GO2-09-043, Independent Spent Fuel Storage Installation - Offsite Dose Calculation Manual, Including Revision Made During 2008: Difference between revisions

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AMENDMENT NO. 47 AUGUST 2008 OFFSITE DOSE CALCULATION MANUAL CHAPTERS 1-5 LIST OF EFFECTIVE PAGES Page    Amendment            Page        Amendment i        38                    18        35 ii      38                    19        35 38                    20        36 iv        38                    21        38 35                    22        38 vi        35                    23        38 1        35                    24        38 2        35                    25          38 3        35                    26          38 4        35                    27        35 5        35                    28        35 6        35                    29        47 7        35                    30        44 8        35                    31        44 9        39                    32        44 10        38                    33        44 11        38                    34        38 12        35                    35        35 13        35                    36        35 14        35                    37        35 15        35                    38        35 16        35                    39        35 17        35                    40        35 LEP-1
 
AMENDMENT NO. 47 AUGUST 2008 OFFSITE DOSE CALCULATION MANUAL CHAPTERS 1-5 LIST OF EFFECTIVE PAGES Page  Amendment              Page      Amendment 41      35                      64        44 42      38                      65        44 43      38                      66        44 44      35                      67        44 45      35                    .68          44 46        35                    69          44 47        35                    70          44 48        35                    71          35 49        38                    72          44 50      47                      73          35 51      47                      74        47 52      35                      75        40 53      35                      76        47 54      35                      77        35 55      35                      78        35 56      35                      79        35 57      35                      80        38 58      35                      81        35 59      35                      82        35 60      35                      83        35 61      44                    84          35 62      44                    85          44 63      44                    86          44 LEP-2
 
AMENDMENT NO. 47 AUGUST 2008 OFFSITE DOSE CALCULATION MANUAL CHAPTERS 1-5 LIST OF EFFECTIVE PAGES Page    Amendment            Page      Amendment 87      44 88      35 89      35 90      35 LEP-3
 
AMENDMENT NO. 38 APRIL 2003 OFFSITE DOSE CALCULATION MANUAL TABLE OF CONTENTS Sectior 1l                        Title                                                                                Page
 
==1.0    INTRODUCTION==
        ..............                    . ... ... ... .... .... ... .... .... . I 2.0      LIQUID EFFLUENT CALCULATIONS .................                                                                      3 2.1  Introduction .....        .............                  ...............  . ............                  3 2.2  Radwaste Liquid Effluent Radiation Monitoring System ...                    .  . . .  .  . .  . . . .  . 3 2.3  10 CFR 20 Concentration Limits ..................                            . ............                  3 2.3.1  Pre-Release Calculation ....................                          . .. .............              4 2.3.2 Post-Release Calculation ...................                          ..... ..........                5 2.4    Radwaste Liquid Effluent Dilution Ratio and Alarm Setpoints Calculations .                            .. 5 2.4.1  Introduction ...........................                                                              5 2.4.2 Methodology for Determining the Total Effluent Concentration Limit (ECL) Fraction .......................                                            6 2.4.3 Methodology for the Determination of Minimum Dilution Factor .....                                      6 2.4.4 Methodology for the Determination of Liquid Effluent Monitor Setpoints 8 2.5    10 CFR 50, Appendix I, Dose Calculation ........................                                            10 2.5. 1 Projection of Doses ...................................                                              12 2.6  Methods for Calculating Doses to Man From Liquid Effluent Pathways .....                                    12 2.6.1  Radiation D oses          ........................                    ........ ..                  13 2.6.2 Plant Param eters ....................................                                                16 2.7    Liquid Holdup Tanks ......................................                                                  17 2.7.1  Maximum Allowable Liquid Radwaste Activity in Temporary Radwaste Hold-Up Tanks ..............................                                              17 LDCN-03-007 i
 
AMENDMENT NO. 38 APRIL 2003 OFFSITE DOSE CALCULATION MANUAL                                                                      I TABLE OF CONTENTS Section                        Title                                                                                    Paae 2.7.2 Maximum Allowable Liquid Radwaste in Tanks That Are Not Surrounded by Liners, Dikes, or Walls                      . . . . . . . . . . . . . . . . . . . . 19 2.8  Other Liquid Radiation Monitors ...........................                                        ... . 20 3.0    GASEOUS EFFLUENT DOSE CALCULATIONS ...................                                                  ....      29 3.1  Introduction..........................................                                              ....      29 3.2  Gaseous Effluent Radiation Monitoring System ..................                                        .... 30 3.2.1  Reactor Building Elevated Discharge Radiation Monitor .......                                ... 30 3.2.2 Radwaste Building Ventilation Exhaust Monitor .............                                    ... 31 3.2.3 Turbine Building Ventilation Exhaust Monitor ..............                                    ... 31 3.3  Release Rate Limits ...            ................................                                  ... 31 3.3.1  Noble Gases .....................................                                              ... 31 3.3.2 Radioiodines and Particulates .........................                                        ... 32 3.3.2.1            Dose Parameter for Radionuclide i (P) .........                            ... 34 3.4  Calculation of Gaseous Effluent Monitor Alarm Setpoints ...........                                  ... 35 3.4.1  Introduction .............................                                                    ... 35 3.4.2 Setpoint Determination for all Gaseous Release Paths                                            ... 35 3.4.2.1      Setpoints Calculations Based on Whole Body Dose Limits                              . . 35 3.4.2.2      Setpoints Calculations Based on Skin Dose Limits .......                                38 3.5  10 CFR 50 Dose Calculation ..................................                                                38 3.5.1  Noble Gases .. ......................................                                                39 LDCN-03-007 ii
 
AMENDMENT NO. 38 APRIL 2003 OFFSITE DOSE CALCULATION MANUAL TABLE OF CONTENTS Section                        Title                                                        Page 3.5.2 Radioiodines, Tritium and Particulates ......................              41 3.5'.2.1  Dose Parameter for Radionuclide i (R1) ..............        42 3.5.3 Annual Doses AtSpecial Locations ........................                  48 4.0    COMPLIANCE WITH 40 CFR 190 - TOTAL DOSE .....................                          80 4.1  Requirement For Operability (RFO) .............................                  80 4.2  ODCM Methodology for Determining Dose and Dose Commitment from Uranium Fuel Cycle Sources ..............................                  80 4.2. 1 Total Dose from Liquid Effluents ..........................              80 4.2.2 Total Dose from Gaseous Effluents ........................                80 4.2.3 Direct Radiation Contribution ....................                        81 5.0    RADIOLOGICAL ENVIRONMENTAL MONITORING ...................                              82 5.1    Radiological Environmental Monitoring Program (REMP) ...............            83 5.2    Land Use Census . ........................................                      83 5.3    Laboratory Intercomparison Program ............................                83 APPENDIX Radiological Effluent and Radiological Environmental Monitoring Controls LDCN-03-007 iii
 
AMENDMENT NO. 38 MARCH 2002 OFFSITE DOSE CALCULATION MANUAL LIST OF TABLES Section                        Title                                                          Page Section 2.0 2-1    Fish Bioaccumulation Factors (BF      3) 1 and Adult Ingestion Dose Conversion Factors (DF*) .......................                21 2-2    Ingestion Dose Factors (Ai,) for Whole Body and Critical Organ          .............. 24 2-3    Input Parameters Used to Calculate Maximum Individual Dose from Liquid Effluents . 27 Section 3.0 3-1    Dose Factors for Noble Gases and Daughters ...........................                  49 3-2    Distances (Miles) to Typical Controlling Locations as Measured from Center of Columbia Generating Station Containment Building ....          50 3-3    Long-Term Average Dispersion (X/Q) and Deposition (D/Q) Values for Typical Locations ..........................                  51 3-4    Dose Rate Parameters Implementation of 10 CFR 20, Airborne Releases .........            52 3-5a    Dose Parameters for 10 CFR 50 Evaluations, Airborne Releases A ge G roup: A dult ............................................                        54 3-5b    Dose Parameters for 10 CFR 50 Evaluations, Airborne Releases Age Group: Teen .............................................                            55 3-5c    Dose Parameters for 10 CFR 50 Evaluations, Airborne Releases Age Group: Child ..............................................                          56 3-5d    Dose Parameters for 10 CFR 50 Evaluations, Airborne Releases Age Group: Infant .............................................                          57 3-6    Input Parameters for Calculating R c. ................................                  58 LDCN-03-007 iv
 
AMENDMENT NO. 35 MARCH 2002 OFFSITE DOSE CALCULATION MANUAL LIST OF TABLES Table                      Title                                                                                                  Pane M
3-7  Input Parameters for Calculating R.......................                                              ...............        59 V
3-8  Input Parameters for Calculating R                                  ...................................                        60 3-9  Input Parameters for Calculating Dose Summaries to the Maximum Individual and the Population Within 50 Miles of Columbia Generating Station from G aseous Effluents ................................................                                                            61 3-10  Reactor Building Stack X/Q and D/Q Values ...............................                                                      63 3-11  Turbine or Radwaste Building X/Q and DIQ Values ..........................                                                    67 3-13  Characteristics of Gaseous Effluent Release Points ...........................                                                71 3-14  References for Values Listed in Tables 3-8 and 3-9 ...........................                                                72 3-15  Design Base Percent Noble Gas (30-minute Decay) ............................                                                  73 3-16  Annual Doses at Typical Locations ......................................                                                      74 3-17  Annual Occupied Air Dose at Typical Locations                          .. ...........................                        75 Section 5.0 5-1    REM P Locations . ....................................                                                        ............. 85 V
 
AMENDMENT NO. 35 MARCH 2002 OFFSITE DOSE CALCULATION MANUAL LIST OF FIGURES Figure                    Title                                                                      Page 2-1    Simplified Block Diagram of Liquid Waste System ...........................                        28 3-1    Site Boundary for Radioactive Gaseous and Liquid Effluents            ....................        76 3-2    Simplified Block Diagram of Gaseous Waste System ...........................                      77 3-3    Simplified Block Diagram of Off-Gas Treatment System ........................                    78 3-4    Auxiliary Boiler ..  ................................................                            79 5-1    Radiological Environmental Monitoring Sample Locations Inside of 10-Mile Radius .....            89 5-2    Radiological Environmental Monitoring Sample Locations Outside of 10-Mile Radius ....            90 vi
 
AMENDMENT NO. 35 MARCH 2002 10    INTRODUCTION The Offsite Dose Calculation Manual (ODCM) is required by Technical Specification Section 5.5.1. This Technical Specification section and the ODCM implement the requirements of 10 CFR 20, 10 CFR 50 and 40 CFR 190 that apply to radioactive effluents from the plant.
The ODCM contains the methodology and parameters used in the calculation of offsite.
radionuclide concentrations and doses resulting from liquid and gaseous radioactive effluents. In most cases, these doses must be calculated, as the amount of dose received outside the plant is too low-to be measurable. The ODCM also describes the methodology used to establish gaseous and liquid effluent monitoring alarm and trip setpoints.
The ODCM also contains the Radiological Effluent and Environmental Monitoring Controls.
The Radiological Effluent Controls Program implements the requirements of Technical Specification Section 5.5.4. The Radiological Effluent Controls were previously known as the Radiological Effluent Technical Specifications (RETS). The Radiological Environmental Monitoring Controls are part of the Radiological Environmental Monitoring Program (REMP).
The REMP controls supplement the effluent controls by verifying that measurable concentrations of radioactive materials in the environment are not greater than predicted by the effluent controls. At Columbia Generating Station, the Radiological Effluent and Environmental Monitoring Controls are contained in the ODCM Appendix. They are written in the same format as the Licensee Controlled Specifications (LCS). This common format allows for ease of use by the plant staff.
The Solid Radioactive Waste or Process Control Program (PCP) is not part of the ODCM, except for Radioactive Effluent Release Report Requirements provided in the ODCM Appendix Section 6.4. FSAR Section 11.4.3 describes the requirements for the PCP. This includes the objective program requirements, process control systems, and waste characterization. Plant Procedures implement the PCP.
The ODCM reflects the requirements of both 10 CFR 20.1-20.602 (the old Part 20) and 10 CFR 20.1001-20.2402 (the new Part 20). Technical Specification 5.5.4c requires monitoring, sampling, and analysis of radioactive liquid and gaseous effluents pursuant to 10 CFR 20.1302. 10 CFR 20.1302 requires surveys of effluents in unrestricted areas for the purpose of demonstrating compliance with 10 CFR 20.1301. Section 1301 sets dose limits for members of the public. These limits are stated in terms of Total Effective Dose Equivalent (TEDE). 10 CFR 20.1-20.602 (the original Part 20 requirements) as well as 10 CFR 50, Appendix I and 40 CFR 190 regulate dose in terms of dose to a critical organ. The NRC addressed this issue in a series of Questions and Answers on this topic in 1995. The Answers state that 10 CFR 50, Appendix I is not changed concurrently with the new Part 20. Therefore, until Appendix I is changed licensees must continue to show compliance to the new Part 20 based on Appendix I, with dose calculations expressed in terms of organ and whole body doses.
The computer codes LADTAP II and GASPAR II may be used to perform these calculations.
For liquid effluents, however, the concentration limits are to be changed to reflect the new values in 10 CFR 20.1001-20.2402, Appendix B. The corresponding bases and certain alarm 1
 
AMENDMENT NO. 35 MARCH 2002 set points for the liquid monitors are also to be changed. The methodology used in the ODCM to determine the necessary dilutions are required to be changed from the old Maximum Permissible Concentration (MPC) to the now required Effluent Concentration Limits (ECLs).
The Columbia Generating Station ODCM has been changed to accommodate the above requirements resulting from the new 10 CFR 20.1001-20.2402.
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AMENDMENT NO. 35 MARCH 2002 2.0      LIQUID EFFLUENT CALCULATIONS Liquid effluent calculations are discussed in this section of the OD.CM. Calculations of liquid effluent concentrations are made to show compliance with the limits of 10 CFR 20. The methodology for calculating dilution factors and liquid effluent monitor setpoints are also discussed. This section outlines the methods used to calculate off-site radiation dose to show compliance to 10 CFR 50, Appendix I. This section also presents the calculations used to establish the maximum allowable liquid radioactive waste activity available in temporary hold-up tanks in accordance with Technical Specification 5.5.8.b.
2.1    Introduction Liquid radwaste released from Columbia Generating Station will meet control limits at a single point of discharge to the Columbia River. Actual discharges of liquid radwaste effluents will only occur on a batch basis, and the average concentration at the point of discharge will be only a small percentage of the allowed limits. A simplified block diagram of the liquid waste management system and effluent pathways is contained in Figure 2-1.
2.2      Radwaste Liquid Effluent Radiation Monitoring System This monitoring subsystem measures the radioactivity in the liquid effluent prior to its entering the cooling tower blowdown line.
All radwaste effluent passes through a four-inch line which has an off-line sodium iodide radiation monitor. The radwaste effluent flow, (variable from 0 to 190 gpm), combines with the 36-inch cooling water blowdown line, (variable from 0 to 7500 gpm) and is discharged to the Columbia River with a total flow based on the total Effluent Concentration Limit (ECL) fraction and cooling water flushing needs, Additional information on this monitoring system can be found in FSAR Section 11.5.2.2.2.3.
2.3      10 CFR 20 Concentration Limits The requirements pertaining to discharge of radwaste liquid effluents to the unrestricted area are specified in Requirement for Operability 6.2.1.1. In order to comply with this requirement, limits will be set to ensure that blowdown line concentrations do not exceed ten times 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2 at any time.
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AMENDMENT NO. 35 MARCH 2002 2.3.1    Pre-Release Calculation The activity of the radionuclide mixture to be discharged and the liquid effluent discharge rate will be determined prior to discharge in accordance with Energy Northwest procedures. The effluent concentration for each nuclide is determined by the following equation:
Conc.      Ci X  fWv ft                                                                (1) where:
Conci    =      Concentration of radionuclide i in the effluent at point of discharge
                        -  fCi/ml.
Ci      =      Concentration of radionuclide i in the batch to be released  -/xCi/ml.
fw      =      Discharge flow rate from sample tank to the blowdown line - variable from 0 to 190 gpm.
tb      =      Blowdown flow rate    - variable from 0 to 7500 gpm.
ft      =      Total discharge (ft = fb + fw) flow rate - variable from 0 to 7690 gpm.
The calculated concentration in the blowdown line must be less than ten times the concentrations listed in 10 CFR 20.1001-20.2402, Appendix B. Before releasing the batch to the environment, the following equation must hold:
m ECLf = *        (Conci /ECLi) < 1                                                              (2) i=.l where:
ECLr =          Total ECL Fraction Conci    =      Concentration of radionuclide i in the effluent at the point of discharge -
ttCi/ml.
ECLi      =      Ten times the effluent concentration of nuclide i listed in 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2.
m        =      Total number of radionuclides in the batch.
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AMENDMENT NO. 35 MARCH 2002 2.3.2 Post-Release Calculation The concentration of each radionuclide in the unrestricted area, following the batch release, will be calculated as follows:
The average activity of radionuclide i during the time period of the release is divided by the Plant Discharge Flow/Tank Discharge Flow ratio yielding the concentration at the point of discharge:
ik X  fw Conc ik  -
ft                                                                          (3) where:
Concik  =      The concentration of radionuclide i in the effluent at the point of discharge during the release period k - (guCi/ml).
Cik            The concentration of radionuclide i in the batch during the release period k
                        - (/ACi/ml).
fw      =      Discharge flow rate from sample tank to the blowdown line - variable from 0 to 190 gpm.
fb      =      Blowdown flow rate - variable from 0 to 7500 gpm.
ft      =      Total discharge (ft = fb + fw) flow rate - variable from 0 to 7690 gpm.
To ensure compliance with 10 CFR 20.1001-20.2401, the following relationships must hold:
m E (Concik /ECLi)        1                                                                    (4) where the terms are as defined in Equation (2) and (3).
2.4    Radwaste Liquid Effluent Dilution Ratio and Alarm Setpoints Calculations 2.4. 1 Introduction The dilution alarm ratio and setpoints of the sample liquid effluent monitor are established to ensure that the limits of ten times 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2, are not exceeded in the effluent at the discharge point (i.e., compliance with RFO 6.2.1.1, as discussed in section 2.3.1 of this manual).
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AMENDMENT NO. 35 MARCH 2002 The alarm (HI) and the alarm/trip (HI-HI) setpoints for the liquid radwaste effluent monitor are calculated from the results of the radiochemical analysis of the pre-release effluent sample. The setpoints are set into the radwaste monitor just prior to the release of each batch of radioactive liquid.
2.4.2 Methodology for Determining the Total Effluent Concentration Limit (ECL) Fraction Radwaste liquid effluents can only be discharged to the environment through the four-inch radwaste line. The maximum radwaste discharge flow rate is 190 gpm. Prior to discharge, the tank is isolated and recirculated for at least thirty minutes, and a representative sample is taken from the tank. An isotopic analysis of the batch will be made to determine the sum of the ECL fraction (ECL) based on ten times the 10 CFR 20.1001-20.2402 limits. From the sample analysis and ten times the effluent concentration values in 10 CFR 20.1001-20.2402, the ECLf is determined using the following equation.
ECLf = E  m      Ci                                                                              (5) i=1  ECLi where:
ECLf            Total ECL fraction in the liquid effluent waste sample.
C        =      The concentration of each measured radionuclide i observed by the radiochemical analysis of the liquid waste sample (AtCi/ml).
ECL    =      Ten times the limiting concentrations of radionuclide i from 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2.OE-04 jiCi/ml total activity.
m      =      The total number of measured radionuclides in the liquid batch to be released.
If the ECLf is less than or equal to 0.8, the liquid batch may be released at any radwaste discharge or blowdown rate. If the ECLf exceeds 0.8, then a dilution factor (Fd) must be determined. The liquid effluent radiation monitor responds proportionally to radioactivity concentrations in the undiluted waste stream. Its setpoint must be determined for diluted releases.
2.4.3    Methodology for the Determination of Minimum Dilution Factor The measured radionuclide concentrations are used to calculate the dilution factor (Fd), which is the ratio of the total discharge flow rates (RW + CBD) to the radwaste tank effluent flow rate (RW) that is required to assure that the limiting concentrations of RFO 6.2. 1. 1 are met at the point of discharge.
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AMENDMENT NO. 35 MARCH 2002 The minimum dilution factor (Fd) is determined according to:
Fd =ECLf x Fs                                                                                    (6) where:
Fd      -    The minimum dilution factor required for compliance with ten times 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2.
Fs      =      The safety factor; a conservative factor used to compensate for statistical fluctuations and errors in measurements. For example, a safety factor (Fs) of 1.5 corresponds to a fifty (50) percent (%) variation. The safety factor is 1.5.
The dilution which is required to ensure compliance with RFO 6.2.1.1 concentration limits will be set such that discharge rates are:
Fd _* fw    + fb fw                                                                                  (7) and follows that:
fb fw*                                                                                          (7a)
Fd-1 or fb _ fw(Fd-1)                                                                                (7b) where:
Fd      -      The minimum dilution factor from Equation (6).
fw      =      The discharge flow rate from the liquid radwaste tank to the blowdown line - variable from 0 to 190 gpm.
fb      =      The cooling tower blowdown flow rate - variable from 0 to 7500 gpm.
7
 
AMENDMENT NO. 35 MARCH 2002 2.4.4 Methodology for the Determination of Liquid Effluent Monitor Setpoints Liquid effluents must meet the restrictions at the point of discharge to the river of ten times ECL or less after dilution. Therefore, the Liquid Effluent Monitor setpoint must be determined such that it will terminate a discharge at less than or equal to that point. The dilution factor must satisfy Equation (7).
Setpoint      CM  (fb  + fw)                                                                    (8) where:
Setpoint      the radwaste effluent monitor setpoint in ItCi/ml.
CM              the maximum permissible diluted concentration, in 1 Ci/ml, at the point of release that is in compliance with ten times 10CFR20.1001-20.2402 Appendix B, Table 2, column 2.
fb and fw are defined in equation 7b.
The Liquid Effluent Monitor measures the undiluted effluent, therefore the term fb  + fw      is used to correct for dilution.
fw The total ECL fraction of the batch to be discharged, ECLf is defined in Equation (5). Since the final concentration must be less than or equal to one ECLf:
M      CL)C        Ci                                                              (8a)
Substituting into Equation (8):
Setpoint  _(ECLf)(*        Ci)  (f,+  _fw)                                        (8b) 8
 
AMENDMENT NO. 39 MAY 2003 The Liquid Effluent Monitor reads out in counts per second (cps), therefore, it is necessary to convert the setpoint from kCi/ml to cps.
SHIMI <      C        i                    fw.
S      ( 0.81    (m                    _
where:
SHIn    =      the trip setpoint in cps E      =      the monitor efficiency for nuclide i, in cps/pCi/ml BKG    =      the monitor background in cps.
At low activity levels, the monitor demonstrates a normal instrument variation. In order to prevent spurious alarms and trips resulting from this variation, the setpoint can be calculated using a 1.0 ECLf representative mixture when the ECL, of the batch is less than 1.0 ECLU.
The effluent monitor also has a high alarm setpoint that will' be set to alarm if the batch contents exceed the concentration expected for the current discharge. This will warn the operator that the batch release is not proceeding as anticipated by the prerelease calculation. The discharge should be stopped and the alarm cause investigated. The Hi alarm setpoint is determined to be at the monitor response for the current batch release multiplied by 1.25 to allow for normal variation in the monitor response. When the ECLf of the batch is less than 1.0 ECLf, the high setpoint will be the greater of either the calculated setpoint, or 80% of the setpoint determined from a 1.0 ECLf mixture.
m SH
* BKG + 1.25E (C1 ) (E1)                                                                    (9a) where:
SH  is the monitor Hi setpoint in cps.
1.25 is a factor to account for normal variation in the monitor reading. It results in a maximum of a 25 % greater than expected count rate before the alarm occurs.
S*-! BKG + (0.8
* One-ECL)                                                                        (9b)
Where one-ECL is the count rate corresponding to a 1.0 ECLf representative mixture. All other terms defined in Equation 9a.
Setpoints established for periods when no release is taking place are described in the implementing procedures.
LDCN-03-010 9
 
AMENDMENT NO. 38 APRIL 2003 2.5      10 CFR 50, Appendix I, Dose Calculation Surveillance Requirement 6.2.1.2.1 requires that the cumulative dose contributions be determined in accordance with the ODCM at least once per 31 days. RFO 6.2.1.2 specifies that the dose to a MEMBER OF THE PUBLIC from radioactive material in liquid effluents released to the UNRESTRICTED AREA shall be limited to:
1.5 mrem/Calendar Quarter - Whole Body and
            <5.0 mrem/Calendar Quarter - Any Organ.
The cumulative dose for the calendar year shall be limited to:
            !3 mrem - Whole Body and
            *10 mrem - Any Organ.
The maximum exposed individual is distinguished as "maximum" with regard to usage factors which describe food intake, occupancy, breathing rate, and other factors of the area in the region of the plant site. It is assumed to be an adult whose exposure pathways include potable water and fish consumption. The choice of an adult as the maximum exposed individual is based on the highest fish and water consumption rates shown by that age group and the fact that most of the dose from the liquid effluent comes from these two pathways.
The dose contribution will be calculated for all radionuclides identified in the liquid effluent released to the unrestricted area, using the following equation:
m Dc=            (Air      &#xfd;,t0CipFd)                                                                (10) i          0=1 where:
D'        =      The cumulative dose commitment to the wholein  body or organ, 'r, from liquid effluents for the total time period ' A t in mrem.
0=1 t
t,          =      The length of the Qth time period over which C, and F, are averaged for all liquid releases, in hours.
m                    The number of releases for the time period under consideration.
C=                  The average concentration of radionuclide i in undiluted liquid effluent during time period m, from any liquid release, in jiCi/ml.
LDCN-03-007 10
 
AMENDMENT NO. 38 APRIL 2003 A=                The site-related ingestion dose commitment factor to the whole body or any organ -r for each identified principle gamma and beta emitter listed in Table 2-2, in mrem/hr per pCi/ml.
F,        =      The near field average dilution factor for Ci, during any liquid waste release. This is defined as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 500.
While the actual discharge structure exit flow is variable from 0 to 17.1 cfs (0 to 7690 gpm), a maximum flow value of 2.0 cfs will be used for dose calculation purposes in accordance with the NUREG-0133 requirement that the product of the average blowdown flow to the receiving water body, in cfs, and the applicable factor (500), is 1000 cfs or less.
(Fe        Liquid Radioactive Waste Flow            _    fw                                      (11)
Discharge Structure Exit Flow x 500          ft x 500 The term A1 , the ingestion dose factors for the whole body and critical organs, are tabulated in Table 2-2. It embodies the dose factor, fish bioaccumulation factor, pathway usage factor, and the dilution factor for the plant diffuser pipe to the Richland potable water intake. The following equation was used to calculate the ingestion dose factors:
AiI = Ko (Uw/Dw      + UF  BFi) DFi                                                          (12) where:
A      =      The composite dose parameter for whole body or critical organ of an adult for nuclide i (in mrem/hr per ACi/ml).
Ko              A conversion factor:
1.14E+05    = (106  pCi/lCi) x (10' ml/liter)/8760 hr/yr.
U      =      730 liter/yr - which is the annual water consumption by the maximum adult (Table E-4 of Regulatory Guide 1.109, Revision 1).
BFi            Bioaccumulation factor for radionuclide i in fish - (pCi/Kg per pCi/liter)
(Table A-I of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013).
DFi            Adult ingestion dose conversion factor for nuclide i - Whole body or critical organ, ti, in (mrem/pCi) (Table E-1 1 of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013).
LDCN-03-007 II
 
AMENDMENT NO. 35 MARCH 2002 D,,            Dilution factor from near field area (within one-quarter mile of the release point) to the Richland potable water intake - 100.
UF              Adult fish consumption, 21 kg/yr (Table E-5 of Regulatory Guide 1.109, Revision 1).
The values of BFi and DF1 are listed in Table 2-1.
The U.S. Nuclear Regulatory Commission's computer program LADTAP II can be used for dose analysis for liquid radioactive effluents from Columbia Generating Station into surface waters. The analysis estimates the radiation dose to individuals, population groups, and biota from ingestion (aquatic foods, water, and terrestrial irrigated foods) and external exposure (shoreline, swimming, and boating) pathways. Dilution assumptions, calculations, and LADTAP II input parameters are provided in Radiological Health Calculation Log 92-2.
2.5.1  Proiection of Doses The projected doses due to releases of radwaste liquid effluents will be calculated for each batch, using Equation (10) or LADTAP II: If the sum of the accumulated dose to date for the month and the projected dose for the remainder of the month exceeds the RFO 6.2.1.3 limits, then the liquid radwaste treatment system shall be used. This is to ensure compliance with RFO 6.2.1.3.
Dose projections are performed by taking the ratio of the time period analyzed to the time period projected as described in plant procedures.
2.6    Methods for Calculating Doses to Man From Liquid Effluent Pathways Dose models presented in NRC Regulatory Guide 1.109, Revision 1, as incorporated in the LADTAP II computer code, will be used for offsite dose calculation. The details of the computer code, and user instruction, are included in NUREG/CR-4013, "LADTAP. II -
Technical Reference and User Guide."
12
 
AMENDMENT NO. 35 MARCH 2002 2.6.1    Radiation Doses Radiation doses from potable water, aquatic food, shoreline deposit, and irrigated food pathways will be calculated by using the following equations:
Potable Water 110UapMp      E Rapi        1100      F      i  QiDaipjexp(-Aitp                                            (13)
Aquatic Foods R.apj        1100 UapMp        i QjBjPDajPjexp(-AitP)                                          (14)
F Shoreline Deposits 1000UapM pW    E Rap    =  Io              F      i QiTiDaipj [exp (-Aitp) (1 - exp(-AXitb)]              (15)
Irrigated foods For all radionuclides except tritium:
Rapj      Uveg      E di  exp(-,&#xfd;ith)Daipj  [r [1 - exp(-    ,Eit)]
ap                              [      Yv ,Ei          + f~jv1 -exp(-kitb)I]
PA1          I U
                +
ppanimal  iY F ADaipj    dexp(-epEite] r[D Qdfexp(-Aith)[-
          " f Biv [1 - exp(-Ajtb)]      + CiAwQAW (16)
PKi For tritium:
apj=Raju ap veg  Cvaj+Uap apj        animal DapjFA(CvQF + CAwQAw)                      (17) 13
 
AMENDMENT NO. 35 MARCH 2002 where:
Bip        The equilibrium bioaccumulation factor for nuclide i in pathway p, expressed as the ratio of the concentration in biota (in pCi/kg) to the radionuclide concentration in water (in pCi/liter), in liters/kg.
B      =,  The concentration factor for uptake of radionuclide i from soil by edible parts of crops, in pCi/kg (wet weight) per pCi/kg dry soil.
CiAw  =    The concentration of radionuclide i in water consumed by animals, in pCi/liter.
C      =-  The concentration of radionuclide i in vegetation, in pCi/kg.
Da~j  =    The dose factor specific to a given age group a, radionuclide i, pathway p, and organ j, which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi, or from exposure to a given concentration of a radionuclide in sediment, expressed as a ratio of the dose rate (in mrem/hr) and the area radionuclide concentration (in pCi/m2 ).
di          The deposition rate of nuclide i in pCi/mi  per hour.
F          The flow rate of the liquid effluent, variable from 0 to 2.0 cfs, for dose calculation purposes.
f      =    The fraction of the year crops are irrigated, dimensionless.
FiA    -  The stable element transfer coefficient that relates the daily intake rate by an animal to the concentration in an edible portion of animal product, in pCi/liter (milk) per pCi/day or pCi/kg (animal product) per pCi/day.
MP          The mixing ratio (reciprocal of the dilution factor) at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food), dimensionless.
P      =    The effective "surface density" for soil, in kg (dry soil)/m 2 (Table E-15, Regulatory Guide 1.109, Revision 1).
QAw  =    The consumption rate of contaminated water by an animal, in liters/day.
QF    =    The consumption rate of contaminated feed or forage by an animal, in kg/day (wet weight).
14
 
AMENDMENT NO. 35 MARCH 2002 Q    =  The release rate of nuclide i in Ci/yr.
r    =  The fraction of deposited activity retained on crops, dimensionless (Table E-15, Regulatory Guide 1.109, Revision 1).
RPj    The total annual dose to organ j of individuals of age group a from all of the nuclides i in pathway p, in mrem/yr.
tb      The period of time for which sediment or soil is exposed to the contaminated water, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).
te      The time period that crops are exposed to contamination during the growing season, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).
th      A holdup time that represents the time interval between harvest and consumption of the food, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).
Tj    =  The radioactive half life of nuclide i in days.
The average transit time required for nuclides to reach the point of exposure. For internal dose, tP is the total time elapsed between release of the nuclides and ingestion of food or water, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).
Uap    A usage factor that specifies the exposure time or intake rate for an individual of age group a associated with pathway p, in hr/yr, L/yr, or kg/yr (Table E-5, Regulatory Guide 1.109, Revision 1).
W      The shoreline width factor, dimensionless (Table A-2, Regulatory Guide 1.109, Revision 1).
Y    =  The agricultural productivity (yield), in kg (wet weight)/m 2 (Table E-15, Regulatory Guide 1.109, Revision 1).
XEi    The effective removal rate constant for radionuclide i from crops, in hr',
where    iE,= Xi + Xw  Xi is the radioactive decay constant, and Xw is the removal rate constant for physical loss by weathering (Regulatory Guide 1.109, Revision 1, Table B-15).
Aj      The radioactive decay constant of nuclide i in hr-'.
15
 
AMENDMENT NO. 35 MARCH 2002 1100  =      The factor to convert from (Ci/yr)/(ft3 /sec) to pCi/liter.
110,000 =        The factor to convert from (Ci/yr)/(ft3 /sec) to pCi/liter and to account for the proportionality constant used in the sediment radioactivity model.
These equations yield the dose rates to various organs of individuals from the exposure pathways mentioned above.
2.6.2 Plant Parameters Columbia Generation Station is a river shoreline site with a variable effluent discharge flow rate 0 to 7690 gpm. The population center nearest the site is the city of Richland, where drinking water withdrawal takes place. The applicable dilution factor is 50,000, using average river flow. The time required for released liquids to reach Richland, approximately 12 miles downstream, is estimated at 4.0 hours. Richland is the "realistic case" location, and doses calculated for the Richland location are typically applicable to the population as a whole.
Individual and population doses based on Richland parameters are calculated for all exposure pathways.
Only the population downstream of the site is affected by the liquid effluents released. There is no significant commercial fish harvest in the 50-mile radius region around Columbia Generating Station. Sportfish harvest is estimated at 14,000 kg/year.
For irrigated foods exposure pathways, it can be assumed that production within the 50-mile radius region around Columbia Generating Station is sufficient to satisfy consumption requirements.
Other relevant parameters relating to the irrigated foods pathways are defined as follows:
Irrigation Rate          Annual Yield        Growing Period Food Type                          (liter/m2 /mo)            (kg/m 2 )              (Days)
Vegetation                              150                    5.0                  70 Leafy Vegetation                        200                    1.5                  70 Feed for Milk Cows                      200                    1.3                  30 Feed for Beef Cattle                    160                    2.0                  130 Source terms are measured based on sampled effluent.
16
 
AMENDMENT NO. 35 MARCH 2002 Table 2-3 summarizes the LADTAP II input parameters. Documentation and/or calculations of these parameters are discussed in detail in PPM 16.12. 1, and Radiological Health Calculation Log 92-2.
2.7      Liquid Holdup Tanks 2.7.1    Maximum Allowable Liquid Radwaste Activity in Temporary Radwaste Hold-Up Tanks The use of temporary liquid radwaste hold-up tanks is planned for Columbia Generating Station.
Technical Specification 5.5.8.b states the quantity of radioactive material contained in any outside temporary tanks shall be limited to the limits calculated in the ODCM such that a complete release of the tank contents would not result in a concentration at the nearest offsite potable water supply that would exceed ten times the limits specified in 10 CFR Part 20.1001-20.2402 Appendix B, Table 2.
Equation (18) will be used to calculate the curie limit for a temporary radwaste hold-up tank.
The total tank activity will be limited to less than or equal to ten (_*10) curies, excluding tritium and dissolved or entrained gases.
The quantity of radioactive material in the hold-up tanks shall be determined to be within the limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
Kd AT =          K fi                                                                              (18)
ECLi eXit where:
AT        =    Total allowed activity in tank (curies).
Ai      =    Activity of radioisotope i (curies).
ECL.    =    Ten times maximum permissible concentration of radionuclide i (10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2).
i=    Decay constant (years-) radioisotope i.
t        =    Transit time of ground water from Columbia Generating Station to WNP-1 well (FSAR Section 2.4) = 67 years.
Fraction of radioisotope        Ai fi        =
17
 
AMENDMENT NO. 35 MARCH 2002
                =        Index for all radioisotopes in tank except tritium and noble gases.
Kd              Dispersion constant based on hydrological parameters, (2.4E+05 Ci per 1 iCi/cc.)
The total allowed activity (AT) is based on limiting WNP-1 well water to less than ten times ECL1 if the entire liquid content of the tank spilled to ground and then migrated via ground water to the WNP-1 well. The WNP-1 well is the location of maximum concentration since it is the nearest source of ground water and conditions are such that no spill of liquid should reach surface water. The 55-60 foot depth of the water table and the low ambient moisture of the soil requires a rather large volume of spillage for the liquid to even reach the water table. However, allowed tank activity (Ar) is conservatively based on all liquid radwaste in the tank instantaneously reaching the water table.
The hydrological analysis performed for the FSAR (Section 2.4) determined that the transit time through the ground water from Columbia Generating Station to the WNP-1 well is 67 years for Strontium and 660 years for Cesium. These two radionuclides are representative of the radionuclides found in liquid radwaste. Strontium is a moderate sorber and Cesium strongly sorbs to soil particles. This calculation conservatively treats all radionuclides as moderate sorbers with a transit time of 67 years.
The concentration of each radionuclide in the well (CWi) is simply the concentration in the tank (CTi) adjusted for radioactive decay during transit (e-t) and divided by the minimum concentration reduction factor (CRFmin). Limiting well concentration to 10 X ECL yields:
CWi                CTie -                                                                    (19)
CRFmin      ECLi ECL, (From Section 2.4 of FSAR) 31 2 (ax ay a.) 1/2 CRIFmin    (4 nt L) 2V                                                                          (20) 2V where:
L                =        Migration distance = 1 mile.
V                =        Volume of tank.
0xX,              =cy,    Dispersion constants.
18
 
AMENDMENT NO. 35 MARCH 2002 Combining Equations (19) and (20) yields:
CTi 2V e-;Lit 1=                                                                                        (21)
(4 r L)fr V(CTx    ay z)1/2 ECLiz Substituting Ai for CTj V and reorganizing terms yields:
(4 rt L) 1/2  (a x Oty  az)U/2            Ai (22) 2                          ECLie 'x;t Making the following substitutions:
Ai = fi AT (4 = L)3 /2 (ax a~ a z)l/2                                        Ci              (23)
K    (4    L310-6                x        Ci/puCi = 2.4 x 10i Ci per A-cc 2                                                    cc yields:
Kd =AT
* ff.
ECLie it or Kd ATK              d
                  " fi ECLie +kit 2.7.2  Maximum Allowable Liquid Radwaste in Tanks That Are Not Surrounded by Liners, Dikes, or Walls Although permanent outside liquid radwaste tanks which are not surrounded by liners, dikes, or walls are not planned for Columbia Generating Station, Equation (18) will be used should such tanks become necessary in the future.
19
 
AMENDMENT NO. 36 JUNE 2002 2.8    Other Liquid Radiation Monitors Other Liquid Radiation Monitors are provided at Columbia Generating Station to monitor the status of normally non-radioactive systems. These monitors are:
* Standby Service Water (SW)
* Turbine Service Water (TSW)
Calculational methodology is not provided in the ODCM for these monitors since their setpoints do not require extensive analysis. Their setpoints are established simply as a fraction of the selected 10 CFR 20 isotope concentration. The ODCM Appendix provides these setpoints in Sections 6. 1. 1 and B6. 1.1. These setpoints are implemented in Plant Procedures.
LDCN-02-007 20
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-1 FISH BIOACCUMULATION FACTORS (BFi)(1)
AND ADULT INGESTION DOSE CONVERSION FACTORS (DFE. -2 Dose Conversion Factor (DF,)
Fish Bioaccumulation  Whole                                            GI Nuclide      Factor (BFL_. Body      Bone          Thyroid      Liver      Tract (pCi/kg per                        (mRem per pCi Ingested) pCi/liter)
H-3              9.OE-01      6.OE-08        (3)      6.OE-08      6.OE-08    6.OE-08 Na-24            1.OE+02      1.7E-06    1.7E-06        1.7E-06    1.7E-06    1.7E-06 P-32            1.OE+05      7.5E-06    1.9E-04        -(3)      1.2E-05    2.2E-05 Cr-51          2.OE+02      2.7E-09      -(3)        1.6E-09    _(3)        6.7E-07 Mn-54          4.OE+02      8.7E-07      -(3)          -(3)      4.6E-06    1.4E-05 Mn-56          4.OE+02      2.OE-08      _(3)        _(3)        1.2E-07    3.7E-06 Fe-55            1.OE+02    4.4E-07    2.8E-06      _(3)        1.9E-06    1.1E-06 Fe-59          1.OE+02      3.9E-06    4.3E-06      _(3)        1.OE-05    3.4E-05 Co-58          5.OE+01      1.7E-06    _(3)          _(3)      7.5E-07    1.5E-05 Co-60          5.OE+01      4.7E-06      _(3)          -(3)      2.1E-06    4.OE-05 Ni-65          1.OE+02      3.1E-08    5.3E-07        -(3)      6.9E-08    1.7E-06 Cu-64          5.OE+01      3.9E-08    _(3)          _(3)        8.3E-08    7.1E-06 Zn-65          2.OE+03      7.OE-06    4.8E-06        -(3)      1.5E-05    9.7E-06 Zn-69m          2.OE+03      3.7E-08    1.7E-07      _(3)        4.1E-07    2.5E-05 As-76          1L.OE+02    4.8E-06    _(3)            -(3).      -(3)    4.4E-05 Br-82          4.2E +02    2.3E-06    _(3)          _(3)        _(3)      2.6E-06 Br-83          4.2E+02      4.OE-08      .(3)          -(3)        -(3)    5.8E-08 Br-84          4.2E+02      5.2E-08      -(3)          -(3)        -(3)    4.1E-13 Rb-89          2.OE+03      2.8E-08    -(3)            -(3)      4.OE-08    2.3E-21 Sr-89          3.OE+01      8.8E-06    3.1E-04        -(3)        -(3)    4.9E-05 Sr-90          3.OE+01      1.8E-04    8.7E-03        -(3)        -(3)    2.2E-04 Sr-91          3.OE+01      2.3E-07    5.7E-06        -(3)        -(3)    2.7E-05 Sr-92          3.OE+01      9.3E-08    2.2E-06        -(3)        -(3)    4.3E-05 Y-90            2.5E+01      2.6E-10    9.7E-09        -(3)        -(3)      1.OE-04 Y-91m          2.5E+01      3.5E-12    9.1E-11            (3)    _(3)      2.7E-10 Y-91            2.5E+01      3.8E-09    1.4E-07        -(3)        -(3)      7.8E-05 LDCN-03-007 21
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-1 (continued)
FISH BIOACCUMULATION FACTORS (BFi)(1)
AND ADULT INGESTION DOSE CONVERSION FACTORS (DFij1)2)
Dose Conversion Factor (DF.)
Fish Bioaccumulation      Whole                                        GI Nuclide    Factor (BFI-        Body      Bone        Thyroid    Liver    Tract (pCi/kg per                          (mRem per pCi Ingested) pCi/liter)
Y-92        2.5E+01              2.5E-11    8.5E-10      -_(3)      -_(3)    1.5E-05 Y-93        2.5E+01              7.4E- 11  2.7E-09      -_(3)      -_(3)    8.5E-05 Zr-95      3.3E+00              6.6E-09  3. IE-08      -_(3)    9.8E-09    3. 1E-05 Nb-95      3.OE+04              1.9E-09  6.2E-09        -_(3)    3.5E-09    2. i E-05 Zr-97      3.3E+00              1.6E-10    1.7E-09      -_(3)    3.4E-10    1.1E-04 Nb-97      3.0E+04            4.8E-12    5.2E- 11    .- (3)      1.3E-11  4.9E-08 Mo-99      1.OE+01              8.2E-07    __(3)        -_(3)    4.3E-06    1.OE-05 Tc-99m      1.5E+01            8.9E-09    2.5E-10        -_(3)    7.OE-10  4. 1E-07 Tc-101      1.5E+01            3.6E-09    2.5E-10        -_(3)    3.7E-10    1.IE-21 Ru- 103    1.OE+01            8.OE-08    1.9E-07      -_(3)      -_(3)  2.2E-05 Ru- 105    1.0E+01            6. 1E-09    1.5E-08
_(3)        -_(3)  9.4E-06 Rh- 105    1.OE+01            .5.8E-08    1.2E-07      -(3)      8.9E-08    1.4E-05 Ru- 106    1.OE+01            3.5E-07    2.8E-06      _(3)          -_(3)    1.8E-04 Ag-110m    2.3E+00            8.8E-08    1.6E-07      -(3)      1.5E-07  6.OE-05 Sb- 124    1.OE+00              1.1E-06  2.8E-06      6.8E-09    5.3E-08  8.OE-05 Sb- 125    1.OE+00            4.3E-07    1.8E-06      1.8E-09    2.OE-08  2.OE-05 Sb- 126    1.0E+00            4.2E-07    1.2E-06      7.OE-09    2.3E-08  9.4E-05 Sb-127      1.0E+00            9.9E-08    2.6E-07      3.1 E-09    5.7E-09  5.9E-05 Te- 127    4.OE+02            2.4E-08    1.1E-07      8.2E-08    4.OE-08  8.7E-06 Te-129m    4.OE+02              1.8E-06  1.2E-05      4.OE-06    4.3E-06  5.8E-05 Te- 129    4.OE+02            7.7E-09    3. 1E-08    2.4E-08    1.2E-08  2.4E-08 Te-131m    4.OE+02            7. 1E-07  1.7E-06      1.3E-06    8.5E-07  8.4E-05 Te- 131    4.OE+02            6.2E-09    2.OE-08      1.6E-08    8.2E-09  2.8E-09 Te- 132    4.0E+02            1.5E-06    2.5E-06      1.8E-06    1.6E-06  7.7E-05 1-131      1.5E+01            3.4E-06    4.2E-06      2.OE-03    6.OE-06  1.6E-06 1-132      1.5E+01            1.9E-07    2.OE-07      1.9E-05    5.4E-07  1.OE-07 1-133      1.5E+01            7.5E-07    1.4E-06      3.6E-04    2.5E-06  2.2E-06 1-134      1.5E+01            1.OE-07    1.1 E-07    5.OE-06    2.9E-07  2.5E-10 LDCN-03-007 22
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-1 (continued)
FISH BIOACCUMULATION FACTORS (BF1 )(')
AND ADULT INGESTION DOSE CONVERSION FACTORS (DF1)_
Dose Conversion Factor (DF,)
Fish Bioaccumulation    Whole                                        GI Nuclide          Factor (BF)_        Body        Bone        Thyroid    Liver    Tract (pCi/kg per                          (mRem per pCi Ingested) pCi/liter) 1-135            1.5E+01            4.3E-07    4.4E-07    7.7E-05      1.2E-06  1.3E-06 Cs- 134        2.GE+03              1.2E-04    6.2E-05    _(3)        1.5E-04  2.6E-06 Cs- 136        2.OE+03              1.9E-05    6.5E-06      _(3)      2.6E-05  2.9E-06 Cs-137          2.OE+03              7.1E-05    8.OE-05    ._(3)      1.1E-04  2.1 E-06 Cs-138          2.OE+03              5.4E-08    5.5E-08    _(3)        1.1 E-07  4.7E-13 Ba- 139        4.0E+00              2.8E-09    9.7E-08    _(3)        6.9E- 11  1.7E-07 Ba- 140        4.OE+00              1'.3E-06    2.OE-05    _(3)        2.6E-08  4.2E-05 La- 140        2.5E+01              3.3E-10    2.5E-09    _(3)        1.3E-09  9.3E-05 La-141          2.5E+01              1.6E-11    3.2E-10    _(3)        9.9E- 11  1.2E-05 La- 142        2.5E+01              1.5E-11    1.3E-10      _(3)      5.8E- 11  4.3E-07 Ce-141          1.OE+00            7.2E-10    9.4E-09    _(3)        6.3E-09  2.4E-05 Ce-143          1.OE+00            1.4E-10    1.7E-09      _(3)      1.2E-06  4.6E-05 Ce- 144          1.OE + 00          2.6E-08    4.9E-07      _(3)      2.OE-07  1.7E-04 Pr-143          2.5E+01              4.6E-10    9.2E-09    _(3)        3.7E-09  4.OE-05 Nd- 147        2.5E+01              4.4E-10    6.2E-09    _(3)        7.3E-09  3.5E-05 Hf- 179m        3.3E+00              4.8E-06      _(3)        _(3)        _(3)    4. 1E-05 Hf-181          3.3E+00              4.3E-06      _(3)        _(3)        __(3)  4. 1E-05 W-185            1.2E+03            1.4E-08    4.1E-07      _(3)      1.4E-07  1.6E-05 W-187            1.2E+03            3.OE-08    1.OE-07      _(3)      8.6E-08  2.8E-05 Np-239          1.OE+01              6.5E-1 1    1.2E-09    _(3)        1.2E-10  2.4E-05 M')NRC NUREG/CR-4013.
12)Reg.' Guide 1.109.
3
("No  data listed in Reg. Guide 1.109.
(Use whole body dose conversion factor as an approximation.)                      I LDCN-03-007 23
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-2 INGESTION DOSE FACTORS (Ai,) FOR WHOLE BODY AND CRITICAL ORGAN (in mrem/hr per jCi/ml)
Liquid Effluent Whole                                              GI Nuclide      Body      Bone            Thyroid  Liver        Tract H-3          1.8E-01    **            1.8E-01  1.8E-01      1'.8E-01 Na-24        4.1E+02  4.1E+02          4.1E+02  4.1E+02      4.1E+02 P-32        1.8E+06  4.6E+07            **    2.9E+06      5.3E+06 Cr-51        1.3E+00    **            7.7E-01    **        3.2E+02 Mn-54        8.3E+02    **              **    4.4E+03      1.3E+04 Mn-56        1.9E+01    **              **    1.6E+02      3.6E + 03 Fe-55        1.1E+02  6.7E+02            **    4.6E+02      2.6E+02 Fe-59        9.4E+02  1.OE+03            **    2.4E+03      8.2E+03 Co-58        2.1E+02    **              **    9.OE+01      1.8E+03 Co-60      5.7E+02      **              **    2.5E+02      4.8E+03 Ni-65      7.5E+00    1.3E+02            **    1.7E+01      4.1E+02 Cu-64      4.7E +00    **              **    1.OE+01      8.6E+02 Zn-65      3.4E+04    2.3E+04          **    7.2E+04      4.7E+04 Zn-69m        1.8E+02  8.1E+02          **    2.OE+03      1.2E+05 As-76        1.2E+03    **              **      **          1.1E+04 Br-82        2.3E+03    **              **      **        2.6E +03 Br-83        4.OE+01    **              **      **        5.8E+01 Br-84        5.2E+01    **              **      **        4.1E-04 Rb-89        1.3E+02    **              **    1.9E+02      1.1E-11 Sr-89        6.4E+02  2.3E+04            **                  3.6E+03 Sr-90        1.3E+04  6.3E+05            **      **        1.6E+04 Sr-91        1.7E+01  4.1E+02            **      **          2.OE+03 Sr-92        6.8E+00    1.6E+02          **      **          3.1E+03 Y-90        1.6E-02  5.9E-01            **      **          6.1E+03 Y-91m        2.1E-04  5.5E-03            **      **          1.6E-02 Y-91        2.3E-01  8.5E+00            **                  4.7E+03 Y-92        1.5E-03  5.2E-02            **      **          9.1E+02 Y-93        4.5E-03  1.6E-01            **      **          5.2E+03 LDCN-03-007 24
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-2 (continued)
INGESTION DOSE FACTORS (A _i FOR WHOLE BODY AND CRITICAL ORGAN (in mrem/hr per /iCi/ml)
Liquid Effluent Nuclide Whole Body        Bone            Thyroid        Liver GI Tract I
Zr-95        5.3E-02    2.5E-01                          7.9E-02      2.5E+02 Nb-95          1.4E+02  4.5E+02                          2.5E+02      1.5E+06 Zr-97          1.3E-03    1.4E-02                        2.7E-03      8.8E+02 Nb-97        3.5E-01    3.7E + 00                        9.3E-01      3.5E+03 Mo-99        2.0E+01                                    1.1E+02      2.5E+02 9.2E-03 Tc-99m        3.3E-01                                    2.6E-02      1.5E+01 9.2E-03 Tc-101        1.3E-01                                    1.4E-02      4.OE-14 Ru- 103      2.OE+00    4.7E+00                                        5.5E+02 Ru- 105      1.5E-01    3.7E-01                                      2.3E+02 Rh-105        1.4E+00    3.OE+00                        2.2E+00      3.5E+02 Ru- 106      8.7E + 00  6.9E+01                                      4.5E+03 Ag-i10m      5.6E-01    1.OE-00                        9.5E-01      3.8E + 02 Sb- 124      3.6E + 00  9.OE+00          2.2E-02        1.7E-01      2.6E + 02 Sb- 125      1.4E+00    5.8E+00          5.8E-03        6.5E-02      6.5E+01 Sb- 126      1.4E+00    3.9E+00          2.3E-02        7.4E-02      3.GE+02 Sb-127        3.2E-01    8.4E-01          1.0E-02        1.8E-02      1.9E+02 Te- 127      2.3E+01    1.1E+02          7.9E+01        3.8E+01      8.3E+03 Te-129m      1.7E+03    1.2E+04          3.8E+03        4.1 E + 03    5.6E+04 Te-129        7.4E + 00  3.0E+01          2.3E+01        1.2E+01      2.3E+01 Te-131m      6.8E+02    1.6E+03          1.3E+03        8.2E+02      8.1E+04 Te-131        5.9E + 00  1.9E+01          1.5E+01        7.9E+00      2.7E+00 Te-132        1.4E+03    2.4E+03          1.7E+03        1.5E+03      7.4E-04 1-131        1.3E+02    1.5E+02          7.4E+04        2.2E + 02    5.9E+01 1-132        7.OE+00    7.4E +00        7.0E +02      2.GE+01      3.7E+00 1-133        2.8E+01    5. IE +01        1.3E +04      9.2E+01      8. 1E + 01 1-134        3.7E + 00  4.0E+00          1.8E+02        1. IE+01      9.2E-03 1-135        1.6E+01    1.6E+01          2.8E +03      4.4E+01      4.8E+01 Cs-134        5.8E+05    3.OE+05                        7.2E + 05    1.3E+04 Cs-136        9. 1E+04  3. 1E +04                      1.3E+05      1.4E+04 Cs-137        3.44E + 05 3.8E+05                        5.3E+05      1.GE+04 Cs-138      2.6E + 02  2.6E+02                        5.3E+02      2.3E-03 LDCN-03-007 25
 
AMENDMENT NO. 38 APRIL 2003 TABLE 2-2 (continued)
INGESTION DOSE FACTORS (A r) FOR WHOLE BODY AND CRITICAL ORGAN                        I (in mrem/hr per pCi/ml)
Liquid Effluent Whole                                                          GI-          I Nuclide        Body_          Bone          Thyroid        Liver          Tract Ba-139        2.9E-02          1.OE-00          **        .7.2E-04        1.8E+00 Ba- 140        1.4E+01        2.1E+02          **          2.7E-01        4.4E + 02 La- 140      2.0E-02          1.5E-01          **
7.9E-02        5.6E+03 La- 141      9.7E-04          1.9E-02          **          6.OE-03        7.3E+02 La- 142      9. 1E-04        7.9E-03          **
3.5E-03        2.6E+01 Ce- 141      2.3E-03        3.OE-02          **
2.OE-02      7.7E+01 Ce-143        4.5E-04        5.5E-03          **
3.9E + 00      1.5E+02 Ce- 144      8.4E-02          1.6E+00          **
6.5E-01      5.5E+02 Pr- 143      2.8E-02        5.6E-01          **          2.3E-01      2.4E+03 Nd- 147        2.7E-02        3.8E-01          **          4.4E-01        2.1 E + 03 Hf-179m        4.2E+01                          **
3.6E + 02 Hf-181        3.8E+01                          **
3.6E + 02 W-185          4.OE+-01        1.2E+03          **
4.OE+02        4.6E + 04 W-187          8.6E+01        2.9E+02          **
2.5E+02        8. 1E+04 Np-239          S*.6E-03      3.OE-02          **
3.OE-03        6.OE+02
**No Ingestion Dose Factor (DFj) is listed in NUREG/CR-4013.      (Whole body dose I
factor value will be used as an approximation.)
LDCN-03-007 26
 
AMENDMENT NO. 35 MARCH 2002 TABLE 2-3 INPUT PARAMETERS USED TO CALCULATE MAXIMUM INDIVIDUAL DOSE FROM LIQUID EFFLUENTS Drinking Water River Dilution:                      50,000 River Transit Time:                  4 hours Usage Factors:                        Adult = 730 1/yr          Teenager = 510 1/yr Child = 510 1/yr          Infant = 330 1/yr Boatingt and Aquatic Food River Dilution:                      500 Transit Time:                        2 hours Usage Factors: (Aquatic Food)        Adult =  21 kg/yr        Teenager = 16 kg/yr Child =  6.9 kg/yr        Infant = 0 (Boating)                Adult =  100 hr/yr        Teenager = 100 hr/yr Child =  85 hr/yr        Infant = 0 Recreation River Dilution:                      20,000 Shoreline Width Factor:              0.2 Usage Factors:                        Shoreline Activities:      Adult    = 90 hr/yr Teenager = 500 hr/yr Child    = 105 hr/yr Infant = 0 Swimming:                  Adult    = 18 hr/yr Teenager = 100 hr/yr Child      = 21 hr/yr Irrigated Foodstuffs River Dilution:                50,000 River Transit Time:            4 hours Leafy Vegetables            Milk      Meat                Vegetables Food Delivery Time:            14 days            48 hours  20 days              24 hours Usage Factors:
Adult                        520 kg/yr          310 1/yr  110 kg/yr            64 kg/yr Teenager                    630 kg/yr          400 1/yr  65 kg/yr              42 kg/yr Child                        520 kg/yr          330 1/yr  41 kg/yr              26 kg/yr 2          2                    2 Monthly Irrigation Rate:      180 1/m2            200 1/M    160 1/M -            200 1/M 2                  2            2                    2 Annual Yield:                5.0 kg/m              1.3 1/M  2.0 kg/m              1.5 kg/mr Annual Growing Period:        70 days              30 days    130 days              70 days Annual 50-Mile Production:    3.5E+09 kg          2.8E+08 L  2.3E+07 kg            1.9E+06 kg 27
 
AMENDMENT NO. 35 MARCH 2002 COLUMBIA RIVER SIMPLIFIED BLOCK DIAGRAM OF LIQUID WASTE SYSTEM Figure 2-1 28
 
AMENDMENT NO. 47 AUGUST 2008 3.0    GASEOUS EFFLUENT DOSE CALCULATIONS Gaseous effluent dose calculations are discussed in this section of the ODCM. Calculations of gaseous effluent dose rate are made to show compliance with the limits of 10 CFR 20. The methodology for calculating dose rate due to noble gases, radionuclides, and particulates are discussed. Methods used to determine the setpoints for the gaseous effluent monitors are also described. Finally, the methods used to calculate off-site radiation dose due to gaseous effluents in order to show compliance with 10 CFR 50, Appendix I are outlined.
3.1    Introduction Gaseous effluents are released on a continuous basis; in addition, batch releases also occur when containment and mechanical vacuum pump purges are performed and when the off-gas treatment system operates in the charcoal bypass mode.
Figure 3-1 delineates the Site boundary, which for dose calculation purposes, is considered circular with a radius of 1950 meters (approximately 1.212 miles). There are several low occupancy unrestricted locations within the site boundary. These locations, with the exception of the visitor center, are not continuously controlled by Energy Northwest. The locations are:
: 1. Wye burial site (burial ground 618-11) - controlled by DOE.
: 2.      DOE railroad - approximately 3 miles of rail line pass through the site.
: 3.      BPA Ashe Substation - occupied 2080 hours/year. These people are not normally controlled by Energy Northwest but are involved in activities directly in support of Columbia Generating Station.
: 4.      Energy Northwest Visitor Center - assumed occupied 8 hrs/yr by non-Energy Northwest individuals.
: 5.      WNP occupied 2080 hrs/yr. This location is controlled by Energy Northwest.
However, activities are not in direct support of Columbia Generating Station.
: 6.      WNP occupied 2080 hrs/yr. This location is controlled by Energy Northwest.
However, activities are not in direct support of Columbia Generating Station.
All other locations shown in Figure 3-1 support Columbia Generating Station activities and are controlled by Energy Northwest. Figure 3-2 provides a simplified block diagram of the gaseous
.radwaste system for the reactor, turbine and radwaste buildings. Figure 3-3 provides a simplified block diagram for the off-gas treatment system.
LDCN-07-033 29
 
AMENDMENT NO. 44 MAY 2005 Examples of air doses and doses to individuals at these locations were calculated based on the NRC GALE code design base mixture, location-specific estimated occupancy, and X/Qs from 1984-1989 meteorological data using desert sigmas. These doses are listed in Tables 3-16 and 3-17 along with the doses to the maximum exposed individual. In the example calculations, the most exposed member of the public beyond the site boundary was considered to be residing in Taylor Flats (4.2 miles ESE of Columbia Generating Station). This was the residential area with the highest X/Q and D/Q values.
Examples of air doses and doses from ground shine and inhalation to National Guard troops stationed at or within the site boundary were also calculated based on the Gale Code design mixture (FSAR Table 11.3-7), a one year duration of stay, and on meteorological data averaged from 1997 to 2000 plus 2002. The results are tabulated in Tables 3-16 and 3-17 for the two sectors with the highest dispersion and deposition factors.
The Auxiliary Boiler and Seal Steam Evaporator B have occasionally become contaminated with tritium. Seal Steam Evaporator B normally supplies heating steam to the Reactor, Radwaste, Turbine, Diesel Generator, and Service buildings as well as the Liquid Nitrogen pad. When Seal Steam Evaporator B is not in operation, the Auxiliary Boiler provides the heating steam. The heating steam system is designed with multiple vents to the atmosphere and provides a possible source of radioactive effluent when in operation. The contamination levels are monitored and controlled through plant procedures. Makeup water is supplied by normally non-radioactive demineralized water. In 1992, an offsite dose calculation was performed using the current administrative limit of 2.0 E+06 picoCuries per liter of tritium activity in the Auxiliary Boiler and Seal Steam Evaporator B systems. Using NRC Regulatory Guide 1.109 methodology with FSAR Low Population Zone (LPZ) X/Q values and assuming one gallon per minute (1 gpm) makeup flow rate for 180 days plus a one time complete boil-off of the total water inventory, the dose contribution from tritium was calculated as less than one tenth of a millirem per year
( < 0. 1 mrem/yr). Figure 3-4 provides a simplified diagram for the Auxiliary Boiler. Estimates of annual releases from the heating steam are reported in the annual effluent report based on sampling the Auxiliary Boiler and Seal Steam Evaporator B and assuming that the entire source term is released during the heating cycle.
Under certain meteorological conditions, tritiated water vapor that has been released from the plant may condense onto surfaces such as rooftops and exterior walls. Subsequently, this condensed, recaptured, tritiated water may be carried with precipitation into the Storm Drain Pond (SDP) which serves as a collection point for rainfall on plant roofs. Influent to the SDP is continuously sampled and periodically analyzed for tritium content.
LDCN-04-012 30
 
AMENDMENT NO. 44 MAY 2005 3.2    Gaseous Effluent Radiation Monitoring System 3.2.1  Reactor Building Elevated Discharge Radiation Monitor The Reactor Building is continuously monitored for gaseous radioactivity prior to discharge to the environment. The effluent is supplied from: the gland seal exhauster, mechanical vacuum pumps, treated off gas, standby gas treatment, and exhaust air from the entire reactor building's ventilation. Further information on this monitoring system can be found in FSAR Section 11.5.2.2.1.5 and the Radioactive Gaseous Effluent Monitoring Instrumentation Bases, ODCM Appendix, B6.1.2.
3.2.2 Radwaste Building Ventilation Exhaust Monitor The radwaste building ventilation exhaust monitoring system monitors the radioactivity in the Radwaste Building exhaust air prior to discharge. Radioactivity can originate from: radwaste tank vents, laboratory hoods, and various cubicles housing liquid process treatment equipment and systems.
Further information on this monitoring system can be found in FSAR Section 11.5.2.2.1.7 and the Radioactive Gaseous Effluent Monitoring Instrumentation Bases, ODCM Appendix, B6.1.2.
3.2.3 Turbine Building Ventilation Exhaust Monitor This monitoring system detects fission and the activation products from the turbine building air which may be present due to leaks from the turbine and other primary components in the building. Further information on this instrumentation can be found in FSAR Section 11.5.2.2.1.6 and the ODCM Appendix, Section B6.1.2.
3.3      Release Rate Limits Limits for release of gaseous effluents from the site to areas at and beyond the site boundary are stated in RFO 6.2.2.1.
3.3.1    Noble Gases In order to comply with RFO 6.2.2.1, the following equations must hold:
Whole body:
Ki  [(X-Q)m  0im  +  (X/Q)g  Qig)]    <500 mrem/yr                                (1) 31
 
AMENDMENT NO. 44 MAY 2005 Skin:
[(Li
[i    +  1.1Mi) ((X/Q)m      0im +  (X/Q)g 0ig)]
* 3000 mrem/yr                          (2) 3.3.2 Radioiodines and Particulates Part "b" of RFO 6.2.2.1 requires that the release rate limit for all radioiodines and radioactive materials in particulate form and radionuclides other than noble gases must meet the following relationship:
Any organ:
Pi [WM Oim + Wg 0ig]      j<1500 mrem/yr                                                    (3)
The terms used in Equations (1) through (3) are defined as follows:
Ki              The whole body dose factor due to gamma emissions for each identified noble gas radionuclide i (mrem/yr per IiCi/m 3).
Li              The skin dose factor due to beta emissions for each identified noble gas radionuclide i (mrem/yr per pCi/m3 ).
Mi              The air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per ,Ci/M 3 (unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose).
P      =      The dose parameter for all radionuclides other than noble gases for the inhalation pathway, (mrem/yr per pCi/m3 ) and for food and ground plane pathways, m2(mrem/yr per tCi/sec). The dose factors are based on the critical individual organ and the most restrictive age group.
Qim            The release rate of radionuclide i in gaseous effluent from mixed mode release (tiCi/sec). The main plant release point is a partially elevated mixed mode release.
Qig            The release rate of radionuclide i in gaseous effluent from all ground level releases (pCi/sec).
(X/Q)m=        The highest calculated annual average relative concentration for any area at and beyond the site boundary due to a partially elevated mixed mode releases from the main plant vent release point (sec/m 3 ).
32
 
AMENDMENT NO. 44 MAY 2005 (X/Q)g  =      The highest calculated ground level annual average relative concentration for any area at and beyond the site boundary due to Turbine And Radwaste Building releases (sec/m 3).
Wg            The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location due to all ground level releases.
Wg              (sec/m 3 ). For the inhalation pathway. The location is at and beyond the site boundary in the sector of maximum concentration.
Wg              m 2 . For ground plane pathways. The location is at and beyond the site boundary in the sector of maximum concentration.
WM              The highest calculated annual average dispersion parameter for estimating the dose to an individual at the. controlling location due to partially elevated releases:
WM              sec/m3 . For inhalation pathway. The location is at and beyond the site boundary in the sector of maximum concentration.
WM              m-2 . For ground plane pathways. The location is at and beyond the site boundary in the sector of maximum concentration.
The factors, Li and Mi, relate the radionuclide airborne concentrations to various dose rates assuming a semi-infinite cloud. These factors are listed in Table B-i of Regulatory Guide 1. 109, Revision 1, and in Table 3-1 of this manual.
The values used in the equations for the implementation of RFO 6.2.2.1 are based upon the maximum long-term annual average X/Q at and beyond the site boundary. Table 3-2 provides typical receptor locations and pathways for use in dose determinations. Table 3-3 provides these typical locations with long term X/Q and D/Q values which may be used if current annual averages are not available.
The X/Q and D/Q values listed in Tables 3-10 and 3-11 may be utilized in GASPAR II computer runs if current meteorological data is unavailable.
33
 
AMENDMENT NO. 38 APRIL 2003 3.3.2.1          Dose Parameter for Radionuclide i (P)
The dose parameters used in Equation (3) are based on:
: 1.      Inhalation and ground plane. (Note: Food pathway is not applicable since no food is grown at or near the restricted area boundary.)
: 2.      The annual average continuous release meteorology at the site boundary.
: 3.      The critical organ for each radionuclide (thyroid for radioiodine).
: 4.      The most restrictive age group.
The following equation will be used to calculate Pi (Inhalation).
ii P.I (Inhalation)    K A(BR) DFAj (mrem/yr per        11Ci/M 3)                            (4) where:
KA      =      A conversion constant,  106 pCi/1 iCi.
BR      =      The breathing rate of the child age group, 3700 m3/yr.
DFA1    =    The critical organ inhalation dose factor for the child age group for the ith radionuclide in mrem/pCi. The whole body is considered as an organ in the selection of DFAi.
The inhalation dose fact6r for DFAj for the child age group is listed in Table E-9 of Regulatory Guide 1.109, Revision 1, and Table 3-4 of this manual. Resolving the units yields:
: p. (Inhalation) -- (3.7E    + 09) (DFAi) (mrem/yr per jzCi/m 3)                                  (5)
I The Pi (Inhalation) values for the child age group are tabulated in Table 3-4 of this manual.
10 LDCN-03-007 34
 
AMENDMENT NO. 35 MARCH 2002 3.4    Calculation of Gaseous Effluent Monitor Alarm Setpoints 3.4.1 Introduction The following procedure is used to ensure that the dose rate at or beyond the site boundary due to noble gases in the gaseous effluent do not exceed 500 mrem/yr to the whole body or 3000 mrem/yr to the skin. The initial setpoints determination was calculated using a conservative radionuclide mix obtained from the GALE code. When sufficient measurable process fission gases are in the effluent, then the actual radionuclide mix will be used to calculate the alarm setpoint.
3.4.2 Setpoint Determination for all Gaseous Release Paths The setpoints for gaseous effluent are based on instantaneous noble gas dose rates. The three release points will be partitioned such that their sum does not exceed 100 percent of the limit.
The setpoints are set at 40 percent for the Reactor Building, 40 percent for the Turbine Building and 20 percent for the Radwaste Building. These percentages could vary at the plant discretion, should the operational conditions warrant such change. However, the combined releases due to variations in the setpoints will notresult in doses which exceed the limit stated in RFO. Both skin dose and whole body setpoints will be calculated and the lower limit will be used. The mechanical vacuum pump discharge radiation monitor setpoint is also established using the methodology of this section.
3.4.2.1        Setpoints Calculations Based on Whole Body Dose Limits The fraction (Trk) of the total gaseous radioactivity in each gaseous effluent release path j for each noble gas radionuclide i will be determined by using the following equation:
7rij=        (dimensionless)                                                                  (6)
MTj where:
Mij            The measured individual concentration of radionuclide i in the gaseous effluent release path j (ItCi/cc).
MTj            The measured total concentration of all noble gases identified in the gaseous effluent release path j (pCi/cc).
35
 
AMENDMENT NO. 35 MARCH 2002 Based on RFO 6.2.2.1, the maximum acceptable release rate of all noble gases in the gaseous effluent release path j is calculated by using the following equation:
QTj =            F. 500          (jACi/sec)
X/Qj    A M.
(Kj)(jij)                                                                (7) i= 1 where:
QTJ      =        The maximum acceptable release rate (IACi/sec) of all noble gases in the gaseous effluent release path j (/tCi/cc).
Fj      =        Fraction of total dose allocated to release path j.
500      =        Whole body dose rate limit of 500 mrem/yr as specified in RFO 6.2.2. 1.a.
X/Qj    =        Maximum normalized diffusion coefficient of effluent release path j at and beyond the site boundary (sec/m 3). Turbine Building and Radwaste Building values are based on average annual ground level values. Main plant vent release values are for mixed mode and may be either short term or average annual value dependent upon type of release.
K        =        The total whole body dose factor due to gamma emission from noble gas nuclide i (mrem/yr per /tCi/m 3 ) (as listed in Table B-1 of Regulatory Guide 1.109, Revision 1).'
7ti      =      As defined in Equation (6).
m        =      Total number of radionuclides in the gaseous effluent.
j        =      Different release pathways.
The total maximum acceptable concentration (CT) of noble gas radionuclides in the gaseous effluent release path j (pCi/cc) will be calculated by using the following equation:
CTj  -    QTj ([tCi/cc)                                                                          (7a)
R.
36
 
AMENDMENT NO. 35 MARCH 2002 where:
CTj              The total allowed concentration of all noble gas radionuclides in the gaseous effluent release path j (pACi/cc).
QTJ              The maximum acceptable release rate (yCi/sec) of all noble gases in the gaseous effluent release path j.
Rj                The effluent release rate (cc/sec) at the point of release.
To determine the maximum acceptable concentration (Cij) of noble gas radionuclide i in the gaseous effluent for each individual noble gas in the gaseous effluent (/tCi/cc), the following equation will be used:
where:
Cij =  Tij CTj  (PCi/cc)                                                                          (7b) mcj  and CTj are as defined in Equations (6) and (7) respectively,    the gaseous effluent monitor alarm setpoint will then be calculated as follows:
m CRj          Z  CijEij(cpm)                                                                    (7c) i=l where:
CRj              Count rate above background (cpm) for gaseous release path j.
Cii        =      The maximum acceptable concentration of noble gas nuclide i in the gaseous effluent release path j (/ACi/cc).
Eii              Detection efficiency of the gaseous effluent monitor j for noble gas i (cpm/ktCi/cc).
37
 
AMENDMENT NO. 35 MARCH 2002 3.4.2.2-      Setpoints Calculations Based on Skin Dose Limits The method for calculating the setpoints to ensure compliance with the skin dose limits specified in RFO 6.2.2.1 .a is similar to the one described for whole body dose limits (Section 3.6.2.1 of this manual), except Equation (7d) will be used instead of Equation (7) for determining maximum acceptable release rate (QTj).
F. 3000 QTJ =            m        3                  (juCi/sec)                                        (7d)
(X/Qj) .*(Li        + 1.1 Mi)(7rij)
I = 1 where:
QTJ              The maximum acceptable release rate of all noble gases in the gaseous effluent release path j in MCi/sec.
X/Qj            The maximum annual normalized diffusion coefficient for release path j at and beyond the site boundary (sec/m 3 ).
Fj      =        Fraction of total allowed dose.
L      =        The skin dose factor due to beta emission for each identified noble gas radionuclide i in mrem/yr per /Ci/m 3 (Li values are listed in Table 3-1).
M      =        The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m3 (Mi values are listed in Table 3-1).
1.1    =        A conversion factor to convert dose in mrad to dose equivalent in mrem.
3000    =        Skin dose rate limit of 3000 mrem/yr as specified in RFO 6.2.2.1.
3.5    10 CFR 50 Dose Calculation The U.S. Nuclear Regulatory Commission's computer program GASPAR II can be used to perform environmental dose analyses for releases of radioactive effluents from Columbia Generating Station into the atmosphere. The analyses estimates radiation dose to individuals and population groups from inhalation, ingestion (terrestrial foods), and external exposure (ground and plume) pathways. The calculated doses provide information for determining compliance with Appendix I of 10 CFR Part 50.
38
 
AMENDMENT NO. 35 MARCH 2002 Both the ODCM equations and the NRC GASPAR II computer program for estimating the highest dose to any organ for a particular age group provides conservatism in calculating maximum organ doses. This conservatism is recognized and is intentional.
Determination of doses due to short-term releases can use the annual average relative concentration (long-term) if it can be demonstrated that past short-term releases were sufficiently random in both time of day and duration (e.g., the short-term release periods were not dependent solely on atmospheric conditions or time of day) to be represented by the annual average dispersion conditions.
The requirements pertaining to 10 CFR 50 dose limits are specified in RFO 6.2.2.2 and 6.2.2.3.
3.5.1 Noble Gases (RFO 6.2.2.2)
RFO 6.2.2.2 deals with the air dose from noble gases and requires that the air dose at and beyond the site boundary due to noble gases released in gaseous effluents shall be limited to the following:
: a.      During any calendar quarter, for gamma radiation:
3.17 x 10-8
* M[(X/Q)gQig    +  (X/q)gqig  +  (X/Q)mQim + (X/q)mqim] *5 mrad                (8)
During any calendar quarter, for beta radiation:
3.17 x 10-8 E Ni [(X/Q) Qjg + (X/q)gqig      +  (X/0)mQi, + (X/q)mqj.m] *10 mrad            (9)
: b. During any calendar year, for gamma radiation:
3.17 x 10-8 E Mi  [(X/Q)gQig  +  (X/q)gqig  + (X/Q)nQim + (X/q)mqim    *10 mrad          (10)
During any calendar year, for beta radiation:
3.17 x 108  E Ni [(X/Q)gQig + (X/q)qqig + (X-Q)inQim + (X/q)mqimj *20 mrad                (11) where:
Mi              The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/yr per /Ci/m 3 (Mi values are listed in Table 3-1).
N.              The air dose factor due to beta emissions for each identified noble gas radionuclide, in mrad/yr per /,Ci/m3 (N1 values are listed in Table 3-1).
39
 
AMENDMENT NO. 35 MARCH 2002 (X/Q)g =  For ground level release points. The highest calculated annual average relative concentration for area at and beyond the site area boundary for long-term releases (greater than 500 hr/yr). (sec/m 3 )
(X/q)g  =  For ground level release points. The relative concentration for areas at and beyond the site area boundary for short-term releases (equal to or less than 500 hr/yr). (sec/m 3 )
(X/Q)m=  For partially elevated release points. The highest calculated annual average relative concentration for areas at and beyond the site boundary for long-term releases (greater than 500 hr/yr). (sec/m 3 )
(X/q)m  =  For partially elevated release points. The relative concentration for areas at and beyond the site boundary for short-term releases (equal to or less than 500 hr/yr). (sec/m 3) qi      = The average release of noble gas radionuclides in gaseous effluents, i, for short-term releases (equal to or less than 500 hr/yr) from the main plant release point, in j.Ci. Releases shall be cumulative over the calendar quarter or year, as appropriate.
qig    = The average release of noble gas radionuclides in gaseous effluents, i, for short-term releases (equal to or less than 500 hr/yr) from Radwaste and Turbine Building, in ItCi. Releases shall be cumulative over the calendar quarter or year, as appropriate.
Qim    = The average release of noble gas radionuclides in gaseous releases, i, for long-term releases (greater than 500 hr/yr) from the main plant release point, in gCi. Release shall be cumulative over the calendar quarter or year, as appropriate.
Qig    = The average release of noble gas radionuclides in gaseous effluents, i, for long-term releases (greater than 500 hr/yr) from Radwaste and Turbine Building, in gCi. Releases shall be cumulative over the calendar quarter or year, as appropriate.
3.17 x 108    = The inverse of the number of seconds in a year (yr/sec).
40
 
AMENDMENT NO. 35 MARCH 2002 3.5.2 Radioiodines, Tritium and Particulates (RFO 6.2.2.3)
RFO 6.2.2.3 deals with radioiodines, tritium, and radioactive materials in particulate form, and requires that the dose to an individual from radioiodines, tritium and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents released to unrestricted areas shall be limited to the following:
: a. During any calendar quarter:
3.17 x 10-1 E Ri I WinQim + Wlqim + WgQig + wgqjg  ] *7.5 mrem                            (12)
: b. During any calendar year:
3.17 x 10-8 E Ri[ WmQim + wnlqim + WgQig + wgqg ] *15 mrem                              (13) i where:
Qim, Qig              The releases of radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases in gaseous effluents, i, for long-term releases greater than 500 hr/yr, in /LCi. Releases shall be cumulative over the calendar quarter or year, as appropriate (m is for mixed mode releases, g is for ground level releases).
qim, qig              The releases of radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases in gaseous effluents, i, for short-term releases equal to or less than 500 hr/yr, in pCi. Releases shall be cumulative over the calendar quarter or year as appropriate (m is for mixed mode releases, g is for ground level releases).
Wm,Wg                  The dispersion parameter for estimating the dose to an individual at the controlling location for long-term (greater than 500 hr) releases (m is for mixed mode releases, g is for ground level releases).
Wm              (X/Q)m for the inhalation pathway and all H-3 doses in sec/m 3 .
Wg              (D/Q)g for the food and ground plane pathways in meters--.
41
 
AMENDMENT NO. 38 APRIL 2003 Wm,Wg                  The dispersion parameter for estimating the dose to an individual at the controlling location for short-term (less than 500 hr) releases (m is for mixed mode releases, g is for ground level releases).
Wm              (X/q).. for the inhalation pathway and all H-3 doses in sec/m 3 .
Wg              (D/q)g for the food and ground plane pathways in meters 2 .
3.17 x 10-8            The inverse of the number of seconds in a year (yr/sec)..
R      =      The dose factor for each identified radionuclide, i, in m2(mrem/yr per ttCi/sec) or mrem/yr per .Ci/m 3 .
3.5.2.1          Dose Parameter for Radionuclide i (R)
The Ri values used in Equations (12) and (13) of this section are calculated separately for each of the following potential exposure pathways:
* Inhalation
* Ground plane contamination
* Grass-cow/goat-milk pathway 0        Grass-cow-meat pathway
*.      Vegetation pathway Monthly dose assessments for gaseous effluent will be done for all age groups.
Calculation of R.I1 (Inhalation Pathway Factor) 3 RI (Inhalation) = K A (BR)a (DFAi)a (mrem/yr per pCi/m  )                                    (14) where:
R                  The inhalation pathway factor (mrem/yr per ACi/m 3 ).
KA                  A conversion constant, 106 pCi/4Ci.
LDCN-03-007 42
 
AMENDMENT NO. 38
                                                                                        -APRIL 2003 (BR)a          =        The breathing rate of the receptor of age group (a) in meter3 /yr.
(Infant = 1400, child = 3,700, teen            8,000, adult = 8,000.
From P.32 NUREG-0133).
(DFAi)a                    The maximum organ inhalation dose factor for receptor of age group a for the ith radionuclide (mrem/pCi). The whole body is considered as an organ in the selection of (DFAi)a. (DFAi)a values are listed in Tables E-7 through E-10 of Regulatory Guide 1.109 manual, Revision 1 and NUREG/CR-4013.
Values of RIare i    listed in Table 3-5.
Calculation of R      (Ground Plane Pathway Factor)
RG (Ground Plane)      K AK B (SF)(DFGi) (1 - e -;Lit)/a (m 2 x mrem/yr per juCi/sec)              (15) where:
R G      =        Ground plane pathway factor (m' x mrem/yr per jCi/sec).
Ka        =        A conversion constant, (106 pCi/1tCi).
K8        =        A conversion constant, (8760 hr/yr).
              &#xfd;.        =        The decay constant for the ith radionuclide (sec').
t        =        Exposure time, 6.31 x 10' sec (20 years).
DFGj      =        The ground plane dose conversion factor for the ith radionuclide, as listed in Table E-6 of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013 (mrem/hr per pCi/m 2).
SF        =        Shielding Factor (dimensionless)--0.7 if building is present, as suggested in Table E-15 of Regulatory Guide 1.109, Revision 1.
The values of RG are listed in Table 3-5 of this manual.
LDCN-03-007 43
 
AMENDMENT NO. 35 MARCH 2002 Calculation of RC1 (Grass-Cow/Goat-Milk Pathway Factor)
R.C (Grass-Cow/Goat-Milk Factor)      =
K QF(Uap) i+
Ffpf)
Fm(r)(DFLi)a
                                              . + (1 -ff)e -Xih 1 j -                          ((16)
(m2 x mrem/yr per 1tCi/sec) where:
KA              A conversion constant    106 pCi//Ci.
QF        =    The cow/goat consumption rate, in kg/day (wet weight).
U          =    The receptor's milk consumption rate for age a, in liters/yr.
Yp        =    The agricultural productivity by unit area of pasture feed grass, in kg/m 2 .
Y          =    The agricultural productivity by unit area of stored feed, in kg/m 2 .
Fm              The stable element transfer coefficients, in days/liter.
r          =    Fraction of deposited activity retained on feed grass.
(DFLj)            =    The maximum organ ingestion dose factor for the ith radionuclide for the receptor in age group a, in mrem/pCi (Tables E- 11 to E- 14 of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013).
S
                        =  The decay constant for the ith radionuclide, in sec-.
X          =  The decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 x 10' sec-' (corresponding to a 14-day half-life).
tf        =    The transport time from pasture to animal, to milk, to receptor, in sec.
t=  The transport time from pasture, to harvest, to animal, to milk, to receptor, in sec.
fp          =  Fraction of the year that the cow/goat is on pasture (dimensionless).
44
 
AMENDMENT NO. 35 MARCH 2002 f=      Fraction of the cow/goat feed that is pasture grass while the cow is on pasture (dimensionless).
C NOTE: For radioiodines, multiply R value by 0.5 to account for the fraction of elemental iodine available for deposition.
C The input parameters teed for calculating R are listed in Table 3-6. The individual pathway dose parameters -for R iare tabulated in Tables 3-5a through 3-5d.
For Tritium:
C In calculating R pertaining to tritium in milk, the airborne concentration rather than the deposition will be used:
RTC(Grass-Cow/Goat-Milk Factor)          =
K AK CFmQFUap (DFLi)a [0.75(0.5/H)] (mrem/yr per                /ACi/m 3)                    (17) where:
KA          =      A conversion constant,    106 pCi/PtCi.
Kc        =      A conversion constant, 10' gm/kg.
H          =      Absolute humidity of the atmosphere, in gm/m 3 .
0.75                The fraction of total feed that is water.
            .0.5        =      The ratio of the specific activity of the feed grass water to the atmospheric water.
Calculation of R M M (Grass-Cow-Meat Pathway Factor)
M RM(Grass-Cow-Meat          Factor)=
1~t A  _______                          (I1-fpf,)e  ih        it KA    QF(Uap)  Ff(r)(DFLi)a    pf  +                    e                        (18) i +
(mremP                                  YS (m2' x mrem/yr per jCi/sec) 45
 
AMENDMENT NO. 35 MARCH 2002 where:
K        =      A conversion constant, 106pCi/4LCi.
Ff      =      The stable element transfer coefficients, in days/kg.
Uap      =      The receptor's meat consumption rate for age a, in kg/yr.
tf      =      The transport time from pasture to receptor, in sec.
th      =      The transport time from crop field to receptor, in sec.
All other parameters are as defined in Equation 16.
NOTE: For radioiodines, multiply R M value by 0.5 to account for the fraction of elemental iodine available for deposition.
                      ~M The input parameters used for calculation R (18) are listed in Table 3-7. The individual (M1 pathway dose parameters for R i are tabulated in Tables 3-5a through 3-5d.
For Tritium:
M In calculating the R . for tritium in meat, the airborne concentration is used rather than the T                                              M deposition rate. The following equation is used to calculate the R T values for tritium:
M R T (Grass-Cow-Meat Pathway) =
KAKc      FIQFUap (DFLi)a  j  0.75(0.5/H) ] (mrem/yr per    1ACi/m 3)                  (19)
M Where the terms are as defined in Equations (16) through (18), R      . values for tritium pertaining to the infant age group is zero since there is no meat consumption by this age group.
46
 
AMENDMENT NO. 35 MARCH 2002 Calculation of RV (Vegetation Pathway Factor)
V RV (Vegetation Pathway Factor)      =
Yj(Xir    ") (DZa          aL fLe -'iL        g)1(0
                                                              + Ua fge -&#xfd;ith K
A        (r)    (
(m2 x mrem/yr per jLCi/sec) where:
KA      =      A conversion constant, 106pCi/JLCi.
UL        =      The consumption rate of fresh leafy vegetation by the receptor in age group a, in kg/yr.
US                The consumption rate of stored vegetation by the receptor in age a
group a, in kg/yr.
fL              The fraction of the annual intake of fresh leafy vegetation grown locally.
fg        -    The fraction of the annual intake of stored vegetation grown locally.
tL        =      The average time between harvest of leafy vegetation and its consumption, in seconds.
th              The average time between harvest of stored vegetation and its consumption, in seconds.
Yv              The vegetation area density, in kg/m 2 .
NOTE: For radioiodines, multiply R V          i value by 0.5 to account for the fraction of elemental iodine available for deposition.
All other items are as defined in Equations (16) through (18).
The input parameters for calculation R V i are listed in Table 3-8. The individual pathway dose parameters for R V are tabulated in Tables 3-5a through 3-5d.
47
 
AMENDMENT NO. 35 MARCH 2002 For Tritium:
In calculating the R T for tritium, the concentration of tritium in vegetation is based on airborne concentration rather than the deposition rate. The following equation is used to calculate R V T for tritium:
R T (Vegetation Pathway Factor)    =
Kc [(ULfL
                        + Uasg) (DFLi)a] [0.75(0.5/H)] (mrem/yr per 1 tCi/m 3)                (21)
Where all terms have been defined above and in Equations (16) through (18), the R T value for tritium is zero for the infant age group due to zero vegetation consumption rate by that age *C group. The input parameters needed for solving Equations (20) and (21) are listed in Table 3-8.
3.5.3 Annual Doses At Special Locations The Radioactive Effluent Release Report submitted within 60 days after January 1 of each year shall include an assessment of the radiation doses from radioactive gaseous effluents to MEMBERS OF THE PUBLIC, due to their activities inside the site boundary during the report period.
Annual doses within the site boundary have been determined for several locations using the NRC GASPARII computer code and source term data from Table 11.3-7 of the FSAR. These values are listed in Tables 3-16 and 3-17. Annual doses to the maximum exposed MEMBER OF THE PUBLIC within the SITE BOUNDARY shall be determined.
48
 
Table 3-1 t"
0 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*
0b Whole Body                                              Gamma Air            Beta Air I
Dose Factor                    Skin Dose Factor          Dose Factor          Dose Factor Kp                            Le e                      Ma                  Na pr i Radionuclide      (mrem/yr per *Ci/m3)          (mrem/yr per *Ci/m3)      (mrad/yr per ,Ci/m3) (mrad/yr per *Ci/m3)
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                      7.11E+02                  3.36E+03            7.39E+02 Xe-135              1.81E+03                      1.86E+03                  1.92E+03            2.46E+03 Xe-137              1.42E+03                      1.22E+04                  1.51E+03            1.27E+04            ~rx1 r-Z 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
    *The listed dose factors are for radionuclides that may be detected in gaseous effluents.                                  z p
00 The values listed above were taken from Table B-1 of NRC Regulatory Guide 1.109, Revision 1.
The values were multiplied by 106 to convert picocuries-' to microcuries-'.
 
AMENDMENT NO. 47 AUGUST 2008 TABLE 3-2 DISTANCES (MILES) TO TYPICAL CONTROLLING LOCATIONS AS MEASURED FROM CENTER OF COLUMBIA GENERATING STATION CONTAINMENT BUILDING Distance in Location*                Sector(s)    miles (meters)          Dose Pathways Site Boundary Reactor Bldg**                        NE              1.212    Air (1950)
Turbine & Radwaste Bldg ***            S              1.212    Air (1950)
Resident                                  NE              4.47      Air, ground, and inhalation (7194)
Resident                                ENE              4.01      Air, ground, meat, and (6453)      inhalation Resident                                  E              4.59      Air, ground, and inhalation (7387)
Resident                                ESE              4.24      Air, ground, and inhalation (6824)
Resident                                  SE              4.5      Air, ground, and inhalation (7242)
* Locations and pathways are based on the 2003 Land Use Census and analysis of joint frequency distribution tables developed from 1997 through 2000 plus 2002 meteorological data. The wind speed and direction data are from the 10 meter elevation of the meteorological tower.
**      highest mixed-level release mode dispersion parameters
***    highest ground-level release mode dispersion parameters LDCN-07-033 50
 
AMENDMENT NO. 47 AUGUST 2008 TABLE 3-3 LONG-TERM AVERAGE DISPERSION (X/Q)
AND DEPOSITION (D/0) VALUES FOR TYPICAL LOCATIONS X/O*        X/O*      X/O*
2.3 Day      8 Day No Decay      Decay      Decay No          No Location              Release Mode      Sector    Distance  Depletion    Depletion  Depleted    D/Q*2 miles      sec/m 3      sec/m 3    sec/mr3      m Site Boundary Reactor Bldg**                Mixed Mode          NE        1.212      3.60E-07    3.60E-07  3.40E-07  4.OOE-09 Ground Mode                              7.80E-06    7.70E-06  6.70E-06    1.30E-08 Turbine/Radwaste Bldg***      Mixed Mode          S        1.212    3.20E-07    3.20E-07  3.OOE-07  3.40E-09 Ground Mode                              1;50E-05    1.50E-05    1.30E-05  1.40E-08 Resident                        Mixed Mode          NE        4.47      8.20E-08    8.1OE-08  7.70E-08  3.30E-10 Ground Mode                              7.30E-07    7. 1OE-07  5.50E-07  6. IE-10 Resident                      Mixed Mode          ENE        4.01      8.60E-07    8.20E-07    7.90E-07  4.60E-10 Ground Mode                              5.40E-07    5.20E-07    4.1OE-07  4.50E-10 Resident                      Mixed Mode            E        4.59      7.70E-07    7.30E-07    7.OOE-07  2.60E-10 Ground Mode                              4.90E-07    4.70E-07    3.70E-07  2.50E-10 Resident                      Mixed Mode          ESE        4.24      7.40E-07    7. 1OE-07  6.40E-07  3.OOE-10 Ground Mode                              4.90E-07    4.70E-07    3.70E-07  2.80E-10 Resident                      Mixed Mode          SE        4.50      1.90E-07    1.80E-07    1.80E-07  2.70E-10 Ground Mode                              4.40E-07    4.20E-07    3.30E-07  2.90E-10
* Dispersion and deposition parameters derived from the average of the following years of meteorological data:
1998, 1999, 2000, 2002, 2003.
**      Location with the highest reactor building dispersion and deposition parameters.
* Location with the highest turbine and radwaste building dispersion and deposition parameters.
LDCN-07-033 51
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-4 DOSE RATE PARAMETERS IMPLEMENTATION OF 10 CFR 20, AIRBORNE RELEASES Child Dose Factor*        p.I I
DFAj          DFGj    Inhalation mrem/hr    mrem/yr Nuclide      X4sec-1    mrem/pCi        pCi/m2    uCi/m3 H-3          1.8E-09      1.7E-07          0.0    6.3E+02 Na-24        1.3E-05    '  4,4E-06        2.9E-08    1.6E+04 Cr-51        2.9E-07      4.6E-06        2.6E-10    1.7E+04 Mn-54        2.6E-08      4.3E-04        6.8E-09  1.6E+06 Mn-56        7.5E-05      3.3E-05        1.3E-08  1.2E+05 Fe-55        8.5E-09      3.OE-05          0.0    1.IE+05 Fe-59        1.8E-07      3,4E-04        9.4E-09  1.3E+06 Co-58        1.1E-07      3.OE-04        8,2E-09  1.1E+06 Co-60        4.2E-09      1.9E-03        2.OE-08  7.OE+06 Cu-64        1.5E-05      9.9E-06        1.7E-09  3.7E +04 Zn-65        3.3E-08      2.7E-04        4.6E-09  1.OE+06 Zn-69m        1.4E-05      2.7E-05        3.4E-09  1.OE+05 As-76        7.3E-06      1.9E-05        1.7E-07  7.OE+04 Br-82        5.5E-06      5.7E-06        2.2E-08  2.1E+04 Sr-89        1.5E-07      5.8E-04        6.5E-13  2.2E+06 Sr-90        7.9E-10      1.OE-02      2.6E-12** 3.7E+07 Zr-95        1.2E-07      6.OE-04        5.8E-09  2.2E+06 Nb-95        2.3E-07        1.7E-04      6.OE-09  6.3E+05 Zr-97        1.1E-05      9.5E-05        6.4E-09  3.5E+05 Nb-97        1.6E-04      7.5E-06        5.4E-09  2.8E+04 Mo-99        2.9E-06      3.7E-05        2.2E-09  1.4E+05 Tc-99m      3.2E-05        1.3E-06        1.1E-09  4.8E+03 Ru- 106      2.2E-08      3.9E-03        1.8E-09  1.4E+07 Ag-lr0m      3.2E-08        1.5E-03      2.1E-08  5.6E+06 Sb- 124      1.3E-07      8.8E-04        1.5E-08  3.3E+06 Sb-125      7.9E-09      6.3E-04        3.5E-09  2.3E+06 Sb- 126      6.5E-07      2.9E-04        1.OE-08  1.1E+06 Sb-127      2.1E-06      6.2E-05        6.6E-09  2.3E+05 Te- 127      2.1E-05      1.5E-05        1.1E-11  5.6E+04 Te-131m      6.4E-06      8.3E-05        9.9E-09  3.1E+05 1-131        1.OE-06      4.4E-03        3.4E-09  1.6E+07 1-132        8.4E-05      5.2E-05        2.OE-08  1.9E+05 1-133        9.2E-06      1.OE-03        4.5E-09  3.7E+06 1-135        2.9E-05      2.1E-04        1.4E-08  7.8E+05 Cs- 134      1. 1E-08      2.7E-04        1.4E-08  1.OE+06 Cs-137      7.3E-10      2.5E-04        4.9E-09  9.3E+05 52
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-4 (continued)
DOSE RATE PARAMETERS IMPLEMENTATION OF 10 CFR 20, AIRBORNE RELEASES Child Dose Factor*            p.I DFAi          DFGj        Inhalation mrem/hr        mrem/yr Nuclide          X(sec')      mrem/pCi          pCi/m2        /&Ci/m 3 Cs-138            3.6E-04      2.3E'07        2.4E-08      8.5E+02 Ba-140            6.3E-07      4.7E-04        2.4E-09        1.7E+06 La- 140          4.8E-06      6.1E-05        1.7E-08      2.3E+05 Ce-141            2.4E-07      1.5E-04        6.2E-10      5.6E+05 Ce-144            2.8E-08      3.2E-03        3.7E-10        1.2E+07 Nd-147            7.2E-07      8.9E-05        1.2E-09      3.3E+05 Hf- 179m,        3.7E-02      2.OE-05      NO DATA        7.4E+04 Hf-181            1.8E-07      6.OE-05        1.2E-08      2.2E+05 W-185            1.1E-07      1.9E-04          0.0        7.OE+05 Np-239          3.4E-06        1.7E-05        9.5E-10      6.4E+04
* Maximum Organ
**No data is listed for Sr-90 in Table E-6 of Regulatory Guide 1.109, Revision 1. Y-90 values were used for dose conversion factor Sr-90.
53
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-5a DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: ADULT                            ORGAN OF
 
==REFERENCE:==
MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES Radionuclide      Inhalation          Ground  Plane        Cow  Milk      Goat Milk    Animal Meat    Vegtables (mrem/yr              (M2*mrem/yr            (M2*mrem/yr      (M2*mrem/yr  (M2*mrem/yr    (M *rnrem/yr per jACi/M 3 )        per pCi/sec)          per #iCi/sec)    per pCi/sec)  per j4Ci/sec)  per pCi/sec)
H 3              7.2E+02              0.OE-01                5.8E+02          1.2E+03      2.4E+02        1.6E+03 NA 24            I.OE+04              1.2E+07                1.2E+06        2.2E+05      7.2E-04        1. IE+05 CR 51            1.4E+04              4.7E+06                3.3E+06          5.9E+05      8.2E +05      2.3E+07 MN 54            1.4E+06              1.4E+09                1.4E+07        2.1E+06        1.5E+07      9.4E+08 MN 56            2.OE+04              9.OE+05                6.2E-02          1.1E-02      0.0E-01        2.OE+02 FE 55            7.2E+04              0.0E-01                1.4E+07        2.2E +06      1.6E+08        1.9E+08 FE 59            1.0E+06              2.7E +08                1.IE+08        2.OE+07      9.8E +08      1.5E +09 CO 58            9.3E+05              3.8E+08                4.7E+07          7.6E+06      1.8E +08      8.OE+08 CO 60            6.OE+06              2.3E+ 10.                1.7E+08        2.5E+07      8.OE+08      2.9E+09 CU 64            4.9E+04              6.1E+05                I.OE+06          1.7E+05      I.IE-05      3,3E+05 ZN 65            8.6E+05              7.5E+08                2.7E +09        4.OE+08        7.OE+08        1.3E+09 ZN 69M            1.4E+05              1.3E+06                1.3E+07        2.4E+06        1.2E-03        1.4E+06 AS 76            1.5E+05              3.8E,+06                2.1E+07        3.8E+06        2.9E+01      8.OE+06 BR 82            1.4E+04              2. IE +07              1.9E+07        3.4E+06        7.OE+02      7.7E+05 SR 89            1.4E+06              2.2E+04                6.9E+08        2.OE +09      1.4E+08        1.5E+10 SR 90            2.9E +07            6.7E+06                3.4E+ 10        8.3E+10        8.9E +09      7.4E+11 ZR 95            1.8E+06              2.5E+08                4.6E+05        7.6E+04      9.2E+08        1.6E+09 NB 95            5.lE+05              1.4E+08                1.3E+08        2.2E+07      3.6E +09      8.4E +08 ZR 97            5.2E+05              3.OE+06                1.4E+04        2.4E+03      6.4E-01        8.8E+06 NB 97            2.4E +03            1.8E+05                1.6E-09        2.9E-10      O0.OE-Ot      8.IE-04 MO 99            2.5E+05              4.0E+06                2.9E +07        5.2E +06      1.2E +05      9.3E+06 TC 99M            4.2E1+03            1.8E+05                2.8E +03        5.E+02        3.6E-18        2.2E +03 RU106            9.44E +06            4.2E+08                7.3E+05        1. IE+05      1.OE+Il      1.2E+10 AG 110M          4.6E +06              3.5E+09                1.2E+ 10        1.8E+09      1.4E+09        4.4E+09 SB124            2.5E+06              6.OE+08                3.5E+08          5.8E+07      2.7E+08        4.OE +09 SB125            1.7E1+06            2.4E1+09                1.3E1+08        1.8E+07      1.2E+08        1.4E+09 SB126            7.7E +05              8.4E+07                2.2E+08          4.OE+07      7.6E+07.      1.6E+09 SB127            3.OE+05              1.7E+07                52E+07          9.3E+06        1.9E+06        t.2E+08 TE127            5.7E+04              3.OE+03                2,6E+04        4.7E+03        8.4E-09      2.OE+05 TEI3[M          5.6E+05              8.OE+06                8.9E+06          1.6E+06      1.IE +04      2,OE+07 1 131            1.2E+07            8.6E+06                3.4E + 10      6. IE+ 10      1.2E1+09      4.4E+ 10 1 132            1.IE+05            6.2E+05                3.9E +00        6,9E+00        0.OE-01        1.IE +03 1133            2.2E+06              1.2E+06                2.5E+08        4.5E +08      2.4E+01        I.IE+08 1135            4.5E+05              1.3E+06                5.5E+05        9.8E+05        1.7E-15        I.4E+06 CS 134          8,5E+05              6.9E+09                7.4E+09        2.7E+ 10      8.6E+08        I.OE + 10 CS136            1,5E+05              1.5E+08              5.0E-+08        2.2E+09      2.3E+07        4.6E+08 CS137            6.2E+05              1.3E+10              6.OE+09          2.1E+10      7. 1E+08      8.6E +09 CS 138          6.2E+02              3.6E+05                1.OE-23        4.6E-23      O.OE-00        3.OE-11 BA 140          1.3E+06              2.IE+07                2.7E+07          4.8E+06      2.8E+07        7.3E+08 LA140            4.6E +05            1.9E1+07              8.4E+04          [,5E+04      7,0E +02      3.3E+07 CE141            3.6E+05              1.4E+07                5.8E +06        I.OE+06        1.7E +07      9.3E+08 CE144.          7.8E+06              7.OE+07                6.4E+07          9.6E+06      2.6E +08      1.IE+ 10 ND147            2.2E+05              8.5E+06                2.5E+05          4.6E +04      1.9E+07      5. 1E + 08 HF 179M          1.6E+05              0.OE-01                0.OE-01          0.OE-01      0.OE-01        0.OE-0O HF181            4.8E +05            2. IE+08              5.5E+05          9.3E+04      1.2E+10        1.8E+09 W 185            4.5E+05              1,8E+04                2.4E +07        3.9E+06      1.9E+07        8.4E +08 NP239            1.2E+05              1.7E+06                3.7E+04          6.7E+03      2.6E+03        1.6E+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.
The PARTS subroutine of GASPAR 11was used to produce this table.
54
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-5b DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: TEEN                          ORGAN OF
 
==REFERENCE:==
MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation          Ground Plane            Cow Milk      Goat  Milk      Animal  Meat  Vegetables (mrem/yr 3        (M2*mrem/yr            (M2 *mrem/yr    (M2 *mrem/yr    (M2*mrem/yr    (M-*nmrem/yr Radionuclide        per ,Ci/M )        per pCi/sec)            per uiCi/sec)  per ,pCi/sec)  per /ICi/sec)  per aCi/sec)
H 3                7.3E+02            0.OE-0I                7.5E+02        1.5E+03        1.5E+02        1.9E+03 NA 24                1.4E+04            1.2E+07                2. IE+06        3.9E+-05        5,8E-04        L.OE+05 CR 51                2. IE+04          4.7E+06                3.9E+06        6.8E+05        4.4E+05        2.5E+07 MN 54                2.OE+06            1.4E+09                1.6E+07        2.3E+06        7.8E+06        9.6E +08 MN 56                5.7E+04            9.OE+05                2.3E-01        4.1E-02        0.OE-00        3.7E+02 FE 55                1.2E+05            O.OE-01                2.4E +07        3.8E+06        1.3E+08        3.0E+08 FE 59                t .5E+06          2.7E+08                1.3E+08        2.5E+07        5.5E+08          1.7E+09 CO 58                1.3E+06            3.8E+08                5.3E+07        8.7E+06        9.4E +07        8.3E+08 CO 60                8.7E+06            2.3E+ 10                2. IE+08      3.OE+07        4.3E +08        3. 1E +09 CU 64                6. 1E+04          6. 1E+05                1.6E+06        2.7E+05        8.OE-06        2.7E+05 ZN 65                1.2E+06            7.5E+08                4.5E +09      6.7E+08        5.4E +08        2.0E+09 ZN 69M              1.7E+05            1.3E+06                2. tE+07      3.8E+06        9.1 E-04        1.tE+06 AS 76                1.5E+05            3.8E+06                2.7E+07        4.9E+06          1.7E+01        5.3E+06 BR 82                1.8E+04            2. IE+07                2.8E +07      5. IE +06      4.9E+02        6. 1E +05 SR 89              2.4E+06            2.2E+04                1.3E+09        3.7E +09        1.2E+08        2.4E + 10 SR 90              3.3E+07            6.7E+06                5. 1E+ 10      1.3E+1        6.2E+09        1.0E+ 12 ZR 95              2.7E+06            2.5E+08                5.8E+05        9.5E+04        5.3E+08        1.8E+09 NB 95              7.5E+05            1.4E+08                1.6E+08        2.7E+07        2.OE+09        9. IE +08 ZR 97              6.3E+05            3.OE+06                2. 1E +04      3.8E+03        4.6E-01        7.OE+06 NB 97              3.9E+03            1.8E+05                1.9E-08        3.3E-09        0.OE-01        4.8E-03 MO 99              2.7E +05            4.OE+06                5. 1E +07      9.2E +06        9.4E+04        1.1 E +07 TC 99M              6. IE+03            1.8E+05                5.3E+03        9.5E+02        3.2E- 18        2. IE+03 RU106                1.6E+07          4.2E+08                9.9E+05        1.5E +05        6.2E+ 10        1.5E+ 10 AG IOM              6.8E1+06          3.5E+09                  1.4E+ 10      2. IE+09        7.6E+08        4.6E+09 SB124              3.8E+06            6.OE+08                4.5E+08        7.3E+07        1.6E+08        4.6E+09 SB 125              2.7E+06            2.4E +09                1.6E+08        2.3E+07        6.8E+07          1.6E+09 SB126                1.2E+06            8.4E+07                2.8E +08        5. tE+07      4.5E+07          1.8E+09 SB127                3.2E+05            1.7E+07                6.9E +07        1.2E+07        1.2E+06        1.2E +08 TE127                8. IE+04          3.OE+03                4.8E +04        8.6E+03        7.OE-09          1.8E+05 TE131M              6.2E+05            8.OE+06                  1.3E+07      2.3E+06        7.4E +03        1.5E+07 1 131              1.5E+07            8.6E +06                5.4E+ 10      9.7E+ 10        9.OE+08        6. lE+ 10 1 132                1.5E+05            6,2E+05                6,4E+00        1.2E+01        0.OE-00        9.3E+02 1133                2.9E+06            1.2E+06                4.2E+08        7.5E+08          1.8E+01        9.6E+07 1135                6.2E+05            1.3E+06                9.3E+05        1.7E+06        1.3E-15        1.2E+06 CS 134              1. I E+06          6.9E+09                1.3E+I0        4.6E + 10      6.8E+08          1.6E+ 10 CS136                1.9E+05            1.5E+08                8.4E +08      3.8E+09          1.8E+07        7.OE+08 CS137                8.5E+05            1.3E+ 10                1.IE+10        3.8E+ 10        5.7E+08          t.4E+ tO CS 138              8.6E+02            3.6E +05                1.8E-23        8.1 E-23        O.OE-00        2.7E-I1 BA 140              2.OE+06            2.1E+07                3.6E+07        6.4E+06        1.8E+07        8.8E+08 LA 140              4.9E+05            1.9E+07                l.IE+05        2. 1E +04      4.4E+02        2.4E+07 CE141                6. IE +05          1.4E+07                7.9E +06      I.4E+06        I.OE+07        1. IE +09 CE 144              1.3E+07            7.OE+07                8.8E+07        1.3E+07        1.6E+08        1.3E+ 1O ND147                3,7E+05            8.5E+06                3.5E+05        6.2E+04        1.2E+07        6. 1E +08 HFI79M              7. IE +04          0.OE-01                0.OE-01        0.OE-01        0.OE-01        0.OE-01 HF181                4.8E+05            2. IE+08                7. 1E +05      1.2E+05        7.OE+09        2. IE+09 W 185                7.7E+05            1.8E+04                313E+07        5.4E +06        1.2E+07        I.E+09 NP239                1.3E+05            1.7E+06                5.3E+04        9.6E + 03      1.7E+03        1.4E+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.
The PARTS subroutine of GASPAR 11was used to produce this table.
55
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-5c DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: CHILD                            ORGAN OF
 
==REFERENCE:==
MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation        Ground Plane            Cow Milk      Goat Milk      Animal Meat    Vegetables (mrem/yr 3        (M2*mrem/yr            (M2*rnrem/yr  (M2*mrem/yr    (M2*mrem/yr    (M *mrem/yr Radionuclide        per ACi/M )        per PCi/sec)            per uCi/sec)  per pCi/sec)  per pCi/sec)  per gCi/sec)
H 3                6.4E+02            0.OE-0I                1.2E+03        2.4E +03        1.8E+02      2.9E+03 NA 24                1.6E+04            1.2E+07                4.5E+06        8.OE+05        9.2E-04        1.6E+05 CR 51                1.7E+04            4.7E+06                2.5E+06        4.4E+05        2.2E+05        1.6E+07 MN 54                1.6E+06            1.4E+09                1. IE+07      1.7E+06        4.3E+06        6.9E+08 MN 56                1.2E+05          9.OE+05                8.813-01        1.6E-01        O.OE-00        1.IE +03 FE 55                1.IE+05          O.OE-0I                6. IE +07      9.6E +06        2.5E+08        7.6E+08 FE 59                1.3E+06          2.7E+08                9.5E+07        1.7E+07        3.0E+08        1.2E+09 CO 58                1.IE+06          3.8E+08                3.4E+07        5.6E+06        4.7E+07        5.3E+08 CO 60              7. IE+06          2.3E+ 10                1.4E+08        2.OE+07        2.2E+08        2. IE +09 CU 64              3.7E+04            6. IE+05                1.7E+06        2.9E+05        6.5E-06        2.2E+05 ZN 65                1.0F+06          7.5E+08                6.8E+09        I.OE+09      6.2E+08        3.0E1+09 ZN 69M              l.OE+05            1.3E+06                2.2E+07        4.OE+06        7.2E-04        9.OE+05 AS 76              7.OE+04            3.8E+06                2.2E +07      4.OE+06        I. IE+01      3.3E+06 BR 82              2. IE+04          2.1 E+07                5.8E+07        1.EO+07      7.6E+02        9.5E+05 SR 89                2.2E+06            2.2E+04                3.1 E +09      9.2E+09        2.3E+08        6.OE + 10 SR 90                3.8E+07            6.7E+06                I.OE+ II      2.6E+11        9.8E+09        2. 1E +12 ZR 95                2.2E+06            2.5E+08                4.2E+05        7.OE+04        3.OE+08        1.3E+09 NB 95                6. tE+05            1.4E+08                1.1E+08        1.8E+07        1.OE+09      6.2E +08 ZR 97                3.5E+05            3.OE+06                2. IE+04      3.8E+03        3.5E-01        5.2E+06 NB 97                2.8E+04            1,8E+05                4.2E-07        7.6E-08        0.OE-01        8.2E-02 MO 99                1.3E+05            4.OE+06                8.7E+07        1.6E+07        1.2E+05        1.6E+07 TC 99M              4.8E+03            1.8E+05                7.4E +03      1.3E+03        3.4E- 18      2.2E+03 RU106                1,4E+07            4.2E+08                7.9E +05      1.2E+05        3.8E + 10      1.2E + I0 AGl IOM            5.5E+06            3.5F+09                9.4E+09        1.4E+09        3.8E+08        3.0E+09 SB124              3.2E1+06            6.OE+08                3.3E+08        5.4E+07        8.8E+07        3.3E+09 SB 125              2.3E+06            2.4E+09                1.2E+08        1.7E+07        3.8E+07        1.2E+09 SB 126              1. tE+06          8.4E+07                2.2E+08        4.OE+07        2.7E+07        1.4E+09 SB t27              2.3E+05            t .7E+07              5.5E+07        I.OE+07        7.2E+05        9.2E+07 TE127              5.6E+04            3.OE+03                5.9E +04        1. 1E+04      6.7E-09        1.7E+05 TE131M              3. I E+05          8.OE+06                  1.IE+07      2.1E+06        5.0E+03        9.9E +06 1 131                1.6E+07          8.6E +06                1.IE+ II      1.9E+ II      1.4E+09        1.2E+ II 1 132                1.9E+05          6.2E+05                  1.5E+01      2.7E+01        0.OE-00        1.6E+03 1 133              3.8E +06            1.2E+06              9.9E +08        1.8E+09        3.3E+01        1.7E+08 1 135              7.9E+05            1.3E+06              2.IE+06        3.8E+06        2.3E-15        2.1 E+06 CS 134              I.OE+06          6.9E+09                2.OE+ 10        7.5E + 10      8.3E+08        2.6E + 10 CS136                1.7E+05            1.5E+08                1.3E+09        6.OE +09      2. 1E +07      1. IE +09 CS137              9. IE+05            1.3E+I0                1.9E+ 10      6.8E+ 10      7.9E+08        2.5E+ 10 CS 138              8.4E+02            3.6E+05                3.2E-23          1.4E-22      0.OE-00        3.6E- II BA140              1.7E+06            2. IE+07              5.6E+07          LOE +07      2. IE+07        1.4E+09 LA 140              2.3E+05            1.9E+07              9.5E+04        1.7E+04        2.8E+02        1.6E+07 CE141                5.4E+05            I.4E+07              6.3E+06        l.IE+06        6.4E+06        9.O0+08 CE144                1.2E+07            7.OE+07                7.OE+07        1. IE+07        1.0E+08        I. IE+ 10 ND147                3.3E+05            8.5E+06                2.8E+05        5.OE+04        7.4E+06        4.8E+08 HF 179M              7.4E+04            O.0E-01                0.OE-0O        0.OE-01        0.OE-01        0.OE-0I HF181                2.2E+05            2.1 E+08              5.9E+05        9.9E+04        4.4E+09        18E +09 W 185                6.9E+05            1.8E+04                2.7E+07        4.3E+06        7.3E-+06      8.3E+08 NP239                6.4E+04            1.7E+06                4.6E+04        8.3E +03      I.IE+03        1.0E+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.
The PARTS subroutine of GASPAR II was used to produce this table.
56
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-5d DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: INFANT                            ORGAN OF
 
==REFERENCE:==
MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation        Ground Plane            Cow Milk      Goat Milk      Animal Meat    Vefetables (mrem/yr            (M2*mrren/yr            (M2 *mnrem/yr  (M2*mrem/yr    (M2*nlrem/yr  (M *mrem/yr Radionuclide        per pCi/M")        per mCi/sec)            per uCi/sec)  per pCi/sec)  per pCi/sec)  per ,uCi/sec)
H 3                3.7E+02            O,OE-01                  1.8E+03      3.7E+03        O.OE-01        O.OE-Ol NA 24                1.IE+04            1.2E+07                7.8E+06        1.4E+06      0.0E-01        0.0E-01 CR 51                1.3E+04            4.7E+06                2,2E+06        3.8E+05      O.OE-01        0.OE-01 MN 54                1.OE+06            1.4E+09                2,IE+07        3. IE +06    O.OE-01        0.OE-01 MN 56              7.2E +04            9.OE+05                  1.3E+00      2.4E-01      0.OE-01        O.OE-01 FE 55              8.7E +04          O.OE-01                  7.4E+07        1.2E+07      O.OE-01        O.OE-01 FE 59                t.0E+06          2.7E+08                  1,8E+08      3.4E+07      0.OE-O0        O.OE-0 I CO 58              7.8E+05            3.8E+08                  2.9E+07        4.8E+06      O.OE-01        0.OE-01 CO 60              4.5E+06            2.3E+ 10                  1.2E+08      1.7E+07      O.OE-01        0.OE-01 CU 64                1.5E+04          6. IE+05                1.9E+06        3.2E+05      0.OE-01        O.0E-01 ZN 65              6.5E +05          7.5E+08                  1.2E + 10      1.7E+09      0.OE-01        O.OE-Ot ZN 69M              4.IE+04            1.3E+06                2.4E+07        4.3E+06      0.OE-01        0.OE-O1 AS 76              2.7E +04          3.8E+06                  2.2E+07        4.UE+06      0.OE-01        0.OE-01 BR 82              1.3E+04            2. IE +07                9.8E+07        1.8E+07      0.OE-01        0.OE-01 SR 89              2.OE+06            2.2E+04                  6.0E+09        1.8E+10      0.OE-01        O.OE-01 SR 90              1.6E+07            6.7E+06                  1.2E+II        2.9E+11      0.OE-01        O.OE-01 ZR 95              1.8E+06            2.5E+08                  4.OE+05        6.5E+04      O.OE-01        O.OE-O0 NB 95              4.8E+05            1.4E+08                9.6E+07        1.7E1+07    O.OE-01        0.OE-01 ZR 97                1.4E+05            3.OE+06                  2.2E+04      4.OE+03        O.OE-01        O.OE-0 I NB 97                2.7E+04            1.8E+05                1.IE-06        1.9E-07      0.OE-01        O.OE-01 MO 99                1.3E+05            4.OE+06                  1.6E+08      2.8E+07        0.OE-01        0.0E-01 TC 99M              2.OE+03            1.8E+05                  8.2E+03        1.5E+03      O.OE-Ol        O.OE-Ol RU106                1.2E+07            4.2E+08                  8.OE +05      1.2E+05      0.OE-01        O.OE-01 AG110M              3.7E +06            3.5E+09                8.2E+09        1.2E+09      O.OE-01      0,E-01 SB 124              2.6E +06            6.OE+08                3.1 E +08      5. IE+07      OE-01        O.OE-01 SB 125              1.6E+06            2.4E+09                  1.IE+08.      1.6E+07      0.OE-01      o.OE-01 SB 126              9.6E+05            8.4E +07                2.IE+08        3.7E+07        O.OE-01      O.OE-01 SB127              2.2E+05            1.7E+07                5.5E+07        9.9E +06      O.OE-01        O.0E-01 TE127              2.4E+04            3.OE+03                6.8E+04        1.2E+04      0.OE-0O        0.OE-01 TE131M              2.OE+05            8.OE+06                  1.2E+07      2.1E+06      0.OE-0O        0.OE-01 1 131                1.5E+07            8.6E +06                2.6E+ 11      4.7E+11      0.OE-01        0.OE-0l 1 132                1.7E+05            6.2E+05                3.4E+01        6.1E+01      O.OE-01        O.OE-0O 1 133              3.6E+06            I 2E+06                2.4E+09        4.3E+09      OOE-0 I        0.OE-0 l 1 135              7.OE +05            1.3E+06                4.9E+06        8.9E+06      0.OE-01        0.OE-01 CS 134              7.OE +05            6.9E+09                3.7E + 10      1.4E+ 11    0.OE-01        0.OE-01 CS 136              1.3E+05            1.5E+08                2.8E+09        1.2E+10      0.OE-01        0.OE-01 CS137              6. IE+05            1.3E+ 10                3.6E+ 10      1.3E+II      O.OE-01        0.OE-0I CS138              8.8E+02            3.6E+05                  1.2E-22      5.6E-22      0.OE-01        O.OE-Ot BA 140              1.6E+06          2. 1E +07                1.2E+08      2. IE+07      O.0E-01        O.OE-01 LA 140              1.7E+05            1.9E+07                9.4E+04        1.7E+04      0.OE-01        O.OE-01 CE141              5.2E+05            I.4E+07                6.4E+06        1.I E+06      0.OE-01        O.OE-01 CE144              9.8E+06            7.0E3+07                7. 1E +07      1.IE+07      0.OE-01        0.OE-01 ND147              3.2E +05          8.5E+06                  2.8E+05        5.0E+04      O.OE-01        O.OE-01 HF179M              2.8E+04            0.OE-01                  0.OE-01        0.OE-01      O.0E-01        O.OE-Ol HFI81              8,4E +04          2. 1E+08                5.9E+05        9.9E+04      O.OE-01        O.OE-01 W 185              6.3E+05            1.8E+04                2.7E+07        4.4E+06      0.OE-01        O.OE-01 NP239              6.OE +04            1.7E+06                4.7E+04        8.5E+03      0.OE-01        O.OE-01 NOTE: The Y-90 ground plane dose factor was used for Sr-90.
The PARTS subroutine of GASPAR 11was used to produce this table.
57
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-6 INPUT PARAMETERS FOR CALCULATING          R Parameter                      Value                      Table*
r (dimensionless)                1.0 for radioiodine          E-15 0.2 for particulates          E-15 Fm (days/liter)                Each stable element            E-1 Up (liters/yr) --Infant          330                          E-5
                  --Child        330                          E-5
                  --Teen          400                          E-5
                  --Adult        310                          E-5 (DFLi)a (mrem/pCi)              Each radionuclide            E-11 to E-14 Yp (kg/M2)                      0.7                          E- 15 Y, (kg/m 2)                    2.0                            E-15 tf (seconds)                      1.73 x 105 (2 days)          E-15 th  (seconds)                    7.78 x 10' (90 days)          E-15 QF (kg/day)                      50 for cow                    E-3 6 for goat                    E-3 fs (dimensionless)              1.0                          NUREG-0133 fp (dimensionless)              0.5 for cow                  site specific 0.75 for goat                site specific
*Of Regulatory Guide 1.109, Revision 1 unless stated otherwise.
58
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-7 INPUT PARAMETERS FOR CALCULATING R .
Parameter                      Value              Table*
r (dimensionless)              1.0 for radioiodine  E-15 0.2 for particulates E-15 F,(days/kg)                    Each stable element  E-1 Up (kg/yr) --Infant            0                    E-5
              --Child            41                    E-5
              --Teen            65                    E-5
              --Adult            110                  E-5 (DFLi), (mrem/pCi)              Each radionuclide    E-11 to E-14 YP (kg/m 2 )                    0.7                  E- 15 Y, (kg/m 2 )                    2.0                  E- 15 tf (seconds)                    1.73 x 106 (20 days)  E-15 th (seconds)                    7.78 x  106 (90 days) E-15 QF (kg/day)                    50                    E-3
*Of Regulatory Guide 1.109, Revision 1.
59
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-8 V
INPUT PARAMETERS FOR CALCULATING R Parameter                      Value              Table*
r (dimensionless)                1.0 for radioiodine  E-1 0.2 for particulates  E-1 (DFL )a (mrem/pCi)              Each radionuclide      E-11 to E-14 UL 0                      E-5 a    (kg/yr) --Infant
                --Child          26                    E-5
                --Teen          42                    E-5
                --Adult          64                    E-5 Us 0                      E-5 a  (kg/yr) --Infant
                --Child          520                    E-5
                --Teen          630                    E-5
                --Adult          520                    E-5 fL (dimensionless)              0.42                  Ref 2**
fg (dimensionless)              0.76                  E  tL (seconds)                    8.6 x 104 (1 day)    E-15 th (seconds)                    5.18 x 106 (60 days)  E-15 Y, (kg/M 2)                      2.0                  E-15
*Of Regulatory Guide 1.109, Revision 1.
**Refer to Table 3-14.
60
 
AMENDMENT NO. 44 MAY 2005 TABLE 3-9 INPUT PARAMETERS FOR CALCULATING DOSE SUMMARIES TO THE MAXIMUM INDIVIDUAL AND THE POPULATION WITHIN 50 MILES OF COLUMBIA GENERATING STATION FROM GASEOUS EFFLUENTS Input Parameter                                      Value      Reference*
Distance to Maine (miles)                            3000        Ref 1 Fraction of year leafy vegetables are grown (FV)      0.42        Ref 2, Fraction of year cows are on pasture (FP)            0.5          Ref 2 Fraction of crop from garden (FG)                    0.76        Ref 3 Fraction of daily intake of cows derived from pasture while on pasture (FPF)                                  1.0        Ref 2 Annual average relative humidity (%) (H)              53.8        Ref 4 Annual average temperature (F') (T)                    53.0        Ref 5 Fraction of year goats are on pasture (FGT)            0.75        Ref 2 Fraction of daily intake of goats derived from pasture while on pasture (FPG)                                1.0        Ref 2 Fraction of year beef cattle are on pasture (FB)      0.5        Ref 2 Fraction of daily intake of beef cattle derived from pasture while on pasture (FBF)                        1.0        Ref 2 Population within 50 miles of plant by direction and radii interval in miles                                356,993    Ref 6 Annual 50-mile milk production (liters/yr)            2.8E+08    Refs 2 & 7 Annual 50-mile meat production (kg/yr)                2.3E+07    Refs 2 & 7 Annual 50-mile vegetable production (kg/yr)            3.5E+09    Refs 2 & 7 Source terms                                                      Ref 8 LDCN-04-012 61
 
AMENDMENT NO. 44 MAY 2005 TABLE 3-9 (continued)
Input Parameter                                  Value              Reference*
X/Q values by sector for each distance            See Tables 3-10 and  Ref 9 (recirculation, no decay) (sec/m 3)                3-11 X/Q values by sector for each distance            See Tables 3-10 and  Ref 9 (recirculation, 2.26 days decay, undepleted)      3-11 (sec/m 3)
X/Q values by sector for each distance              See Tables 3-10 and  Ref 9 (recirculation, 8.0 days decay, depleted)          3411 (sec/m 3 )
D/Q values by sector for each distance (1/mr)      See Tables 3-10 and  Ref 9 3-11 Quarterly Input Parameters Month        FV      FP      FG      FPE    H      T    FGT  FPG      FB  FBF Jan-Mar    0.01    0.01    0.76    1.0                  0.33  1.0    0.01 1.0 Apr-Jun    0.5    0.83    0.76    1.0                  1.0  1.0    0.83  1.0 Jul-Sep    1.0    1.0      0.76    1.0                  1.0  1.0    1.0  1.0 Oct-Dec    0.17    0.17    0.76    1.0                  0.67  1.0    0.17  1.0
*References are listed in Table 3-14.
LDCN-04-012 62
 
I-z7 TABLE 3-10*
REACTOR BUILDING STACK X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS I
A) NO DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE        .5-1        1-2          2-3          3-4        4-5        5-10      10-20      20-30    30-40      40-50 S 7.233E-07  2.479E-07    9.489E-08    5.568E-08  3.999E-08  2.375E-08  2.942E-07  2.492E-07  1.680E-07 1,253E-07 SSW  6.151E-07  1.952E-07    6.890E-08    3.959E-08  2.828E-08  1.787E-08  2.911E-07  2.500E-07  1.695E-07 I.269E-07 SW  5.252E-07  1.699E-07    6.213E-08    3.600E-08  2.480E-08  3.197E-08  7.124E-08  4.933E-08 3.231E-08  2.361E-08 WSW    2.380E-07  7.862E-08    2.917E-08    1.730E-08  1.219E-08  1.137E-08  7.239E-08  5.943E-08 4.062E-08  3.060E-08 W  4.040E-08  1.526E-08    6.870E-09    4.921E-09  3.869E-09  2.400E-09  6.163E-08  5.262E-08 3.596E-08  2.708E-08, WNW    1.006E-07  4.327E-08    1.762E-08    1.015E-08  6.869E-09  3.999E-09  4.582E-08  3.813E-08 2.567E-08  1.913E-08 NW  3.535E-07  1.115E-07    4.115E-08    2.387E-08  1.692E-08  1.107E-08  4.969E-08  3.987E-08 2.653E-08  1.962E-08 NNW    8.116E-07  2.537E-07    9.528E-08    5.516E-08  4.251E-08  2.564E-08  7.337E-08  5.778E-08 3.843E-08  2.841E-08 N  9.786E-07  2.693E-07    9.262E-08    5.237E-08  3.635E-08  1.259E-07  1.513E-07  8.039E-08 5.332E-08  3.932E-08 NNE  8.968E-07  2.408E-07    8.399E-08    4.833E-08  3.714E-08  2.717E-07  2.161E-07  1.185E-07 8.013E-08  5.991E-08 NE  7.711 E-07  2.623E-07    1.014E-07    6.451E-08  1.267E-07  3.922E-07  2.312E-07  1.264E-07 8.540E-08  6.382E-08 ENE  6.527E-07  2.596E-07    1.948E-07    6.621E-07  7.164E-07  3.445E-07  1.395E-07  7;490E-08 5.OOIE-08  3.706E-08 E  4.084E-07  1.746E-07    2.109E-07    8.577E-07  8.018E-07  3.851E-07  1.558E-07  8.349E-08                            rn 5.567E-08  4.121E-08 ESE  3.220E-07  1.821E-07  2.632E-07      6.970E-07 6.760E-07  2.577E-07  9.473E-08  6.573E-08 4.325E-08  3.171E-08    0 LA SE  3.789E-07  1.320E-07  6.120E-08      1.293E-07  1.887E-07  9.740E-08  9.029E-08  6.192E-08 4.060E-08  2.970E-08 SSE zp 5.081E-07  1.667E-07  6.492E-08      3.811E-08 2.717E-08  1.665E-08  8.252E-08  6.906E-08 4.675E-08  3.500E-08      4m
  *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.
 
TABLE 3-10 (continued)*
REACTOR BUILDING STACK X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS I
;z B) 2.260 DAY DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE        .5-1        1-2      .2-3          3-4          4-5        5-10      10-20    20-30    30-40      40-50 S 7.216E-07  2.466E-07  9.389E-08      5.480E-08 3.915E-08    2.283E-08  2.568E-07 2.054E-07 1.293E-07  9.044E-08 SSW  6.135E-07  1.939E-07  6.803E-08      3.887E-08 2.761E-08    1.711E-08  2.523E-07  2.037E-07 1.283E-07  8.958E-08 SW  5.237E-07  1.688E-07  6.135E-08      3.534E-08 2.420E-08    2.951E-08  5.919E-08  3.753E-08 2.252E-08  1.524E-08 WSW    2.372E-07  7.796E-08  2.868E-08      1.686E-08  1.178E-08  1.068E-08  6.478E-08 5.075E-08 3.271E-08  2.329E-08 W  4.030E-08  1.516E-08  6.788E-09      4.842E-09 3.791E-09    2.319E-09  5.320E-08 4.275E-08 2.701E-08  1.885E-08 WNW    1.003E-07  4.296E-08  1.740E-08    9.980E-09 6.722E-09    3.856E-09  3.750E-08 2.877E-08  1.749E-08 1.184E-08 NW  3.527E-07  1.108E-07 4.070E-08      2.348E-08  1.656E-08  1.061E-08  4.212E-08 3.152E-08  1.929E-08 1.318E-08 NNW    8.102E-07  2.527E-07  9.458E-08      5.456E-08 4.185E-08    2.489E-08  6.578E-08 4.959E-08 3.141E-08  2.222E-08 N  9.771E-07  2.683E-07  9.200E-08      5.185E-08  3.585E-08  1.170E-07  1.354E-07 6.754E-08 4.242E-08  2.980E-08 NNE  8.955E-07  2.400E-07  8.338E-08      4.780E-08 3.652E-08    2.593E-07  1.975E-07 1.022E-07 6.538E-08  4.637E-08 NE  7.701E-07  2.615E-07  1.008E-07    6.397E-08  1.249E-07  3.741E-07  2.117E-07 1.091E-07 6.964E-08  4.924E-08 ENE  6.517E-07  2.587E-07  1.930E-07    6.384E-07 6.797E-07    3.168E-07  1.185E-07 5.737E-08 3.483E-08  2.362E-08 E  4.076E-07  1.734E-07 2.077E-07      8.230E-07 7.608E-07    3.544E-07  1.323E-07 6.397E-08 3.876E-08  2.622E-08 ESE  3.214E-07  1.808E-07  2.594E-07      6.752E-07 6.457E-07    2.419E-07  8.182E-08 5.288E-08 3.237E-08  2.223E-08  0.
SE  3.782E-07  1.313E-07  6.058E-08      1.264E-07  1.83 1E-07  9.206E-08  7.688E-08 4.861E-08 2.937E-08  1.995E-08 rri SSE  5.072E-07 1.659E-07  6.436E-08      3.761E-08 2.667E-08    1.602E-08  7.307E-08 5.778E-08 3.662E-08  2.574E-08      z
  *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.                z p
 
TABLE 3-10 (continued)*
REACTOR BUILDING STACK X/Q AND D/Q VALUES FROM 33- MET TOWER INSTRUMENTS I
CQ 8.000 DAY DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE          .5-1      1-2        2-3          3-4        4-5          5-10    10-20      20-30      30-40    40-50 S 6.870E-07  2.328E-07  8.735E-08    5.063E-08  3.617E-08    2.136E-08 2.685E-07  2.124E-07  1.324E-07 9.232E-08 SSW  5.873E-07  1.834E-07  6.329E-08    3.591E-08  2.550E-08    1.609E-08 2.654E-07  2.126E-07  1.331E-07 9.302E-08 SW  5.000E-07  1.589E-07. 5.674E-08    3.244E-08  2.213E-08    2.939E-08 6.169E-08  31893E-08  2.327E-08 1.572E-08 WSW    2.277E-07  7.372E-08  2.673E-08    1.569E-08  1.097E-08    1.039E-08 6.379E-08  4.882E-08  3.088E-08 2.177E-08 W  3.885E-08  1.453E-08  6.502E-09    4.682E-09  3.690E-09    2.284E-09 5.618E-08  4.473E-08  2.820E-08 1.980E-08 WNW    9.689E-08  4.142E-08  1.660E-08    9.466E-09  6.357E-09    3.678E-09 4.026E-08 LA                                                                                                  3.099E-08  1.907E-08 1.313E-08 NW  3.314E-07  1.027E-07  3.695E-08    2.114E-08  1.490E-08    9.821E-09 4.455E-08  3.327E-08  2.034E-08 1.398E-08 NNW    7.535E-07  2.309E-07  8,426E-08    4.795E-08  3.710E-08    2.238E-08 6.621E-08  4.896E-08  3.022E-08 2.097E-08 N  9.087E-07  2.430E-07  8.053E-08    4.462E-08  3.063E-08    1.170E-07  1.345E-07 6.441E-08  3.943E-08 2,716E-08 NNE    8.254E-07  2.140E-07  7.180E-08    4.061E-08  3.146E-08    2.547E-07 1.888E-07  9.353E-08  5.841E-08 4.079E-08 NE  7.158E-07  2.419E-07  9,164E-08    5.820E-08  1.213E-07    3.649E-07 1.977E-07  9.751E-08  6.074E-08 4.234E-08 ENE  6.161E-07  2.429E-07  1.735E-07    6.038E-07  6.510E-07    2.951E-07 1.058E-07  5.000E-08  3.012E-08 2.040E-08 E  3.818E-07  1.637E-07  1,920E-07    7.910E-07  7.284E-07    3.299E-07 1.181E-07  5.574E-08  3.355E-08 2.271E-08 ESE  2.985E-07  1.727E-07  2.452E-07    6.085E-07  5.804E-07    2.113E-07 6.544E-08  4.013E-08 2.362E-08  1.572E-08 SE  3.531E-07  1.220E-07  5.628E-08    1.251E-07  L.831E-07    9.018E-08 z
7.343E-08  4.569E-08  2.730E-08 1.843E-08 LA SSE  4.702E-07  1.524E-07  5.815E-08    3.373E-08  2.395E-08  I1.474E-08 6.877E-08                                      tTl 5.340E-08  3.326E-08 2.318E-08      z 0z
  *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.
 
TABLE 3-10 (continued)*
F-REACTOR                        BUILDING STACK X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS
* D) REACTOR BUILDING D/Q RELATIVE DEPOSITION PER UNIT AREA (m- 2) BY DOWNWIND SECTORS SEGMENT BOUNDARIES IN MILES DIRECTION FROM SITE        .5-1        1-2          2-3          3-4          4-5'      5-10      10-20      20-30      30-40      40-50 S 8.885E-09  2.377E-09    7.431E-10    3.700E-10  2.217E-10  9.108E-11  6.624E-11  3.829E-11  2.048E- I1 1.269E-II SSW  9.023E-09  2.273E-09    6.700E-10    3.268E-10    1.949E-10  8.066E-11  5.408E-II  3.067E-11    1.642E-11 I.018E-I1 SW  6.286E-09  1.610E-09    4.833E-10    2.342E-10  1.392E-10  7.102E-I1  3.623E-11  1.531E-11  8.210E-12  5.095E-12 WSW    2.845E-09  7.281E-10    2.180E-10    1.048E-10  6.225E- tI  3.339E-11  1.808E-11  7.657E-12  4.11IE-12  2.554E-12 W  5.769E-10  1.568E-10    4.928E-11    2.395E-11    1.420E-11  5.747E-12  6.235E-12  3.797E-12  2.030E-12  1.258E-12 WNW    1.240E-09  3.567E-10    1.181E-10    5.911E-11  3.535E711  1.628E-11  9.618E-12  4.787E-12  2.567E-12  1.593E-12 0&#xfd;'
as~
NW  3.659E-09  9.268E-10    2.776E-10    1.359E-10  8.094E-11  3.338E-11  2.197E-II  1.227E-11  6.572E-12  4.076E-12 NNW    7.346E-09  1.901E-09    5.865E-10    2.895E-10  1.767E-10  7.266E-11  4.503E-11  2.482E-II  1.327E-11  8.222E-12 N  1.193E-08  2.866E-09    8.251E-10    3.964E-10  2.342E-10  1.336E-10  7.241E-11  2.881E-11  L540E-11  9.540E-12 NNE  1.124E-08  2.537E-09    6.895E-10    3.220E-10  1.900E-10  1.548E-10  6.506E-II  2.593E-11  1.386E-11  8.586E-12 NE  1.112E-08  2.809E-09    8.728E-10    4.390E-10  4.039E-10  2.720E-10  8.331E-11  3.305E-11  1.765E-11  1.093E-II ENE  8.557E-09  2.413E-09    8.729E-10    5.940E-10  3.639E-10  1.401E-I0  4.062E-1I  1.614E-11  8.626E-12  5.343E&2 E  5.410E-09  1.341E-09    5.043E-10      4.333E-10  2.720E-10  1.046E-10  3.064E-11  1.228E-11  6.566E-12  4.069E-12 ESE  3.905E-09  9.113E-10    3.806E-10    4.051E-10  2.678E-10  1.030E-10  2.993E-11  1.207E- 1I  6.455E-12  3.999E-12 SE  5.287E-09  1.313E-09    3.966E-10      2.773E-10  2.729E-10    1.185E-10  3.999E-I1  1.647E-11  8.809E-12  5.457E-12      z SSE  6.171E-09  1.554E-09    4.791E-10      2.344E-10  1.384E-10  5.601E-11  3.363E-11  1.869E-11  9.998E- 12 6.195E-12 C.,'
rri
    *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 -                                    z December 2003.
z p
 
t" TABLE 3-11*
z TURBINE OR RADWASTE BUILDING X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS I
A) NO DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE        .5-1      1-2          2-3          3-4        4-5      5-10      10-20      20-30      30-40    40-50 S  3.868E-05  1.063E-05    3.796E-06    2.113E-06  1.409E-06  6.988E-07  2.950E-07  1.623E-07 1.098E-07 8.197E-08 SSW  3.547E-05  9.889E-06    3.578E-06    2.003E-06  1.340E-06  6.684E-07  2.848E-07  1.579E-07 1.073E-07 8.038E-08 SW  1. 118E-05 2.834E-06    9.522E-07    5.220E-07  3.457E-07  1.694E-07  6.985E-08  3.760E-08 2.502E-08 1.846E-08 WSW    8.08 1E-06 2.242E-06    8.126E-07    4.532E-07  3.023E-07  1.507E&07  6.441E-08  3.592E-08 2.453E-08 1.846E-08 W  6.440E-06  1.863E-06    6.919E-07    3.899E-07  2.616E-07  1.312E-07  5.649E-08  3.161E-08  2.160E-08 1.626E-08 WNW    5.951E-06  1.621E-06    5.769E-07    3.218E-07  2.150E-07  1.070E-07  4.535E-08  2.501E-08  1.692E-08 1.264E-08 NW  7.966E-06  2.086E-06    7.159E-07    3.920E-07  2.587E-07  1.262E-07  5.195E-08  2.806E-08  1.876E-08 1.390E-08 NNW    1.214E-05  3.174E-06    1.084E-06    5.899E-07  3.874E-07  1.876E-07  7.652E-08  4.11IE-08  2.743E-08 2.030E-08 N  1.573E-05  4.150E-06    1.434E-06    7.902E-07  5.241E-07  2.573E-07  1.066E-07  5.776E-08  3.863E-08 2.861E-08 NNE  1.819E-05  5.020E-06    1.797E-06    9.996E-07  6.657E-07  3.297E-07  1.391E-07  7.660E-08  5.183E-08 3.874E-08 NE  2.031E-05  5.567E-06    1.976E-06    1.092E-06  7.234E-07  3.557E-07  1.487E-07  8.147E-08  5.502E-08 4.109E-08 ENE  1.320E-05  3.531E-06    1.234E-06    6.840E-07  4.554E-07  2.250E-07  9.432E-08  5.153E-08  3.466E-08 2.578E-08 E  1.466E-05  3.959E-06    1.390E-06    7.709E-07  5.130E-07  2.531E-07  1.057E-07  5.763E-08  3.868E-08 2.873E-08 4 >r ESE  1.359E-05  3.615E-06    1.246E-06    6.839E-07  4.519E-07  2.202E-07  9.012E-08  4.833E-08  3.210E-08 2.365E-08 SE  1.377E-05  3.608E-06    1.228E-06    6.702E-07  4.412E-07  2.137E-07  8.676E-08  4.627E-08  3.065E-08 2.255E-08 SSE  1.150E-05  3.131E-06    1E-06 -06    6.057E-07  3.992E-07  1.9-07 07  8.066E-08  4.401E-08  2.969E-08 2.217E-08    z
  *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.              z
 
TABLE 3-11 (continued)*
TURBINE OR RADWASTE BUILDING X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS I
B) 2.260 DAY DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE          .5-1      1-2        2-3          3-4        4-5        5-10      10-20      20-30    30-40      40-50 S  3.843E-05 1.050E-05  3.714E-06    2.049E-06  1.354E-06  6.557E-07 2.603E-07  1.325E-07 8.336E-08  5.820E-08 SSW    3.523E-05 9.762E-06  3.499E-06    1.940E-06  1.286E-06  6.257E-07 2.500E-07  1.277E-07 8.036E-08  5.602E-08 SW  1.108E-05 2.784E-06  9.227E-07    4.990E-07  3.262E-07  1.542E-07 5,816E-08  2.798E-08  1.685E-08 1.137E-08 WSW    8.038E-06 2.219E-06  7.987E-07    4.424E-07  2.931E-07  1.434E-07 5.839E-08  3.059E-08  1.969E-08 1.400E-08 W  6.400E-06 1.841E-06  6.775E-07    3.784E-07  2.517E-07  1.233E-07 4.986E-08  2.572E-08  1.626E-08 1.135E-08 WNW      5.898E-06 1.594E-06  5.597E-07    3.082E-07  2.034E-07  9.779E-08 3.794E-08  1.867E-08 00                                                                                                            1.135E-08  7.666E-09 NW    7.909E-06 2.057E-06  6.985E-07    3.785E-07  2.472E-07  1.173E-07 4.493E-08  2.213E-08 1.358E-08  9.297E-09 NNW    1.208E-05 3.141E-06  1.064E-06    5.748E-07  3.747E-07  1.779E-07 6.895E-08  3.480E-08 2.197E-08  1.547E-08 N  1.563E-05 4.IOOE-06  1.404E-06    7.669E-07  5.043E-07  2.419E-07 9.463E-08  4.775E-08 3.001E-08  2.102E-08 NNE    1.811E-05 4.976E-06  1.770E-06    9.782E-07  6.474E-07  3.153E-07 1.273E-07  6.625E-08 4.248E-08  3.017E-08 NE  2.023E-05 5.525E-06  1.950E-06    1.070E-06  7.052E-07  3.412E-07 1.367E-07  7.076E-08  4.525E-08  3.205E-08 ENE    1.310E-05 3.476E-06  1.200E-06    6.572E-07  4.324E-07  2.069E-07 7.981E-08  3.921E-08 2.391E-08  1.624E-08 E  1.454E-05 3.896E-06  1.351E-06    7.404E-07  4.870E-07  2.327E-07 8.954E-08  4.391E-08  2.672E-08  1.810E-08 ESE    1.349E-05 3.564E-06  1.215E-06    6.601E-07  4.317E-07  2.045E-07 7.801E-08  3.827E-08  2.346E-08  1.606E-08 SE  1.366E-05 3.554E-06  1.196E-06    6.453E-07  4.200E-07  1.974E-07 7.415E-08                                    0 z
3.583E-08 2.171E-08  1.472E-08  0 SSE  1.144E-05 3.098E-06  1.084E-06    5.899E-07  3.859E-07  1.844E-07 7.237E-08  3.683E-08 2.327E-08  1.633E-08 z
  *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.                z p
 
6" TABLE 3-11 (continued)*
TURBINE OR RADWASTE BUILDING X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS C5                                                                                                                                I C) 8.000 DAY DECAY, DEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE          .5-1      1-2          2-3          3-4        4-5        5-10      10-20    20-30    30-40    40-50 S  3.458E-05 9.073E-06    3.065E-06    1.637E-06  1.054E-06    4.875E-07  1.785E-07 8.488E-08 5.106E-08 3.445E-08 SSW    3.171E-05 8.437E-06    2.889E-06    1.551E-06  1.002E-06    4.659E-07  1.721E-07 8.237E-08 4.973E-08 3.363E-08 SW  9.993E-06 2.416E-06    7.670E-07    4.028E-07  2.573E-07    1.172E-07 4.157E-08  1.910E-08 1.118E-08 7.378E-09 WSW    7.227E-06 1.914E-06    6.570E-07    3.518E-07  2.268E-07    1.055E-07 3.928E-08  1.903E-08 1.162E-08 7.941E-09 W  5.756E-06  1.590E-06    5.589E-07    3.021E-07  1.958E-07    9.157E-08 3.419E-08  1.653E-08 1.004E-08 6.824E-09 WNW    5.318E-06  1.382E-06  4.648E-07      2.484E-07  1.602E-07    7.408E-08 2.703E-08  1.274E-08 ON                                                                                                              7.590E-09 5.070E-09 NW  7.123E-06  1.780E-06  5.778E-07      3.033E-07  1.933E-07    8.783E-08 3.127E-08  1.452E-08 8.596E-09 5.727E-09 NNW    1.087E-05 2.711E-06    8.763E-07    4.577E-07  2.905E-07    1.313E-07 4.660E-08  2.171E-08 1.294E-08 8.690E-09 N  1.407E-05 3.543E-06    1.158E-06    6.124E-07  3.923E-07    1.796E-07 6.462E-08  3.027E-08 1.804E-08 1.209E-08 NNE    1.627E-05 4.288E-06    1.454E-06    7.765E-07  4.998E-07    2.313E-07 8.507E-08  4.076E-08 2.470E-08 1:678E-08 NE  1.817E-05 4.758E-06    1.600E-06    8.484E-07  5.435E-07    2.498E-07 9.11OE-08  4.344E-08 2.628E-08 1.784E-08 ENE    1.180E-05 3.011E-06  9.949E-07      5.285E-07  3.396E-07    1.561E-07 5.639E-08  2.638E-08 1.565E-08 1.043E-08 E  1.311E-05 3.376E-06    1.121E-06    5.956E-07  3.826E-07    1.756E-07 6.325E-08  2.953E-08 1.749E-08 1.163E-08 ESE  1.215E-05 3.085E-06    1.005E-06    5.291E-07  3.376E-07    1.532E-07 5.425E-08  2.502E-08 1.473E-08 9.769E-09 SE  1.231E-05 3.078E-06  9.910E-07    5.182E-07  3.293E-07    1.485E-07 5.205E-08  2.380E-08 1.393E-08 9.199E-09  I'J z 0
SSE  1.029E-05                                                                                                        0 2.674E-06  8.928E-07    4.699E-07  2.992E-07    L.363E-07 4.907E-08  2.319E-08 1.396E-08 9.439E-09 z
    *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.                H z
9
 
z-                                                            TABLE 3-11 (continued)*
TURBINE OR RADWASTE BUILDING X/Q AND D/Q VALUES FROM 33' MET TOWER INSTRUMENTS
&                                                                                                                                          I D) TURBINE OR RADWASTE DEPOSITION, D/Q.
RELATIVE DEPOSITION PER UNIT AREA (m-2) BY DOWNWIND SECTORS SEGMENT BOUNDARIES IN MILES DIRECTION FROM SITE          .5-1      1-2          2-3          3-4        4-5        5-10        10-20        20-30    30-40      40-50 S 4.538E-08  9.296E-09    2.427E-09    1.090E-09  6.166E-10  2.371E-10  6.860E-II    2.719E-11  1.452E-11  8.986E-12 SSW  3.690E-08  7.559E-09    1.973E-09    8.862E-10  5.013E-10  1.928E-10  5.577E-11    2.211E-1i  1.180E-II  7.307E-12 SW    1.972E-08  4.039E-09    1.054E-09    4.735E-10  2.679E-10  1.030E-10  2.980E-11      1181E-11 6.307E- 12  3.904E-12 WSW    9.730E-09  1.993E-09    5.203E-10    2.337E-10  1.322E-I0  5.083E-11. 1.471E-11    5.829E-12  3.113E-12  1.927E-12 W  4.377E-09  8.965E-10    2.340E-10    1.051E-10  5.946E-11  2.287E-11  6.616E-12    2.622E-12  1.400E-12  8.666E-13
--        WNW    5.900E-09  1.209E-09    3.155E-10    1.417E-10  8.016E- 11  3.083E-11  8.918E-12    3.534E-12  1.887E-12  1.168E- 12 NW    1.485E-08  3.042E-09  7.942E-10      3.567E-10  2.018E-10  7.760E-11  2.245E-11    8.898E-12  4.751E-12  2.941E-12 NNW    3.024E-08  6.193E-09    1.617E-09    7.261E-10  4.108E-10  1.580E-10  4.5"70E- I1  1.811E-I1  9.672E-12  5.987E-12 N  3.749E-08  7.678E-09  2.005E-09      9.003E-10  5.093E-10  1.959E-10  5.666E-I1    2.246E-11  1.199E-1I  7.423E-12 NNE  3.428E-08  7.021E-09    1.833E-09    8.232E-10  4.657E-10  1.791E-10  5.181E-11    2.053E-1I  1.096E-11  6.787E-12 NE  4.442E-08  9.098E-09'  2.375E-09      1.067E-09  6.034E-10  2.321E-10  6.713E-I I    2,661E-11  1.421E-11  8.795E-12 ENE  2.574E-08  5.273E-09    1,377E-09    6.183E-10  3.498E-10  1.345E-10  3.891E-11    1.542E-11  8.236E-12  5.098E-12 E  1.914E-08  3.920E-09    1.023E-09    4.597E-10  2.600E-10  1.000E-10  2.893E-I1    1.147E-11  6,123E-12  3.790E-12 ESE  1.851E-08  33791E-09    9.897E-10      4.445E-10  2.515E-10    9.670E-11  2.798E-1I    1.109E-11                          04 rT, SE 5.921E-12  3.665E-12 z
2.192E-08  4.490E-09    1.172E-09    5.264E-10  2.978E-10    1.145E-10  3.313E-11    1.313E-11  7.012E-12  41340E-12 SSE  21446E-08  5.011E-09    1.308E-09      5.875E-10                                                                              rr, 3.324E-10    1.278E-10  3.698E-I1    1.465E-11  7.826E-12  4.844E-12        z
    *Based on meteorological data at 33-ft level collected January 1998 - December 2000 and January 2002 - December 2003.                    zp
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-13 CHARACTERISTICS OF GASEOUS EFFLUENT RELEASE POINTS Reactor            Radwaste              Turbine Building          Building              Building Height of release point above ground level (m)                  70.3              20.4                  36.3 Annual average rate of air flow from release point (m3/sec)                                37.8              39.2                  169.9 Annual average heat flow from release point (cal/sec)            1.06 x 106        2.9 x 106            9.1 x 10i Type and size of release                Duct                3 Louver houses,    4 Exhaust point (m)                                1.14 x 3.05        Each                fans, Each 1.37 x 2.44 x 0.75  1.45 x 2.01 Effective vent area (Mi)                3.48                11.58**              8.74***
Vent velocity (m/sec)*                  10.9                3.4                  19.5 Effective diameter (m)                  2.1                3.8                  3.3 (rr2 = area)
Building height (m)                    70.1                31.8****            42.4
*Reactor Building exhaust in vertical direction. Radwaste and Turbine Building exhaust in horizontal plane.
**Equivalent to two vents (as per FSAR Section 9.4.3.2, 2 of 3 will normally be in operation).
***Equivalent to three vents (as per FSAR Section 9.4.6.2.2, 3 of 4 will normally be in operation).
****Height of the Radwaste and Control Building.
71
 
AMENDMENT NO. 44 MAY 2005 TABLE 3-14 REFERENCES FOR VALUES LISTED IN TABLES 3-8 and 3-9 Reference 1  U.S. Map Reference 2  Health Physics Calculation Log 93-2 Reference 3  Regulatory Guide 1.109, Revision 1, Table E-15 Reference 4  Columbia Generating Station FSAR, Table 2.3-1 Reference 5  Columbia Generating Station FSAR, Page 2.3-2 Reference 6  Columbia Generating Station FSAR Table 2.1-2 Reference 7  1986 50-Mile Land Use Census, Supply System REMP Reference 8  Effluent Analysis for Applicable Time Period Reference 9  NUREG/CR-2919, XOQDOQ: Computer Program For The Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, September 1982.
LDCN-04-012 72
 
AMENDMENT NO. 35 MARCH 2002 TABLE 3-15 DESIGN BASE PERCENT NOBLE GAS (30-MINUTE DECAY)*
Isotone                      Percent of Total Activityj Kr-83M                      2.9 Kr-85M                      5.6 Kr-85                        0 Kr-87                        15 Kr-88                        18 Kr-89                        0.2 Xe-131M                      0.02 Xe-133M                      0.3 Xe-133                      8.2 Xe-135M                      6.9 Xe-135                      22 Xe-137                      0.7 Xe-138                      21
*From FSAR Table 11.3-1 73
 
AMENDMENT NO. 47 AUGUST 2008 TABLE 3-16 ANNUAL DOSES AT TYPICAL LOCATIONS****
Source: Gaseous Effluent Whole Distance        Occupancy        Body Dose  Thyroid Dose Location                            (Miles)        (hrs/yr)        (mrem/yr)    (mrem/yr)
BPA Ashe Substation            0.5 N            2080              1.1E+00      1.7E+00 DOE Train                      0.5 SE*          78              6.7E-02      1.OE-01 Wye Burial Site                0.5 WNW          8                4.1E-03      6.5E-03 WNP-1                          1.2 ESE          2080              3.8E-02      1.3E-01 WNP-4                          1.0 ENE          2080              7.OE-02      1.1E-01 Visitor Center                  0.08 ESE        8                8.6E-02      1.3E-01 Taylor Flats**                  4.2 ESE          8760              3.1E-02      5.2E+00 Site Boundary***                1.212 SE        8760              1.1E+00      1.7E+00 National Guard                  0.5 N            8760              4.1E+00      7.7E+00 National Guard                  0.5 SE          8760              4.4E+00      8.9E+00 National Guard                  1.2 N            8760              1.1E+00      2.4E+00 National Guard                  1.2 SE          8760              3.2E-01      2.0E+00
* The sector with the highest X/Q values (within 0-0.5 mile radius) was used.
** Closest residential area representative of maximum individual dose from plume, ground, ingestion, and inhalation exposure pathways. Included for comparison.
*** Assumed continuously occupied. Actual occupancy is very low. Doses from Inhalation and Ground Exposure pathways. No food crops.
****Calculation inputs as described in Section 3.1.
LDCN-07-033 74
 
AMENDMENT NO. 40 JULY 2003 TABLE 3-17 ANNUAL OCCUPIED AIR DOSE AT TYPICAL LOCATIONS**
Annual Beta Air dose        Annual Location                                        (mrad)      Gamma Air Dose BPA Ashe Substation                8.9E-01                1.5E+00 DOE Train                          5.3E-02                9.2E-02 Wye Burial Site                    3.2E-03                5.7E-03 WNP-1                              3.3E-02                2.8E-02 WNP-4                              5.3E-02                8.5E-02 Visitor Center                    7.OE-02                1.2E-01 Taylor Flats*                      2.3E-02                1.4E-02 Site Boundary                      8.7E-01                1.5E+00 National Guard (0.5 mi N)          3.2E+00                5.7E + 00 National Guard (0.5 mi SE)        3.5E+00                6.2E+00 National Guard (1.2 mi N)          8.6E-01                1.5E+00 National Guard (1.2 mi SE)        2.5E-01                3.6E-01
*Closest residential area.
**Calculation inputs as described in Section 3.1.
LDCN-03-037 75
 
AMENDMENT NO. 47 AUGUST 2008 90OM3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS Figure 3-1 76
 
AMENDMENT NO. 35 MARCH 2002 ReaQtQLBtld Refueling Pump Rooms Main Steam Tunnels Offgas Treatment System Standby Gas Mechanical Vacuum Pumps R-adwaste-Bldg To Hot Machine Shop                                    Atmosphere Chemistry Labs Demineralizer Room Radwaste Proc. Area Control Room SIMPLIFIED BLOCK DIAGRAM OF GASEOUS WASTE SYSTEM Figure 3-2 77
 
AMENDMENT NO. 35 MARCH 2002 SIMPLIFIED BLOCK DIAGRAM OF OFF-GAS TREATMENT SYSTEM Figure 3-3 78
 
AMENDMENT NO. 35 MARCH 2002
                                        -      AUXILIARY STEAM LINE
                                        -. TO CONDENSATE RETURN TANK VIA TRAP FROM BOILER FEED PUMPS
                                        -    SERVICG AIR FOR ATOMIZINO
                                        - STEAM FOIl ATOMIZING
                                        " FROM FUEL OIL SERVICE PUMP FC--P-2A(B)
TO AUX. GOILER
]LOWDOWN TANK AUXILIARY BOILER Figure 3-4 79
 
AMENDMENT NO. 38 APRIL 2003 4.0    COMPLIANCE WITH 40 CFR 190 - TOTAL DOSE 4.1    Requirement for Operability (RFO)
RFO 6.2.3 states, "The annual dose or dose commitment to any MEMBER OF THE PUBLIC in the UNRESTRICTED AREA beyond the site boundary, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid and to less than or equal to 75 mrems to the thyroid."
4.2    ODCM Methodology for Determining Dose and Dose Commitment from Uranium Fuel Cycle Sources The annual dose or dose commitment to a MEMBER OF THE PUBLIC for the uranium fuel cycle sources is determined as:
a)    Dose to the whole body due to the release of radioactive materials in liquid effluents.
b)      Dose to any organ due to the release of radioactive materials in liquid effluents.
c)    Air doses due to noble gases released in gaseous effluents.
d)      Dose to any organ due to the release of radioiodines, tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents.
e)      Dose due to direct radiation from the plant.
The annual dose or dose commitment to a MEMBER OF THE PUBLIC from the uranium fuel cycle sources is determined whenever the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceed twice the limits in RFO 6.2.1.2.a, 6.2.1.2.b, 6.2.2.2.a, 6.2.2.2.b, 6.2.2.3.a, or 6.2.2.3.b. Direct radiation measurements may also be made to determine if the limits of RFO 6.2.3.1 have been exceeded.
4.2.1 Total Dose from Liquid Effluents The annual dose to a MEMBER OF THE PUBLIC from liquid effluents will be determined using NRC LADTAP ii computer code or the methodology presented by Equation (10) in Section 2.5. It is assumed that .dose contribution pathways to a MEMBER OF THE PUBLIC do not exist for areas within the site boundary.
4.2.2 Total Dose from Gaseous Effluents The annual dose to a MEMBER OF THE PUBLIC from gaseous effluents will be determined using NRC GASPAR II computer code or the methodology presented by Equations (10), (11) and (13) in Section 3.5. Appropriate atmospheric dispersion parameters will be used.
LDCN-03-007 80
 
AMENDMENT NO. 35 MARCH 2002 4.2.3 Direct Radiation Contribution The dose to a MEMBER OF THE PUBLIC due to direct radiation from the reactor plant and the Independent Spent Fuel Storage Installation (ISFSI) will be determined using thermoluminescent dosimeters (TLDs) or may be calculated. TLDs are placed at sample locations and analyzed as per Table 6.3.1-1.
81
 
AMENDMENT NO. 35 MARCH 2002 5.0    RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological environmental monitoring is intended to supplement radiological effluent monitoring by verifying that measurable concentrations of radioactive materials and levels of radiation in the environment are not greater than expected based on effluent measurement and dose modeling of environmental exposure pathways. The Radiological Environmental Monitoring Program (REMP) for Columbia Generating Station provides for measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides for which the highest potential dose commitment to a MEMBER OF THE PUBLIC would result due to plant operations. The REMP implements Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50. It also provides the information required by Technical Specification Section 5.6.2.
The REMP is designed to conform to regulatory guidance provided by Regulatory Guide 4.1, 4.8, 4.15 and the Radiological Assessment Branch Technical Position (BTP), taking into consideration certain site specific characteristics. The unique nature of the site on Federally owned and administered land (Hanford Site) dedicated to energy facilities, research, waste management and as a natural reserve, forms the basis for many of the site specific parameters.
Among the many site specific parameters considered is demographic data such as:
"      No significant clusters of population including schools, hospitals, business facilities or primary public transportation routes are located within 8 km (5 mile) radius of the plant.
"      No private residences are located on the Hanford Site.
* The closest resident is east of the Columbia River at a distance of approximately 4 miles.
82
 
AMENDMENT NO. 35 MARCH 2002 5.1    Radiological Environmental Monitoring Program (REMP)
Environmental samples for the REMP are collected in accordance with ODCM RFO 6.3.1 and Table 6.3.1-1. This table provides a detailed outline of the environmental sampling plan items by sample type, sampling and collection frequency, and type and frequency of analysis.
Deviations from the sampling frequency detailed in Table 6.3.1-1 may occur due to circumstances such as hazardous conditions, malfunction of automatic sampling equipment, seasonal unavailability, or other legitimate reasons. When sample media is unobtainable due to equipment malfunction, special actions per program instruction shall be taken to ensure that corrective action is implemented prior to the end of the next sampling period. In some cases, alternate sample collection may be substituted for the missing specimen.
Sampling stations are described in Table 5-1. Each station is identified by an assigned number or alphanumeric designation, meteorological sector (16 different, 22-1/2' compass sections) in which the station is located, and radial distance from containment as estimated from map positions. Also included in Table 5-1 is information identifying the type(s) of samples collected at each station. Figures 5-1 and 5-2 depict the geographical locations of each of the sample stations listed in Table 5-1.
5.2    Land Use Census A Land Use Census shall be conducted in accordance with the requirements of ODCM RFO 6.3.2. It shall identify within a distance of 8 km (5 miles) in each of the 16 meteorological sectors, the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 50m 2 (500 ft2) producing broad leaf vegetation. Field activities pertaining to the Land Use Census will be initiated during the growing season and completed no later than September 30 each year. The information obtained during the field survey is used along with other demographic data to assess population changes in the unrestricted area that might require modifications in the sampling plan to ensure adequate evaluation of dose or dose commitment.
5.3    Laboratory Intercomparison Program A Laboratory Intercomparison program shall be conducted in accordance with the requirements of ODCM RFO 6.3.3. Analysis of REMP samples is contracted to a provider of radiological analytical services. By contract, this analytical service vendor is required to conduct all activities in accordance with Regulatory Guides 4.1, 4.8, and 4.15 and to include in each quarterly report, actions pertinent to their participation in the Interlaboratory Comparison Program. A precontract award survey and periodic audit at the contractor's facility ensure that the contractor is participating in the Crosscheck Program, as reported.
83
 
AMENDMENT NO. 35 MARCH 2002 Besides the vendor's required participation in the Interlaboratory Comparison Program, the Department of Health (DOH) of the State of Washington oversees an analytical program for the Energy Facility Site Evaluation Council (EFSEC) to provide an independent test of Columbia Generating Station REMP sample analyses. The Columbia Generating Station/DOH split samples are analyzed by Washington State's Office of Public Health Laboratories and Epidemiology, Environmental Radiation Laboratory (ERL). The results of the ERL analysis and Interlaboratory Comparison Program data are included in an annual report, "Environmental Radiation Program, Environmental Health Surveillance, State of Washington" and are available for comparison with the Columbia Generating Station data.
Energy Northwest participates in the International Intercomparison of Environmental Dosimeter Program. Results of this intercomparison program are reported in the REMP Annual Report, when available.
84
 
z TABLE 5-1 t
REMP LOCATIONS Station Sector Radial Milesa TLD  AP/AI    SW  DW GW SE MI FI GP 1      S        1.25      0 2    NNE        1.45      0 4    SSE      9.57      0    0 5    ESE      7.72      0 6      S        7.72      0 7    WNW        2.83      0 8    ESE      4.39      0    0 00  9A*    WSW      28.35      0    0 9B    WSW      32.82                                  0    0 9C    WSW      32.15                                        0 10      E        3.16      0 13    SW        1.26      0 14    WSW        1.26      0
                                                                      ,m 15    W        1.24      0                                    k&#x17d;Z 16    WNW        1.21      0 z-17    NNW        1.19      0                                    z p.
 
z TABLE 5-1 (continued) 6*
REMP LOCATIONS Station Sector Radial Milesa TLD    AP/AI      SW    DW GW SE MI Fl GP 18      N        1.16    0 19    NE        1.74    0 20    ENE        1.93    0 22      E        2.08      0 24      SE        1.87    0 25    SSE        1.50    0 0'  26*      E        3.19                          0    0 a,%
27      E        3.19                          0 29    SSE      11.57                              0 30      E      3.28                                            0 31    ESE      1.06                                  0 34    ESE      3.32                                      0 36    ESE      7.33                                        0 37    SSE      14.62                                              0 38*    ESE        26.5 0    m z
0--
 
z                                  TABLE 5-1 (continued)
REMP LOCATIONS Station Sector Radial Milesa TLD AP/AI    SW        DW GW SE MI Fl GP 40      SE      6.51      0    0 41      SE      5.79      0 42    ESE      5.85. 0 43      E        5.16      0 44    ENE        5.90      0 45    ENE        4.45      0 46      NE      4.99      0 00  48      NE      4.59          0 49    NW        1.19      0 50    SSW        1.26      0 51    ESE      2.14      0 52      N        0.07                                0 53      N        7.54      0                                          m}
54    NNE        6.08      0                                          z 0
55    SSE      6.05      0                                        0 (A
m 56    SSW        6.65      0                                          z.
z 57      N        0.70          0                                      p 102G    SSE      0.56                                            0
 
TABLE 5-1 (continued)
REMP LOCATIONS Station      Sector      Radial Milesa  TLD        AP/AI      SW        DW  GW SE MI FI GP
    *Control location.
O        -  Radiological Environmental RFO sample collected at station.
a        -  Estimated from center of Plant Containment from map positions.
TLD    =  Thermoluminescent dosimeter AP/AI =    Air Particulate and Iodine SW      =  Surface Water (River Water)
DW      =  Drinking Water 00  GW      =  Ground Water SE      =  Shoreline Sediment MI      =  Milk FI      =  Fish GP      =  Garden Produce M
                                                                                              *Z z
t~JIr 9
 
AMENDMENT NO. 35 MARCH 2002 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE LOCATIONS INSIDE OF 10 MILE RADIUS Figure 5-1 99
 
F-0 OtheUo                              WASHINGTON 0 Connell mile radius
* Pomeroy Clarkston 0 Daytoni
                                                  & Eureka 0
                                                                *Waia walls OREGON Columbia 1 inch = 16 miles n-I 0        8        16 z
A Sample Locations n-I 90028aA1 0
k) z SEP 2001 z
p Radiological Environmental Monitoring Sample Locations Outside of 10-Mile Radius Figure 5-2
 
ODCM APPENDIX COLUMBIA GENERATING STATION Radiological  Effluent And Radiological  Environmental  Monitoring Controls
 
OFFSITE DOSE CALCULATION MANUAL APPENDIX LIST OF EFFECTIVE PAGES PAGE      AMEND. NO.                    PAGE        AMEND. NO.
6.2.2.4-1          35 35                                            35 35                        6.2.2.4-2 i6 6.0-I                                    6.2.2.5-1          47 38                                            47 6.2.2.5-2 6.0.1-1        35                                            35 35                        6.2.2.6-1 6.0.1-2                                  6.2.2.6-2          35 6.0.1-3        47 6.2.3-1            38 6.1.1-1        35                                            35 35                        6.2.3-2 6.1.1-2                                  6.2.3-3            35 6.1.1-3        35 6.1.1-4        37 6.3.1-1            35 6.1.2-1        35                                            35 35                        6.3.1-2 6.1.2-2                                  6.3.1-3            35 6.1.2-3        45                                            35 43                        6.3.1-4 6.1.2-4                                  6.3.1-5            35 6.1.2-5        43                                            35 46                        6.3.1-6 6.1.2-6                                  6.3.1-7            35 6.1.2-7        44                                            35 6.3.1-8 6.3.1-9            35 6.2.1.1-1      35                                            35 35                        6.3. 1-10 6.2.1.1-2                                6.3. 1-11          35 6.2.1.1-3      35                                            35 6.3. 1-12 6.3. 1-13          46 6.2.1.2-1      38 6.2.1.2-2      35                                            35 6.3.2-1 6.3.2-2            35 6.2.1.3-1      38 6.2.1.3-2      35                                            35 6.3.3-1 6.2.2.1-1      38                                            35 35                        6.4.1-1 6.2.2.1-2 6.2.2.1-3      35                                            35 35                        6.4.2-1 6.2.2.1-4                                6.4.2-2            35 6.2.2.1-5      35 6.2.2.1-6      35                                            35 35                        6.4.3-1 6.2.2.1-7                                6.4.3-2            35 6.2.2.2-1      35 6.2.2.2-2      35 6.2.2.3-1      35 6.2.2.3-2      35 ODCM LEP- I              Amendment 47, AUG 2008
 
OFFSITE DOSE CALCULATION MANUAL APPENDIX LIST OF EFFECTIVE PAGES AMEND. NO.                    PAGE        AMEND. NO.
PAGE 35                          6.2.2.2-1        35 B 6.1.1-1                                  B 36                          6.2.2.2-2        35 B 6.1.1-2                                  B 36                          6.2.2.2-3        35 B 6.1.1-3 B 6.1.1-4        37 36                        B 6.2.2.3-1        35 B 6.1.1-5 36                        B 6.2.2.3-2        35 B 6.1.1-6 36                        B 6. 2.2.3-3        35 B 6.1.1-7 B 6.1.1-8        36 B 6.2.2.4-1        35 44                        B 6. 2.2.4-2        35 B 6.1.2-1 44                        B 6. 2.2.4-3        35 B 6.1.2-2 B 6.1.2-3        35 42                        B 6.2.2.5-1        47 B 6.1.2-4 35                        B 6,2.2.5-2        47 B 6.1.2-5 35                        B 6.2.2.5-3        47 B 6.1.2-6 B 6.1.2-7        35 35                        B 6.2.2.6-1        35 B 6.1.2-B 45                        B 6.2 .2.6-2      35 B 6.1.2-9 6.1.2-10        45                        B 6.2.2.6-3        35 B
B 6.1.2-11        44 6.1.2-12        42                        B 6.2.3-1          38 B
6.1.2-13        46                      B 6.2.3-2          35 B
46                        B 6.2.3-3          38 B 6.1.2-14 35                      B  6.3.1-1          35 B 6.2.1.1-1 35                      B  6.3.1-2          35 B 6.2.1.1-2 35                      B  6.3.1-3          35 B 6.2.1.1-3 35                        B 6.3.1-4          35 B 6.2.1. 1-4 B  6.3.1-5          35 38                      B  6.3.1-6          35 B 6.2.1.2-1 38                      B  6.3.1-7          35 B 6.2.1.2-2 B 6.2.1.2-3      35 B 6.3.2-1          35 35                      B 6.3.2-2          35 B 6.2.1.3-1 38                      B 6.3.2-3          35 B 6.2.1.3-2 B 6.2.1.3-3      35 B 6.3.3-1          35 38                      B 6.3.3-2          35 B 6.2.2.1-1 B 6.2.2.1-2      35 B 6.2.2.1-3      35 B 6.2.2.1-4      35 B 6.2.2. 1-5      35 B 6.2.2.1-6      35 B 6.2.2.1-7      35 B 6.2.2.1-8      35 B 6.2.2. 1-9      35 ODCM LEP-2                          47, AUG 2008 ODCM EP-2Amendment
 
OFFSITE DOSE CALCULATION MANUAL APPENDIX TABLE OF CONTENTS RADIOLOGICAL    EFFLUENT AND RADIOLOGICAL ENVIRONMENTAL MONITORING CONTROLS
 
==6.0        INTRODUCTION==
.................            .........................                                  .. 6.0-1 6.0.1      DEFINITIONS .........................................                                            ... 6.0.1-1 6.1        INSTRUMENTATION 6.1.1            Radioactive Liquid Effluent Monitoring Instrumentation ...........                    ....................                                6.1.1-1 6.1.2            Radioactive Gaseous Effluent Monitoring Instrumentation .............              ....................                            ... 6.1.2-1 6.2      RADIOACTIVE EFFLUENTS 6.2.1            Liquid Effluents ............              ..................                          ...      6.2.1.1-1 6.2.1.1        Liquid Concentration ..........                  ................                      ...      6.2.1.1-1 6.2.1.2        Liquid Dose .............            .....................                                  ."6.2.1.2-1 6.2.1.3        Liquid Waste Management System ..... ...........                                        ..      6.2.1.3-1 6.2.2          Gaseous Effluents ...............                    ..................                        6.2.2.1-1 6.2.2.1        Dose Rate ...............          ......................                                  .. 6.2.2.1-1 6.2.2.2        Dose - Noble Gases .......... ..................                                        .. 6.2.2.2-1 6.2.2.3        Dose - 131, 133, Tritium, and Radionuclides in Particulate Form ...              .................                                          6.2.2.3-1 6.2.2.4        Gaseous Offgas Treatment System ....                          ...........              ...      6.2.2.4-1 6.2.2.5        Reactor Building Sump Vent Exhaust Filter System                                      .          6.2.2.5-1 6.2.2.6        VENTING or PURGING ..............                        .................                        6.2.2.6-1 6.2.3,          Total Dose ................              ......................                                .. 6.2.3-1 6.3        RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.1          Radiological Environmental Monitoring Program                                    ......            6.3.1-1 6.3.2          Land Use Census            .....................                                                    6.3.2-1 6.3.3          Interlaboratory Comparison Program ... ...........                                                  6.3.3-1 6.4            REPORTING REQUIREMENTS 6.4.1          Annual Radiological Environmental Operating Report                                                  6.4.1-1 6.4.2          Radioactive Effluent Release Report ....                              ..........                    6.4.2-1 6.4.3          Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems ......... ................                                          ... 6.4.3-1 Columbia Generating Station                i                                        Amendment 35, MAR 2002
 
OFFSITE DOSE CALCULATION MANUAL APPENDIX BASES BASES FOR RADIOLOGICAL EFFLUENT AND RADIOLOGICAL ENVIRONMENTAL            MONITORING CONTROLS B 6.1      INSTRUMENTATION B 6.1.1          Radioactive Liquid Effluent Monitoring Instrumentation . . . . . . . . . . .                          B 6.1.1-1 B 6.1.2          Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . . . . .          B 6.1.2-1 B  6.2      RADIOACTIVE EFFLUENTS B  6.2.1. I      Liquid Concentration        . . . . . . . . . . . . . .      B  6.2.1.1-1 B  6.2.1.2      Liquid Dose                                                  B  6.2.1.2-1 B  6.2.1.3      Liquid Waste Management System        . . . .....            B  6.2.1.3-1 B  6.2.2.1      Dose Rate . . . . . . . . . . . . . . . . . . . .            B  6.2.2.1-1 B  6.2.2.2      Dose - Noble Gases        . . . . . . .                      B  6.2.2.2-1 B  6.2.2.3      Dose  - 131,  133,  Tritium,    and Radionuclides in Particulate Form . . . . . . . . . . . . . . .            B  6.2.2.3-1 B  6.2.2.4      Gaseous Offgas Treatment System .........                    B  6.2.2.4-1 B  6.2.2.5      Reactor Building Sump Vent Exhaust Filter System              B 6.2.2.5-1 B  6.2.2.6      VENTING or PURGING        . . . . . . . . . . . . . . .      B 6.2.2.6-1 B  6.2.3        Total Dose      . . . . . . . . . . . . . . . . . . .          B 6.2.3-1 B 6.3      RADIOLOGICAL ENVIRONMENTAL MONITORING B 6.3.1          Radiological Environmental Monitoring Program                  B 6.3.1-1 B 6.3.2        Land Use Census . . . . . . . . . . . . . . .                    B 6.3.2-1 B 6.3.3          Interlaboratory Comparison Program        .....                B 6.3.3-1 Columbia Generating Station                    ii                Amendment 35, MAR 2002
 
Introduction 6.0 Introduction
 
==6.0        INTRODUCTION==
 
The Columbia Generating Station Off-Site Dose Calculation Manual (ODCM) Appendix contains the Radiological Effluent and Radiological Environmental Monitoring Controls required by Technical Specification 5.5.1.b. These controls are written in the format of the Columbia Generating Station Licensee Controlled Specifications (LCS) to provide a consistent set of requirements for the station. The definitions contained in the Technical Specifications and the LCS apply to this Appendix. The provisions of Technical Specification Surveillance Requirements (SR) 3.0.2 and 3.0.3 are applicable to ODCM surveillance frequency. In addition, Section 1.0 of the LCS that gives the Requirements for OPERABILITY (RFO) Applicability (RFOs 1.0.1 through 1.0.6) and SR Applicability (SR 1.0.1 through SR 1.0.4) also apply to all RFOs and SRs in this Appendix.
Columbia Generating Station            6.0-1            Amendment 38, APRIL 2003
 
Definitions 6.0.1 6.0.1  Definitions The following definitions and definitions of Technical Specifications and Licensee Controlled Specifications apply to the Appendix of the ODCM.        They are defined so that uniform interpretation of these requirements for OPERABILITY may be achieved. The defined terms appear in capitalized type and shall be applicable throughout the ODCM.
Term                    DEFINITIONS GASEOUS OFFGAS          A GASEOUS OFFGAS TREATMENT SYSTEM (e.g., the TREATMENT SYSTEM        "augmented offgas system") is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
LOWER LIMIT OF          The LLD is defined, for purposes of this ODCM DETECTION (LLD)          Specification, 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:
4.66 Sb LLD =
E e V e 2.22 x 106
* Y e exp (-XAt)
Where:
LLD is the "a priori" lower limit of detection as defined above as microcuries per unit mass or volume.
S, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate as counts per minute.
E is  the counting efficiency as counts per disintegration.
V is the sample size in units of mass or volume.
(continued)
Columbia Generating Station          6.0.1-1                  Amendment 35, MAR 2002
 
Definitions 6.0.1 6.0.1  Definitions LOWER LIMIT OF            2.22 x 10' is the number of disintegrations per DETECTION (LLD)          minute per microcurie.
(continued)
Y is the fractional  radiochemical yield when applicable.
A is the radioactive decay constant for the particular radionuclides and Ar for plant effluents is the lapsed time between the midpoint of sample collection and time of counting.
Typical values of E, V, Y, and Atshould be used in the calculation.
It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (afer the fact) limit for a particular measurement.
MEMBER(S)  OF THE        MEMBERS OF THE PUBLIC shall include all persons who PUBLIC                  are not occupationally associated with the plant.
This category does not include employees of the utility, its contractors or vendors.      Also excluded from this category are persons who enter the site to service equipment or to make deliveries.      This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
OPERATIONAL CONDITION    See Technical    Specification definition of MODE.
CONDITION PURGE - PURGING          PURGE - PURGING shall be the controlled process of discharging air or gas from a containment to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the containment.
SITE BOUNDARY            The SITE BOUNDARY shall be as generally depicted in ODCM, Figure 3-1.
The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
(continued)
Columbia Generating Station            6.0. 1-2              Amendment 35, MAR 2002
 
Definitions 6.0.1 6.0.1  Definitions  (continued)
SOURCE CHECK              A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
UNRESTRICTED AREA        An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
VENTILATION EXHAUST      A VENTILATION EXHAUST TREATMENT SYSTEM shall be any TREATMENT SYSTEM        system designed and installed to reduce gaseous radioiodine or 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 Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
I VENTING                  VENTING shall be the controlled process of discharging air or gas from a containment to 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 system names, does not imply a VENTING process.
Columbia Generating Station          6.0.i1-3            Amendment 47, AUG 2008
 
Radioactive Liquid Effluent Monitoring Instrumentation 6.1.1 6.1  INSTRUMENTATION 6.1.1    Radioactive  Liquid Effluent Monitoring Instrumentation RFO 6.1.1.1        The radioactive  liquid effluent monitoring    instrumentation channels  in Table 6.1.1-1 shall  be OPERABLE.
APPLICABILITY:      In accordance with Table 6.1.1-1.
COMPENSATORY MEASURES
----------------------------------- NOTE------------------------------------
Separate condition entry is allowed for each channel.
CONDITION              REQUIRED COMPENSATORY MEASURE        COMPLETION TIME A. One or more required      A.1    Enter the condition          Immediately radioactive liquid                referenced in effluent monitoring              Table 6.1.1-1 for the instrumentation                  channel.
channels inoperable.
B. As required by              B.1  Perform SR 6.2.1.1.1 on      Prior to Compensatory Measure              two independent samples      radioactive A.1 and referenced in            of the batch to be            liquid release Table 6.1.1-1.                    released.                    through the radwaste AND                                effluent line B.2  Verify the associated        Prior to release rate                radioactive calculations and the,        liquid release discharge valve lineup      through the using two qualified          radwaste members of the technical    effluent line staff.
AND (continued)
Columbia Generating Station              6.1. 1-1              Amendment 35, MAR 2002
 
Radioactive Liquid Effluent Monitoring Instrumentation 6.1.1 COMPENSATORY MEASURES                                            __________
CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME B. (continued)              B.3    Restore the channel to      30 days OPERABLE status.
C. As required by            C.1  Analyze a grab sample        Once per 12 Compensatory Measure            for radioactivity (beta      hours A.1 and referenced in          or gamma) of the Table 6.1.1-1.                associated pathway. The LLD shall be < 1E-7 4sCi/ml.
AND C.2  Restore the channel to      30 days OPERABLE status.
D. As required by          D.1 Estimate the flow rate        At the beginning Compensatory Measure          through the associated      of the release A.1 and referenced in        pathway,                    and once per 4 Table 6.1.1-1.                                            hours during releases through the associated li ne AND D.2 Restore the channel to        30 days OPERABLE status.
'E. Required Compensatory    E.1 Prepare and submit, in        Upon submittal Measure B.3, C.2, or          the Radioactive Effluent    of current D.2 and associated            Release Report, the          calendar year Completion Time not          reason the channel was      Radioactive met.                          not restored to OPERABLE    Effluent Release status within 30 days.      Report Columbia Generating Station          6.1. 1-2 6112Aedet3,MR20      Amendment 35, MAR 2002
 
Radioactive Liquid Effluent Monitoring Instrumentation 6.1.1 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.1.1.1      Perform CHANNEL CHECK.                      24 hours SR  6.1.1.2      Perform SOURCE CHECK.                        Prior to each radioactive rel ease SR  6.1.1.3      Perform SOURCE CHECK.                        31 days SR  6.1.1.4      Perform CHANNEL FUNCTIONAL TEST.            92 days SR  6.1.1.5      Perform CHANNEL CALIBRATION.                18 months Columbia Generating Station          6. 1. 1-3          Amendment 35, MAR 2002
 
Radioactive Liquid Effluent Monitoring                    Instrumentation 6.1.1 Table 6.1.1-1 Radioactive ,Liquid Effluent Monitoring    Iristrumentat ion APPLICABLE                  CONDITIONS MODES OR        REOUIRED  REFERENCED OTHER          CHANNELS      FROM                        ALARM/
SPECIFIED            PER  COMPENSATORY    SURVEILLANCE      TRIP FUNCTION                CONDITIONS        FUNCTION  MEASURE A.I    REQUIREMENTS  SETPOINT I. Liquid Radwaste Effluent Line                (a)              I          B          SR 6.1.1.1      (b)
Gross Radioactivity Monitor                                                        SR 6.1.1.2 SR 6..1.4 SR 6.1.1.5
: 2. Deleted
: 3. Turbine Service Water System              (d)                1        C          SR  6.1.1.1    8E-6 Gross Radioactivity Monitor                                                        SR  6.1.1.3  1 CCi/ml SR  6.1.1.4  Cs-137 SR  6.1.1.5
: 4. Standby Service Water Gross Radioactivity Monitor (d)            I  per loop C          SR SR SR 6.1.1.1 6.1.1.3 6.1.1.4 8E-6 pCi/ml Cs-137 I
SR  6.1.1.5
: 5. Liquid Radwaste Effluent    Line            (a)                1          0          SR 6.1.1.1      (c)
Flow Rate Monitor                                                                  SR 6.1.1.4 SR 6.1.1.5
: 6. Plant Discharge Blowdown  Line          (a)                I          D        SR 6.1.1.1      (c)
Flow Rate Monitor                                                                  SR 6.1.1.4 SR 6.1.1.5 (a)      When radioactive effluients are being discharged through this pathway.
(b)      The alarm/trip setpoint of the Liquid Radwaste Effluent Line Gross Radioactivity Monitor shall be set to ensure the limits of RFO 6.2.1.1 are not exceeded for each batch of radioactive 'liquid effluent released. The alarm/trip setpoint of this channel shall be determined and adjusted in accordance with the methodology and parameters described in the O0CM and plant procedure&#xfd;.
(c)        No alarm setpoints are required for these record only instruments.
(d)      When there is flow in the system identified by this function.
I Columbia Generating Station                            6.1. 1-4              Amendment 37, January 2003
 
Radioactive Gaseous Effluent Monitoring      Instrumentation 6.1.2 6.1  INSTRUMENTATION 6.1.2  Radioactive Gaseous Effluent Monitoring      Instrumentation RFO 6.1.2.1        The radioactive  gaseous effluent monitoring      instrumentation channels in Table 6.1.2-1    shall be OPERABLE.
APPLICABILITY:    In accordance with Table 6.1.2-1.
COMPENSATORY MEASURES
                -------------------- NOTE ---------------------------------------
Separate condition entry is  allowed for each channel.
CONDITION            REQUIRED COMPENSATORY MEASURE        COMPLETION TIME A. One or more required      A.1  Enter the condition            Immediately channels inoperable,            referenced in Table 6.1.2-1 for the channel.
B. As required by            B.1  Verify the Reactor            Immediately Compensatory Measure            Building Elevated A.1 and referenced in          Discharge Radiation Table 6.1.2-1.                  Monitor is OPERABLE.
OR B.2  Verify the Condition C        Immediately Compensatory Actions are being met.
(continued)
Columbia Generating Station          6.1.2-1                Amendment 35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation 6.1.2 COMPENSATORY  MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME C. As required by            C.A  Take a noble gas grab        8 hours Compensatory Measure            sample at the associated A.1 and referenced in          sample location.            AND Table 6.1.2-1.
once per 8 hours thereafter AND C.2    Analyze the grab sample    24 hours after for gamma emmiters as      completion of required by Required        Required Compensatory Measure        Compensatory C.I.                        Measure C.1 AND 30 days C.3    Restore the channel to OPERABLE status.
: 0. As required by          0.1    Estimate the flow rate      4 hours Compensatory Measure            through the associated A.1 and referenced in          pathway.                    AND Table 6.1.2-1.
Once per 4 hours thereafter AND D.2    Restore the channel to      30 days OPERABLE status.
(continued)
Columbia Generating Station          6. 1.2-2              Amendment 35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation 6.1.2 COMPENSATORY  MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE        COMPLETION TIME E. As required by Compensatory Measure E.1.1      Establish auxiliary sampling equipment to 4 hours I
A.1 and referenced in              continuously collect      AND Table 6.1.2-1.                      samples from the associated effluent        Continuously release pathway as        thereafter required in Table 6.2.2.1-1.
OR E.1.2    Collect relevant          Immediately information to provide an estimate of effluent releases.
AND E.2      Restore the channel OPERABLE status.
to  30 days I
F. Required Compensatory    F.1    Prepare and submit, in        Upon submittal Measure C.3, D.2, or            the Radioactive Effluent      of current E.2 and associated              Release Report, the          calendar year Completion Time not            reason the channel was        Radioactive met or Compensatory              riot restored to OPERABLE    Effluent Release Measure E.1.2                    status within the            Report performed.                      required completion time or the use of Compensatory Measure E.1.2.
I (continued)
Columbia Generating Station            6. 1.2-3              Amendment 45, FEB 2006
 
Radioactive Gaseous Effluent Monitoring    Instrumentation 6.1.2 COMPENSATORY MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME G. As required by          G.1  Verify the Offgas            Immediately Compensatory Measure          Treatment System A.1 and referenced            (RECHAR) is not in Table 6.1.2-1.            bypassed.
AND G.2  Install a temporary        In 4 hours portable Area Radiation Monitor and record        AND radiation levels.
Every 4 hours thereafter AND G.3  Take a grab sample and      In 8 hours analyze to verify that the noble gas gross        AND gamma activity rate is
                                    < 332 mCi/sec.            Every 24 hours thereafter AND G.4  Restore the channel to    30 days OPERABLE status.
H. Required Compensatory    H.1  Be in MODE 3.              12 hours Measure and associated Completion Time of Condition G not met.
Columbia Generating Station          6.1.2-4              Amendment 43, JAN 2005
 
Radioactive Gaseous Effluent Monitoring  Instrumentation 6.1.2 SURVEILLANCE REQUIREMENTS I
SURVEILLANCE                              FREQUENCY SR  6.1.2.1      Perform CHANNEL CHECK.                      24 hours SR  6.1.2.2      Perform SOURCE CHECK.                        24 hours SR  6.1.2.3      Perform CHANNEL CHECK.                        7 days SR  6.1.2.4      Perform SOURCE CHECK.                        31 days SR 6.1.2.5      Perform CHANNEL FUNCTIONAL TEST.            92 days SR  6.1.2.5a      Perform CHANNEL CHECK.                      Prior to mechanical vacuum pump use SR  6.1.2.6      Perform CHANNEL CALIBRATION.                18 months SR' 6.1.2.7      Perform CHANNEL CALIBRATION.                Each refueling outage Columbia Generating Station          6. 1.2-5          Amendment 43, JAN 2005
 
Radioactive Gaseous Effluent Monitoring Instrumentation 6.1.2 Table 6.1.2-1 (page I of 2)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABLE                CONDITIONS MODES OR      REQUIRED    REFERENCED OTHER      CHANNELS      FROM                          ALARM/
SPECIFIED        PER  COMPENSATORY    SURVEILLANCE          TRIP FUNCTION                CONDITIONS    FUNCTION  MEASURE A.1    REQUIREMENTS      SETPOINT
: 1. Main Condenser Offgas Post-                  (a)          I            C          SR 6.1.2.1          (c)
Treatment Radiation Monitor                                                    SR 6.1.2.2 SR 6.1.2.5 SR 6.1.2.7 I
: 2. Main Condenser Offgas Pre-                  (a)          I            G        SR 6.1.2.1          1.5 Treatment Radiation Monitor                                                      SR 6.1.2.4        times SR 6.1.2.5        steady SR 6.1.2.6        state fu I I power 2a. Mechanical Vacuum Pump Discharge            (f)            I          B        SR 6.1.2.5a          (c)
Radiation Monitor                                                                SR 6.1.2.7
: 3. Reactor Building Elevated Discharge Radiation Monitor (b) a)      Low Range Activity Monitor          (e)            1          C        SR  6.1.2.1          (c)
SR  6.1.2.4 SR  6.1.2.5 SR  6.1.2.6 b)      Iodine Sampler                    (e)            1            E        SR 6.1.2.3          N/A c)      Particulate Sampler                (e)            I            E        SR 6.1.2.3          N/A d)    Effluent System Flow Rate          (e)            1            D        SR 6.1.2.1          (d)
Monitor                                                                  SR 6.1.2.5 SR 6.1.2.6 e)    Sampler Flow Rate Monitor          (e)            1          0        SR 6.1.2.1          (d)
SR 6.1.2.5 SR 6.1.2.6 (continued)
Cal      When an Offgas Steam Jet Air Ejector is operating.
1        When building exhaust is secured collect building ambient air samples and other data to characterize building nut-leakage..
(c)      Alarm/trip setpoinnts shall be determined in accordance with ODCM methodology and set to ensure the limits of RFO 6.2.2.1, Dose Rate-Gaseous Effluents are not exceeded.
(d)      Alarm/trip setpoints shall be determined in accordance with the associated instrumentation specifications and set to ensure a representative sample is obtained.
(e)      During building exhaust system operation.
(f)      During startup and shutdown when vacuum is being maintained with a Mechanical Vacuum Pump and the Steam Sea] System is operational.
Columbia Generating Station                            6.1.2-6                    Amendment 46, OCT 2006
 
Radioactive Gaseous Effluent Monitoring Instrumentation 6.1.2 Table 6.1.2-i (page 2 of 2)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABLE                CONDITIONS MODES OR      REQUIRED  REFERENCED OTHER      CHANNELS      FROM                    ALARM/
SPECIFIED        PER  COMPENSATORY  SURVEILLANCE  TRIP FUNCTION                  CONDITIONS    FUNCTION  MEASURE A.1  REQUIREMENTS SETPOINT
: 4. Turbine Building Ventilation Exhaust Radiation Monitor (b) a)      Noble Gas Activity Monitor            (e)            I          C.      SR 6.1.2.1      (c)
Low Range                                                                SR 6.1.2.4 SR 6.1.2.5 SR 6.1.2.6 b)      Iodine Sampler                      (e)            I          E        SR 6.1.2.3    N/A c)      Particulate  Sampler                (e)            1          E        SR 6.1.2.3    N/A d)      Effluent System Flow Rate            (e)            1          D        SR 6.1.2.1    (d)
Monitor                                                                  SR 6.1.2.5 SR 6.1.2.6 e)      Sampler Flow Rate Monitor            (e)            1          D        SR 6.1.2.1    (d)
SR 6.1.2.5 SR 6.1.2.6
: 5. Radwaste Building Ventilation Exhaust Radiation Monitor (b) a)      Noble Gas Activity Monitor          (e)            1          C        SR 6.1.2.1    (c)
Low Range                                                                SR 6.1.2.4 SR 6.1.2.5 SR 6.1.2.6 b)      Iodine Sampler                      (e)            I          E        SR 6.1.2.3    N/A c)      Particulate  Sampler                (e)            I          E        SR 6.1.2.3    N/A d)      Effluent System Flow Rate            (e)            1          D        SR 6.1.2.1    (d)
Monitor                                                                  SR 6.1.2.5 SR 6.1.2.6 e)      Sampler Flow Rate Monitor            (e)            1          D        SR 6.1.2.1    (d)
SR 6.1.2.5 SR 6.1.2.6 (b)      When building exhaust is secured collect building ambient air samples and other data to characterize building out-leakage..
(c)      Alarm/trip setpoints shall be determined in accordance with OCM methodology and set to ensure the limits of RFO 6.2.2.1, Dose Rate-Gaseous Effluents are not exceeded..
I (d)      Alarm/trip setpoints shall be determined in accordance with the associated instrumentation specifications and set to ensure a representative sample is obtained..
(e)      During building exhaust system operation.
Columbia Generating Station                              6.1.2-7                      Amendment 44, May 2005
 
Liquid Concentration 6.2.1.I 6.2  RADIOACTIVE    EFFLUENTS 6.2.1    Liquid Effluents 6.2.1.1    Liquid Concentration RFO 6.2.1.1        The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 3-1) shall be limited to:
: a. Ten times the concentration specified in Table 2, Column 2 of Appendix B to 10 CFR Part 20 for radionuclides other than dissolved or entrained noble gases; and
: b. 2 X 1O-4,Ci/ml total activity concentration for all dissolved or entrained noble gases.
APPLICABILITY:      At all times.
COMPENSATORY MEASURES CONDITION              REQUIRED COMPENSATORY MEASURE    COMPLETION TIME A. Concentration of            A.1  Initiate action to        Immediately radioactive material              restore concentration to released to                      within limits.
UNRESTRICTED AREAS not within limits.
Columbia Generating Station            6.2. 1. 1-1          Amendment 35, MAR 2002
 
Liquid Concentration 6.2.1.1 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 6.2.1.1.1      The radioactivity content of each batch      Prior to each of radioactive liquid waste shall be          batch release determined prior to release by sampling and analysis in accordance with Table 6.2.1.1-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of RFO 6.2.1.1.
SR  6.2.1.1.2      Post-release analysis of samples              In accordance composited from batch releases shall be      with Table performed in accordance with Table            6.2.1.1-1 6.2.1.1-1. The results of the post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the limits of RFO 6.2.1.1.
Columbia Generating Station          6.2.1. 1-2            Amendment 35, MAR 2002
 
Liquid Concentration 6.2.1.1 Table 6.2.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program MINIMUM                      SAMPLE LOWER SAMPLE        TYPE OF          LIMIT OF LIQUID RELEASE        SAMPLE      ANALYSIS        ACTIVITY        DETECTION TYPE          FREQUENCY    FREQUENCY        ANALYSIS            (LLD) 7 Batch waste tank    Prior to      Prior to        Principal          5 X 10-release            release of    release of    Gamma                  gCi /ml each batch    each batch    Emitters 1-131              1 X 10-6 MCi /ml Dissolved and        1 X 10-5 entrained              4Ci/ml gases (gamma emitters)
Batch waste tank    Prior to    31 days        Tritium              1 X 10-5 release            release of  composite                              gCi /ml each batch Gross Alpha          1 X 10-7 4Ci /ml 6
Batch waste tank  Prior to      92 days        Fe-55                1 X 10-release            release of  composite                            'Ci/ml each batch Sr-89                5 X 10-8 1 CCi /ml Sr-90                5 X 10-8 4Ci /ml Columbia Generating Station        6.2.1.1-3              Amendment 35, MAR 2002
 
Liquid Dose 6.2.1.2 6.2 RADIOACTIVE EFFLUENTS 6.2.1  Liquid Effluents 6.2.1.2  Liquid Dose RFO 6.2.1.2      The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see ODCM Figure 3-1) shall be limited to:
: a.    < 1.5 mrem to the whole body and during any calendar quarter; and
                                                          < 5 mrem to any organ    I
: b.    < 3 mrem to the whole body and < 10 mrem to any organ during any calendar year.
(contnued APPLICABILITY:  When radioactive liquid effluents are released.
(continued)
Columbia Generating Station          6.2. 1.2-1        Amendment 38, APRIL 2003
 
Liquid Dose 6.2.1.2 COMPENSATORY MEASURES
----------------------------------- NOTE------------------------------------
Enter applicable Conditions and Required Compensatory Measures of OOCM 6.2.3, "Total Dose," when the calculated dose from liquid effluents exceeds twice the limits of RFO 6.2.1.2.
CONDITION            REQUIRED COMPENSATORY MEASURE    COMPLETION TIME A. Calculated dose from      A.1  Submit a Special Report    30 days the release of                  to the NRC that radioactive materials          identifies causes for in liquid effluents to          exceeding limits, Unrestricted Areas not          corrective actions taken within limits,                  to reduce releases, corrective actions to assure that subsequent releases will be in compliance with the required limits, results of radiological analyses of drinking water source, and the radiological impact on potentially affected drinking water supplies with regard to 40 CFR 141.
SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.1.2.1    Verify the cumulative dose contributions      31 days from radioactive liquid effluents for the current calendar quarter and current calendar year are within limits in accordance with the methodology and parameters in the ODCM.
Columbia Generating Station        6.2.1.2-2              Amendment 35,  MAR 2002
 
Liquid Waste Management System 6.2.1.3 6.2  RADIOACTIVE EFFLUENFS 6.2.1  Liquid Effluents 6.2.1.3    Liquid Waste Management System RFO 6.2.1.3        The Liquid Waste Management System shall be OPERABLE.
Appropriate portions of the system shall be used to reduce the radioactive materials in liquid waste prior to discharge when the projected doses due to the liquid.effluent, from the reactor unit to UNRESTRICTED AREAS, would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.
APPLICABILITY:    At all times.
COMPENSATORY MEASURES CONDITION            REQUIRED COMPENSATORY  MEASURE  COMPLETION TIME A. Liquid waste being        A.1    Submit a Special Report    30 days discharged without              to the NRC that includes treatment when RFO              explanation of why the 6.2.1.3 limits are              Liquid Waste Management exceeded.                      System was not OPERABLE or why liquid radwaste was being discharged without treatment. The Special Report shall identify any required inoperable equipment and the reasons for the inoperability, the corrective actions taken to restore the required inoperable equipment to OPERABLE status, and the corrective actions to prevent recurrence.
Columbia Generating Station          6.2. 1.3-1          Amendment 38, APRIL 2003
 
Liquid Waste Management System 6.2.1.3 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.1.3.1    Dose due to liquid releases to                Prior to each UNRESTRICTED AREAS shall. be projected in    batch release accordance with the methodology and parameters in the ODCM.
SR  6.2.1.3.2      The installed Liquid Waste Management        See RFOs System shall be demonstrated OPERABLE by      6.2.1.1 and meeting RFOs 6.2.1.1 and 6.2.1.2.            6.2.1.2 Columbia Generating Station          6.2.1.3-2            Amendment 35, MAR 2002
 
Dose Rate 6.2.2.1 6.2  RADIOACTIVE EFFLUENTS 6.2.2    Gaseous Effluents 6.2.2.1    Dose Rate RFO 6.2.2.21        The dose rate at and beyond the SITE BOUNDARY (See Figure 3-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following:
: a. For nobles gases, less than or equal to 500 mrem per year to the whole body and less than or equal to 3000 mrem per year to the skin; and I
: b. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives >8 days, less than or equal to 1500 mrem per year to any organ.
APPLICABILITY:      At all times.
COMPENSATORY MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE    COMPLETION TIME A. Dose rate from the          A.1  Initiate action to        Immediately release of radioactive            restore dose rate to materials in gaseous              within limits.
effluents at and beyond the SITE BOUNDARY not within limits.
Columbia Generating Station            6.2.2.1-1          Amendment 38, APRIL 2003
 
Dose Rate 6.2.2.1 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.2.1.1      Verify the  dose rate due to noble gases      In accordance in gaseous effluents is within limits in    with fable accordance with methodology and              6.2.2.1-1 parameters  in the ODCM.
SR  6.2.2.1.2    Verify dose rate due to iodine-131,          In accordance iodine-133, tritium, and all                with Table radionuclides in particulate form with      6.2.2.1-1 half lives >8 days in gaseous effluents is within limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses.
Columbia Generating Station          6.2.2.1-2            Amepdment 35, MAR-2002
 
Dose Rate 6.2.2.1 Table 6.2.2.1-1 Part 1 (Page 1 of 5)
Radioactive Gaseous Waste Sampling and Analysis Program Requirements Applicable At All Times SAMPLE TYPE            SAMPLE        MINIMUM  TYPE OF      SAMPLE. LOWER FREQUENCY        ANALYSIS  ACTIVITY          LIMIT OF FREQUENCY  ANALYSIS        DETECTION LLD)
Primary Containment          31 days        31 days    Tritium          I X 10.6 Tritium Grab Sample(a)                                                    pCi/ml Reactor Building              7 days          7 days    Tritium          1X 10-6 Elevated Discharge Vent                                                    &#xfd;iCi /ml Tritium Grab Sample Reactor Building              31 Days        31 Days  Principal          I X 10i Elevated Discharge                                        Gamma            UCi/ml Noble Gas Grab Sample                                  Emitters Turbine and Radwaste          31 Days        31 Days  Principal          I X 10i Building Vent Noble Gas                                    Gamma          pCi/ml and Tritium Grab                                        Emitters Samples Tritium          I X 10.6 ACi /ml Charcoal Filter Sample      Continuous        7 days      1-131        1 X 10-12 (All Building Vents)            (b)(c)                                    /ICi /ml 1-133        1 X 10.10 PCi/ml Particulate Filter          Continuous        7 days  Princi pal      1 X 10-11 Sample (All Building          (b)(c)                    Gamma          mCi/ml Vents)                                                  Emitters (continued)
(a)      Not required when Primary Containment is not required.
(b)    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 RFOs 6.2.2.1, 6.2.2.2, and 6.2.2.3.
(c)    Analyses shall be completed within 48 hours after changing, or after removal from the sampler.
Columbia Generating Station              6.2.2. 1-3          'Amendment 35,    MAR 2002
 
Dose Rate 6.2.2.1 Table 6.2.2.1-1 Part 1 (Page 2 of 5)
Radioactive Gaseous Waste Sampling and Analysis Program Requirements Applicable At All Times SAMPLE TYPE            SAMPLE        MINIMUM    TYPE, OF    SAMPLE LOWER FREQUENCY      ANALYSIS  ACTIVITY        LIMIT OF FREQUENCY  ANALYSIS      DETECTION (LLD)
Composite of                Continuous      31 days  Gross Alpha        I X i0-1 Particulate Filter              (b)                                      kiCi /ml Samples (All Building Vents)
Composite of                Continuous      92 days    Sr-89,          1 X i0-1 Particulate Filter              (b)                      Sr-90            pCi/ml Samples (All Building Vents)
Noble Gas (All Building    Continuous                Gross Beta        I X 10.6 Vents)                                                or Gamma          1uCi/ml (Xe-133 Equivalent)
(b)    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 RFOs 6.2.2.1, 6.2.2.2, and 6.2.2.3.
Columbia Generating Station            6.2.2. 1-4            Amendment 35,    MAR 2002
 
Dose Rate 6.2.2.1 Table 6.2.2.1-1 Part 2 (Page 3 of 5)
Radioactive Gaseous Waste Sampling and Analysis Program Requirements Applicable During PURGE and VENT SAMPLE TYPE            SAMPLE        MINIMUM  TYPE OF    SAMPLE LOWER FREQUENCY      ANALYSIS  ACTIVITY        LIMIT OF FREQUENCY  ANALYSIS      DETECTION (LLD)
Primary Containment          8 hours        8 hours  Principal      1    10Xi Purge and Vent Noble        prior to      prior to    Gamma          kCi/ml Gas, Iodine, and          each PURGE    each PURGE  Emitters Particulate Grab            and VENT      and VENT Sample(d)(e)
(d)    Sampling and analysis is not required for a Primary Containment VENT when the VENT path is through the Standby Gas Treatment System via the two-inch bypass line and when containment noble gas monitoring instrumentation indicates less than the alarm setpoint.
(e)    If the Reactor Building Elevated Discharge Low Range Radiation Monitor is not OPERABLE, sampling and analysis of the containment shall be completed at least once per 12 hours during VENT or PURGE.
Columbia Generating Station          6.2'.2.1-5          Amendment 35, MAR 2002
 
Dose Rate 6.2.2.1 Table 6.2.2.1-1 Part 3 (Page 4 of 5)
Radioactive Gaseous Waste Sampling and Analysis Program Requirements During Startup and Shutdown SAMPLE TYPE                  SAMPLE        MINIMUM        TYPE OF      SAMPLE LOWER FREQUENCY      ANALYSIS      ACTIVITY          LIMIT OF FREQUENCY      ANALYSIS        DETECTION (LLD)
Primary Containment and              Once          Once        Principal        I X 10-1 Reactor Building                    following      following        Gamma            UCi/ml Elevated Discharge                startup or    startup or      Emitters Noble Gas, Iodine, and              shutdown      shutdown Particulate Grab Sample Charcoal Filter Sample            Continuous          (f)          1-131          1 'X 10.2 (All Building Vents)                    (b)                                        4Ci/ml 1-133          1 X 1010 pCi /ml Particulate Filter                Continuous          (f)      Principal        1 X 10-"
Sample (All Building                    (b)                        Gamma          UCi/ml Vents)                                                            Emitters (continued)
(b)    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 RFOs 6.2.2.1, 6.2.2.2, and 6.2.2.3.
(f)    Filters shall be removed from the continuous sampler at least once per 24 hours for at least 7 days following each shutdown or startup, and analyses shall be completed within 48 hours of filter removal.        When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10.
Requirement  (f)    does not apply if:
(1) a.      Analysis shows that the DOSE EQUIVALENT 1-131 concentration      in the primary coolant is less than or equal to L.OE-03 ACi/cc.
Or
: b. When the DOSE EQUIVALENT 1-131 concentration in the primary coolant is greater than 1.OE-03 pCi/cc, but has rot increased more than a factor of 3; And (2)  a. When any building vent noble gas monitor is less than or equal to 2.0% of the setpoint determined in accordance with ODCM Section 3.4.
Or
: b. When any building vent noble gas monitor is greater than 2.0% of its setpoint, but shows that effluent activity has not increased by more than a factor of 3.
Columbia Generating Station                    6.2.2.1-6                Amendment 35, MAR 2002
 
Dose Rate 6.2.2.1 Table 6.2.2.1-1 Part 4 (Page 5 of 5)
Radioactive Gaseous Waste Sampling and Analysis Program Requirements During Reactor Power Changes SAMPLE TYPE                  SAMPLE        MINIMUM          TYPE OF          SAMPLE LOWER FREQUENCY      ANALYSIS          ACTIVITY            LIMIT OF FREQUENCY        ANALYSIS          DETECTION (LLD)
Primary Containment and                  (g)            (g)          Principal            1 X 10i1 Reactor Building                                                        Gamma              pCi/ml Elevated Discharge                                                    Emitters Noble Gas, Iodine, and Particulate Grab Sample Charcoal Filter Sample              Continuous          (h)            1-131            1 X 10-12 (All Building Vents)                    (b)                                                4Ci/ml 1-133            1 X 10I-4Ci/ml Particulate Filter                  Continuous          (h)          Principal            1 X 10-11 Sample (All Building                      (b)                            Gamma              4Ci/ml Vents)                                                                  Emitters (b)      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 RFOs 6.2.2.1, 6.2.2.2, and 6.2.2.3.
(g)      Grab sampling and analysis shall be performed following a THERMAL POWER change exceeding 15%
of RATED THERMAL POWER within a P-hour period when the noble gas release rate, as indicated by the main condenser offgas pretreatment monitor, is greater than 15,000 jCi/sec.
(h)      Filters shall be removed from the continuous samplers at least once per 24 hours for at least 7 days following each THERMAL POWER change exceeding 15% of RATED THERMAL POWER in one hour when the noble gas release rate, as indicated by the main condenser offgas pretreatment monitor, is greater than 15,000 pCi/sec.      Analyses shall be completed within 48 hours of filter    change. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10.
Requirement (h)    does not apply if:
(1)    a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration  in  the primary coolant is less than or equal to I.OE-03 uCi/cc.
Or
: b. When the DOSE EQUIVALENT 1-131 concentration in the primary coolant    is greater than 1.OE-03 pCi/cc, but has not increased more than a factor of 3; And (2)    a. When any building vent noble gas monitor is less than or equal  to 2.0% of the setpoint determined in accordance with OOCM Section 3.4.
Or
: b. When any building vent noble gas monitor is greater than 2.0% of its setpoint, but shows that effluent activity has not increased by more than a factor of 3.
Columbia Generating Station                      6.2.2.1-7                    Amendment 35, MAR 2002
 
Dose  - Noble gases 6.2.2.2 6.2    RADIOACTIVE EFFLUENTS 6.2.2    Gaseous Effluents 6.2.2.2    Dose - Noble Gases RFO 6.2.2.2        The air dose at and beyond the SITE BOUNDARY from noble gases in gaseous effluents from the site shall be limited to the following:
: a.  <  5 mrads gamma radiation and <= 10 mrads beta radiation during any calendar quarter; and
: b.    < 10 mrads gamma radiation and <= 20 mrads beta radiation during any calendar year.
APPLICABILITY:      At all times.
COMPENSATORY MEASURES
------------------------------------    NOTE------------------------------------
Enter applicable Conditions and Required Compensatory Measures of 6.2.3, Total Dose, when gaseous effluent (noble gas) dose exceeds twice the limits of RFO 6.2.2.2.
CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A. Calculated air dose        A.1  Submit a Special Report      30 days from radioactive noble          to the NRC that gases in gaseous                identifies causes for effluents in the                exceeding the limits, UNRESTRICTED AREA not            corrective actions taken within limits,                  to reduce releases, and corrective actions to assure that subsequent releases are within limits.
Columbia Generating Station            6.2.2.2-1              Amendment 35, MAR 2002
 
Dose - Noble gases 6.2.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.2.2.1    Verify the cumulative dose contributions    31 days from noble gases in gaseous effluents for the current calendar quarter and current calendar year are within limits in'.  .
accordance with the methodology and parameters in the ODCM.
I.' ,
Columbia Generating Station        6.2.2.2-2            Amendment 35, MAR 2002
 
Dose                    -  1-131,  1-133,  Tritium,    and Radionuclides in Particulate Form 6.2.2.3 6.2                          RADIOACTIVE EFFLUENTS 6.2.2                            Gaseous Effluents 6.2.2.3                                  Dose 131,                                1-133,  Tritium,  and Radionuclide-s in Particulate Form RFO 6.2.2.3                                                                    The dose to any organ of a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to the UNRESTRICTED AREA shall be limited to the following:
: a.    <  7.5 mrems during any calendar quarter; and
: b.    <  15 mrems during any calendar year.
APPLICABILITY:                                                                At all    times.
COMPENSATORY MEASURES
                                                                    -------------------- NOTE ---------------------------------------
Enter applicable Conditions and Required Compensatory Measures of 6.2.3, Total Dose, when gaseous effluent dose from 1-131, 1-133, Tritium, and Radionuclides in Particulate Form exceeds twice the limits of RFO 6.2.2.3.
CONDITION                                            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A.                Calculated dose from                                                      A.1    Submit a Special Report      30 days the release of iodine-                                                          to the NRC that 131, iodine-133,                                                                identifies causes for tritium, and                                                                    exceeding the limits, radionuclides in                                                                corrective actions taken particulate form with                                                          to reduce releases, and half-lives > 8 days,                                                            corrective actions to in gaseous effluents                                                            assure that subsequent in the UNRESTRICTED                                                            releases are within AREA not within                                                                  limits.
limits.
Columbia Generating Station                                                                            6.2.2.3-1              Amendment 35,  MAR 2002
 
Dose  -  1-131,  1-133,  Tritium,  and Radionuclides in  Particulate Form
: 6. 2. 2.3 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.2.3.1. Verify the cumulative dose contributions        31 days from iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives > 8 days, in gaseous effluents for the current calendar quarter and current calendar year are within limits in accordance with the methodology and parameters in the ODCM.
Columbia Generating Station          6.2.2.3-2              Amendment 35,  MAR 2002
 
Gaseous Offgas Treatment System 6.2.2.4 6.2  RADIOACTIVE EFFLUENTS 6.2.2    Gaseous Effluents 6.2.2.4    Gaseous Offgas Treatment System RFO 6.2.2.4        The Gaseous Offgas Treatment System shall be in operation in either:
: a. The normal mode; or
: b. The charcoal bypass mode, provided the offgas post-treatment radiation monitor is OPERABLE as specified in Table 6.1.2-1.
APPLICABILITY:      Whenever the main condenser steam jet air      ejector  system is in operation.
COMPENSATORY    MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A. Gaseous Offgas              A.1    Place Gaseous Offgas        7 days Treatment System not              Treatment System in the in the normal mode of              normal mode of operation.                        operation.
B. Required                  B.1  Submit a Special Report      30 days Compensatory Measure            to the NRC that A.1 and associated              identifies the required Completion Time not              inoperable equipment and met.                            the reasons for the inoperability, corrective actions taken to restore the required inoperable equipment to OPERABLE status, and the corrective actions to prevent recurrence.
Columbia Generating Station            6.2.2.4-1                Amendment 35, MAR 2002
 
Gaseous Offgas Treatment System 6.2.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR  6.2.2.4.1    Verify Gaseous Offgas Treatment System in    7 days
                  - operation in either:
a) The normal mode; or b) The charcoal bypass mode, provided the offgas post-treatment radiation monitor is OPERABLE.
Columbia Generating Station          6.2.2.4-2            Amendment 35, MAR 2002
 
VENTILATION EXHAUST TREATMENT SYSTEM 6.2.2.5 I
6.2 RADIOACTIVE EFFLUENTS 6.2.2  Gaseous Effluents 6.2.2.5  VENTILATION EXHAUST TREATMENT SYSTEM I
RFO 6.2.2.5        The appropriate portions of the VENTILATION EXHAUST I
TREATMENT SYSTEM shall be OPERABLE. The system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when-the projected doses due to gaseous effluent, from the reactor unit to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREAS when averaged over 31 days, would exceed 0.3 mrem to any organ in a 31 day period.
APPLICABILITY:    At all times.
COMPENSATORY  MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A. Gaseous waste being      A.1  Submit a Special Report      30 days discharged without              to the NRC that VENTILATION EXHAUST TREATMENT SYSTEM treatment when identifies inoperable equipment or subsystems and the reasons for I
projected organ dose            inoperability, the exceeds 0.3 mrem.              corrective actions taken to restore the inoperable equipment to OPERABLE status, and the corrective actions to prevent recurrence.
Columbia Generating Station          6.2.2.5-1              Amendment 47, AUG 2008
 
VENTILATION EXHAUST  FREATMENT SYSTEM 6.2.2.5 I
SURVEILLANCE  REQUIREMENTS SURVEILLANCE                                FREQUENCY SR  6.2.2.5.1    Determine  the projected organ doses due      31 days to gaseous  releases to a MEMBER OF THE PUBLIC in  the UNRESTRICTED AREA in accordance  with the methodology and parameters  in the ODCM.
SR  6.2.2.5.2      The installed VENTILATION EXHAUST              Once per 92 TREATMENT SYSTEM shall be demonstrated          days OPERABLE by complying with RFOs 6.2.2.1 and 6.2.2.3.
Columbia Generating Station            6.2.2.5-2            Amendment 47, AUG 2008
 
VENTING or  PURGING 6.2.2.6 6.2 RADIOACTIVE EFFLUENTS 6.2.2  Gaseous  Effluents 6.2.2.6  VENTING or PURGING RFO 6.2.2.6        VENTING or PURGING of the primary containment    shall be:
: a. Through one functional-for-filtration Standby Gas Treatment (SGT) System train during MODES I, 2,-or 3 provided the other train is OPERABLE during the first 24 hours of any VENTING or PURGING operation; or
: b. Through one or two functional-for-filtration SGT System train(s) when deinerting the containment in MODE 4 during the first 24 hours of any VENTING or PURGING operation; or
: c. Through the Primary Containment VENT and PURGE System when not using SGT following the first 24 hours of any VENTING and PURGING operation.
APPLICABILITY:. MODES 1, 2, and 3 when VENTING or PURGING and when deinerting the containment. Not applicable during containment depressurization following 10 CFR 50, Appendix J, Type A testing.
COMPENSATORY MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A. Requirements of          A.1  Suspend VENTING and          Immediately 6.2.2.6 not met.                PURGING of the containment.
Columbia Generating Station          6.2.2.6-1              Amendment 35, MAR 2002
 
VENTING or  PURGING 6.2.2.6 SURVEILLANCE  REQUIREMENTS SURVEILLANCE                            FREQUENCY SR  6.2.2.6.1    Verify the containment is aligned for      Within 4 hours VENTING or PURGING through the SGT system  prior to the or the Primary Containment VENT and PURGE  start and once System.                                    per 12 hours during PURGING or VENTING Columbia Generating Station          6.2 .2.6-2        Amendment 35, MAR 2002
 
Total  Dose 6.2.3 6.2 RADIOACTIVE EFFLUENTS 6.2.3  Total Dose RFO 6.2.3          The annual dose or dose commitment to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA beyond the SITE BOUNDARY due to releases of radioactivity and radiation from uranium fuel I
cycle sources shall be limited to:
: a.      25 mrem to the whole body or any organ except the thyroid; and I
: b.  <  75 mrem to the thyroid.
APPLICABILITY:      At all times.
COMPENSATORY  MEASURES CONDITION              REQUIRED COMPENSATORY MEASURE      COMPLETION  TIME A. Calculated dose to any    A.1 --------- NOTE-----------      Immediately MEMBER OF THE PUBLIC              Calculations shall in the UNRESTRICTED              include direct radiation AREA exceeds twice the            contributions from the limits of 6.2.1.2 or              reactor, outside storage 6.2.2.2 or 6.2.2.3.              tanks, and the Independent Spent Fuel Storage Installation (ISFSI).
Calculate the dose or dose commitment to a MEMBER OF -THE PUBLIC from all uranium fuel cycle sources to determine if specified limits of RFO 6.2.3 were exceeded.
(continued)
Columbia Generating Station              6.2.3-1            Amendment 38, APRIL 2003
 
Total Dose 6.2.3 COMPENSATORY  MEASURES CONDITION          REQUIRED COMPENSATORY MEASURE    COMPLETION TIME B. Calculated dose or    B.1    Submit a Special Report    30 days dose commitment              to the NRC pursuant to exceeds specified            10 CFR 50, Appendix I, limits of RFO 6.2.3.          Section IV.A, that includes corrective actions taken to prevent recurrence, the schedule for achieving conformance with required limits, an analysis that estimates the radiation exposure to a MEMBER OF THE PUBLIC from uranium fuel cycle sources for the calendar year, descriptions of the levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.
AND (continued)
Columbia Generating Station      6.2 .3-2              Amendment 35, MAR 2002
 
Total Dose 6.2.3 COMPENSATORY MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE      COMPLETION TIME B.    (continued)              B.2 --------- NOTE-----------      30 days
: 1. Only applicable if the release condition resulting in violation of 40 CFR 190 has not been corrected.
: 2. Special Report submitted is considered a timely request and a variance is granted until NRC action on the request is complete.
Submit a request for a variance in accordance with 40 CFR 190 in the Special Report to the NRC.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR  6.2.3.1        Cumulative dose contributions from liquid      31 days and gaseous effluents shall be determined in accordance with SRs 6.2.1.2.1, 6.2.2.2.1, and 6.2.2.3.1.
Columbia Generating Station            6.2.3-3                Amendment 35, MAR 2002
 
Radiological  Environmental Monitoring Program 6.3.1 6.3  RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.1  Radiological  Environmental  Monitoring Program RFO 6.3.1          The Radiological  Environmental  Monitoring Program shall be as follows:
: a. Radiological Environmental Monitoring samples shall be collected at locations and analyzed as specified in Table 6.3.1-1;
: b. The level of radioactivity as the result of plant effluents for a single radionuclide in each environmental sampling medium at a required location shall be less than the limits specified in Table 6.3.1-2, when averaged over the calendar quarter;
: c. The total level of radioactivity in multiple detected radionuclides as the result of plant effluents in each environmental sampling medium at a required location shall be less than the limit specified in Table 6.3.1-2, when averaged over the calendar quarter; and
: d. The potential annual dose to the MEMBER OF THE PUBLIC from all radionuclides other than those in Table 6.3.1-2 in each environmental sampling medium at a required location shall be less than the calendar year limits of RFOs 6.2.1.2, 6.2.2.2, and 6.2.2.3.
APPLICABILITY:    At all  times.
(continued)
Columbia Generating Station            6.3.1-1              Amendment 35, MAR 2002
 
Radiological Environmental Monitoring Program 6.3.1 COMPENSATORY MEASURES          _________________                        _________
CONDITION            REQUIRED COMPENSATORY MEASURE          COMPLETION TIME A. one or more samples        A.1        Prepare and submit, in      Upon submittal not collected or                      the Annual Radiological    of current analyzed as specified                Environmental Operating    Annual in Table 6.3.1-1.                    Report, a description        Radi-ological for not conducting the      Environmental Radiological                Operating Report Environmental Monitoring sampling and analysis requirements as required and the corrective actions to prevent recurrence.
B.--------- NOTE ------------ 8.1      Identify locations for        30 days Separate Condition                obtaining replacement entry is allowed for              samples and replace, in each sample location,              the Radiological
                      --- ---            Environmental Monitoring Program, the location(s)
One or more sample                from which samples are locations required by              unavailable with the new Table 6.3.1-1 with                location(s).
milk or fresh leafy vegetable samples        AND unavailable.
B.2 Prepare and submit, in            Upon submittal the Radiological              of current Environmental Operating        Radiological Report, the cause of the      Environmental unavailability of              Operating Report samples, the new locations for obtaining replacement samples, and the revised figure(s) and table(s) for the ODCM reflecting the new l~ocat ions.
(continued)
Columbia Generating Station                6.3. 1-2 6312Aedet3,MR20      Amendment 35, MAR 2002
 
Radiological  Environmental Monitoring Program 6.3.1 COMPENSATORY MEASURES CONDITION        IREQUIRED COMPENSATORY MEASURE      COMPLETION TIME C. - -------NOTE---------    C.I  .Submit a Special Report      30 days Separate Condition          to the NRC which entry is allowed for        includes the cause(s) each sample location.        for exceeding the limit(s) and the corrective actions to One or more sample            reduce radioactive locations with the            effluents so that the level of radioactivity        potential annual dose to for one radionuclide          a MEMBER OF THE PUBLIC as the result of plant        is less than the effluents in an              calendar year reporting environmental sampling        limits of RFOs 6.2.1.2, medium not within the        6.2.2.2, and 6.2.2.3.
limits of Table 6.3.1-2 when averaged over the calendar quarter.
OR One or more sample locations with the total level of radioactivity for more than one radionuclide as a result of plant effluents in an environmental sampling medium not within the limits of Table 6.3.1-2 when averaged over the calendar quarter.
(continued)
Columbia Generating Station          6.3.1-3              Amendment 35, MAR 2002
 
Radiological            Environmental Monitoring Program 6.3.1 COMPENSATORY MEASURES CONDITION          REQUIRED COMPENSATORY MEASURE                COMPLETION TIME D. One or more sample    D.1 --------- NOTE----------                30 days locations with the            Only required if the potential annual dose        radionuclides are the to the MEMBER OF THE          result of plant PUBLIC from                  effluents.
radionuclides other          ------------------------
than those in Table 6.3.1-2 not within          Submit a Special Report limits.                      to the NRC which includes the cause(s) for exceeding the limit(s) and the corrective actions to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of RFOs 6.2.1.2, 6.2.2.2, and 6.2.2.3.
AND D.2 ---------- NOTE---------                Upon submittal Only required if the                  of the current radionuclides are not                  calendar year the result of plant                    Annual effluents.                            Radiological
                                  ------------------------              Environmental Operating Report Describe the condition in the Annual Radiological Environmental Operating Report.
Columbia Generating Station        6.3.1-4                          Amendment 35, MAR 2002
 
Radiological Environmental Monitoring Program 6.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR  6.3.1.1      Verify radiological environmental            In accordance monitoring samples collected at the          with Table locations and analyzed as specified in      6.3.1-1 Table 6.3.1-1 are within limits.
Detection capabilities for the analysis are specified in Table 6.3.1-3.
Columbia Generating Station          6.3.1-5            Amendment 35, MAR 2002
 
Radiological    Environmental Monitoring Program 6.3.1 TABLE 6.3.1-1 (page I of 6)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SA\MPLES          SAMPLING AND          TYPE AND AND/OR SAMPLE                  AND SAMPLE LOCGATON.(a                  COLLECTION        FREQUENCY OF FREQUENCY          ANALYSIS I.      DIRECT            Thirty-four  routine monitoring          Quarterly.          Gamma dose RADIATION(b)      stations either with two or more                                quarterly.
dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:
An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY.
An outer ring of stations, one in each of the meteorological sectors of NE, ENE, E, ESE, SE in the six to nine km range from the site, and one in each of the meteorological sectors of N, NNE, SSE, S, SSW in the nine to twelve km range from the site.
The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, arid one or two areas to serve as control stations.
Additional monitoring stations for the Independent Spent Fuel Storage Installation (ISFSI).
(continued)
(a)        Specific  parameters of distance and direction sector relative to the reactor are provided for each  sample location in Table 5.1.        Refer to Radiological Assessment Branch Technical Position,  Revision 1, November    L979. Deviations are permitted from the required sampling schedule  if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal  unavailability, and malfunction of automatic sampling equipment.
(b)      One or more instruments, such as a 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 thermoluminescent dosimeter (TLD) is considered to be one phosphor card with multiple readout areas; a phosphor card in a packet is considered to be equivalent to two or more dosimeters.          Film badges shall not be used as dosimeters for measuring direct radiation.        (The number of direct radiation monitoring stations may be reduced according to geographical limitations.          The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.)
Columbia Generating Station                          6.3.1-6                    Amendment 35,    MAR 2002
 
Radiological      Environmental    Monitoring Program 6.3.1 TABLE 6.3.1-1 (page 2 of 6)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SAMPLES        SAMPLING AND          TYPE AND AND/OR SAMPLE                  AND SAMPLE LOCATIONSda)              COLLECTION        FREQUENCY OF FREQUENCY            ANALYSIS
: 2.      AIRBORNE Radioiodine        Samples from six locations:            Continuous sampler    Radioiodine arid                                                      operation with        Canister:
Particulates      One sample from close to the SITE      sample collection      1-131 analysis BOUNDARY location, having a high        weekly, or more      weekly.
calculated annual average ground-      frequently if level D/Q.                            required by dust      Particulate loading.              Sampler:
Three samples from close to the                              Gross beta three Columbia River locations                                radioactivity having the highest calculated D/0.                            analysis following One sample from the vicinity of a                            filter community having the highest                                change,(c) calcula-ted annual average ground-level D/0.
One sample from a control location,                          Gamma isotopic as for example 30-50 km distant and                          analysis  d) in the least prevalent wind                                  of composite direction.                                                  (by location) quarterly.
(continued)
(a)      Specific  parameters of distance and direction sector relative to the reactor are provided for each  sample location in Table 5.1.      Refer to 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, and malfunction of automatic sampling equipment.
(c)      Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon arid thorium daughter decay.        If gross beta activity  in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
(d)      Gamma isotopic analysis means the identification and quantification of gamma-emitting radionucl ides that may be attributable to the effluents from the facility.'
Columbia Generating Station                        6.3. 1-7                    Amendment 35,      MAR 2002
 
Radiological    Environmental      Monitoring Program 6.3.1 TABLE 6.3.1-I (page 3 of 6)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SAMPLES          SAMPLING AND          TYPE AND AND/OR SAMPLE                  AND SAMPLE LOCATIONSWa)              COLLECTION        FREQUENCY OF FREQUENCY          ANALYSIS
: 3.      WATERBORNE One sample upstream                    Composite sample      Gamma isotopic
: a. Surface(e)      One sample downstream                  over on? month        analysis period.fT            monthly.
Composite for tritium analysis quarterly.
Gamma Samples from one or two source&#xfd;        Quarterly.          isotopic(d)
: b. Ground          only if likely to be affected.    )                        and tritium analysis quarterly.
(continued)
(a)        Specific  parameters of distance and direction sector relative to the reactor are provided for each  sample location in Table 5.1.      Refer to 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, and malfunction of automatic sampling equipment.
(d)      Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(e)      The "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near the mixing zone.
(f)      A composite sample is one in which the quantity (aliquot) of liquid 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 (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtai~ning a representative sample.
(g)        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.
Columbia Generating Station                          6.3.1-8                    Amendment 35,    MAR 2002
 
Radiological      Environmental      Monitoring Program 6.3.1 FABLE 6.3.1-I (page 4 of 6)
RAOIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SAMPLES          SAMPLING AND            TYPE AND AND/OR SAMPLE                  AND SAMPLE LOCATIONS(a)                COLLECTION          FREQUENCY OF FREQUENCY            ANALYSIS
: 3.      WATERBORNE (continued)
One sample of each of one to three      Composite sample        1-131 analysis
: c. Drinking        of the nearest water supplies that      over twoweek          on each could be affected by its discharge.      period    when        composite when 1-131 analysis is      the dose performed; monthly    calculated for composite              the otherwise,              consumption of the water is greater than one mrem per year,(
One sample from a control  location.                          Composite for gross beta and gamma isMpic analysis monthly.
Composite for tritium analysis quarterly.
: d. Sediment        One sample from downstream area          Semiannually.        Gamma ist.pic from              with existing or potential                                    analysisd) shoreline          recreational value.                                            semiannually.
(continued)
(a)      Specific  parameters of distance and direction sector relative to the reactor are provided for each  sample location in Table 5.1. Refer to Radiological Assessment Branch Technical Position,  Revision 1, November 1919. Deviations are permitted from the required sampling schedule  if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal  unavailability, and malfunction of automatic sampling equipment.
(d)      Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(f)      A composite sample is one in which the quantity (aliqiiot) of liquid 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 (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.
(h)        the dose shall  be calculated for the maximum organ and age group,      using the methodology and parameters    in the ODCM.
Columbia Generating Station                        6.3. 1-9                    Amendment 35,      MAR 2002
 
Radiological      Environmental      Monitoring Program 6.3.1 FABLE 6.3.1-1 (page 5 of 6)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SAMPLES          SAMPLING AND            TYPE AND AND/OR SAMPLE                  AND SAMPLE LOCATIONS(T)                COLLECTION          FREQUENCY OF FREQUENCY            ANALYSIS
: 4. INGESTION
: a. Milk      Samples from milking animals in          Semimonthly when      Gamma three locations within five km            animals are on        isotopic(d)(k) distance having the highest dose          pasture, monthly      and 1-131 potential. If there are none, then    at other times,        analysis semi-one sample from milking animals in                                monthly when each of three areas between 5-16 km                              animals are on distant where doses are calculated                                pasture; to be(reater than I mrem per                                      monthly at year.                                                            other times.
One sample from milking animals at a control location, 30-50 km distant and in the least prevalent wind direction.
: b. Fish and      One sample of each of three              Sample annually,      Gamma isotpic Invertebrates    recreationally important species          unless an impact      analysis    " on (one anadromous and two resident)        is indicated, then    edible in vicinity of plant discharge            semiannually.          portions.
area.
One sample of same species in areas not influenced by plant discharge.
(continued)
(a        Specific  parameters of distance and direction sector relative to the reactor are provided for each  sample location in Table 5.1.      Refer to 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, and malfunction of automatic sampling equipment.
(d        Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility, (h)        The dose shall be calculated for the maximum organ and age group,          using the methodology and parameters in the ODCM.
(i        If any of the analytical results for Columbia River fish samples are significantly higher than the results of the Snake River samples or the results of previous fish samples, sampling will be conducted semiannually.
(k)        If Cs-134 or Cs-137 is measured in an individual      milk sample in excess of 30 pCi j,    then Sr-90 analysis shall be performed.
Columbia Generating Station                        6.3.1-10                    Amendment 35,      MAR 2002
 
Radiological      Enviro'nmental    Monitoring Program 6.3.1 TABLE 6.3.1-I (page 6 of 6)
RADIOLOGICAL ENVIRONMENFAL MONITORING PROGRAM EXPOSURE PATHWAY          NUMBER OF REPRESENTATIVE SAMPLES          SAMPLING AND          TYPE AND AND/OR SAMPLE                AND SAMPLE LOCATIONS(a)                COLLECTION        FREQUENCY OF FREQUENCY            ANALYSIS
: 4. INGESTION (continued)
: c. Food      1 sample of each principal class of    At time of            Gamma isotpic Products        food products from any area that is      harvest. J            analyses    on irrigated by water in which liquid                            edible plant wastes have been discharged.                            portion.
Samples of three different kinds of    Monthly during        Gamma broad leaf vegetation grown nearest      growing season.        isotopic(d) each of two different offsite                                  and 1-131 locations of highest predicted                                analysis.
annual average ground-level D/Q if milk sampling is not performed, One sample of each of the similar        Monthly during        Gamma broad leaf vegetation grown 30-50        growing season.      isotopic(d) km distant in the least prevalent                              and 1-131 wind direction if milk sampling is                            analysis.
not performed.
(a)        Specific parameters of distance and direction sector relative to the reactor are provided for each sample location in Table 5.1.        Refer to 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, seasonalunavailability, and malfunction of automatic sampling equipment.
(d)        Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(j)        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 i ncl uding samples of tuberous and root food products.
Columbia Generating Station                          6.3.1-11                    Amendment 35,      MAR 2002
 
Radiological      Environmental    Monitoring Program 6.3.1 TABLE 6.3.1-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES(2)
AIRBORNE                                          FOOD PARTICULATE                                        PRODUCTS WATER          OR GASES              FISH          MILK      (pCi/kg, ANALYSIS          (pCi/L)          (pCi/m 3 )      (pCi/kg, wet)      (pCi/L)        wet)
H-3(1)          2 x 10' Mn-54            1 x 10:3                            3 x  104 Fe-59          4 x 102                                1 x 10i Co-58          I x 10i                              3 x 104 Co-60          3 x 102                              1 x 10' Zn-65          3 x 102                              2 x 104 Zr-Nb-95          4 x 102 1-131              2              0.9                                  3        1 x 10O Cs-134              30                10              1 x 10:'          60        1 x 10' Cs-137              50                20              2 x 10'          70  /    2 x 10' Ba-La-140        2 x 102                                              3 x 102 (1)      For drinking water samples.          The value given is the 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/L may be used.
(2)    The limits are for samples that have only one radionuclide detected.
When a sample contains more than one radionuclide, the total level of radioactivity limit is concentration (1)      concentration (2)        < 1.0 reporting level (1)    reporting level (2)
Columbia Generating        Station            6.3. 1-12                  Amendment 35,    MAR 2002
 
Radiological Environmental      Monitoring Program 6.3.1 TABLE 6'.3.1-3 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS(a)
LOWER LIMIT OF DETECTION (LLD)(b)(c)
AIRBORNE                                      FOOD PARTICULATE            FISH                  PRODUCTS    SEDIMENT WATER        OR GASES            (pCi/kg,      MILK      (pCi/kg,    (pCi/kg, ANALYSIS        (pCi/L)        (pCi/mr)            wet)      (pCi/L)      wet)          dry)
Gross beta            4          1 x 10.1 H-3            2000*
Mn-54            15                                130 Fe-59            30                                260 Co-58,60            15                                130 Zn-65            30                                260 Zr-Nb-95            15                                                                              I 1-131                          7 x 10.'                            1          60 Cs-134            15          5 x 10.2              130          15          60          150 Cs-137            18          6 x  10-2            150          18          80          180 Ba-La-140            15                                              15                              I
(*)      If no drinking water pathway exists,    a value of 3,000 pCi/L may be used.
(a)      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.
(b)      Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13, except for specification regarding energy dependence.      Correction factors shall be provided for energy ranges not meeting the energy dependence specification.
(c)      The LLD is defined in the ODCM definitions section. 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.
Columbia Generating Station                      6.3.1-13                  Amendment 46, OCT 2006
 
Land Use Census 6.3.2 6.3 RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.2  Land Use Census RFO 6.3.2          A Land Use Census shall be conducted  and:
: a. Shall identify the location of the nearest milk animal, residence, and garden of greater than 500 ft 2 producing broad leaf vegetation in each of the 16 meteorological sectors within a distance of 5 miles;
: b. The calculated dose and dose commitment at each newly identified census location shall be less than the most recent values calculated by SR 6.2.2.3.1 at the previous census location.
APPLICABILITY:    At all times.
COMPENSATORY  MEASURES CONDITION            REQUIRED COMPENSATORY MEASURE    COMPLETION TIME A. Land Use Census not      A.1  Prepare and submit the      Upon submittal conducted.                      reasons for not            of current conducting the Land Use    calendar year OR                                  Census and the              Radiological corrective actions to      Environmental All required locations          prevent recurrence.        Operating Report not identified.
B. One or more newly        B.1  Identify new location(s)    Upon submittal identified census            in the Radioactive            of the current locations with the            Effluent Release Report.      calendar year calculated dose or                                          Radioactive dose commitment                                            Effluent Release greater than the                                            Report values calculated by SR 6.2.2.3.1 at the current census locations.
(continued)
Columbia Generating Station            6.3.2-1              Amendment 35, MAR 2002
 
Land Use Census 6.3.2 COMPENSATORY MEASURES CONDITION              REQUIRED COMPENSATORY MEASURE      COMPLETION TIME C. One or more newly          C.1    Add the new location to    30 days identified census                the Radiological locations with the                Environmental Monitoring calculated dose or                Program.
dose commitment, via the same exposure          AND pathway, > 120% of the calculated dose or        C.2    Delete the sampling        After October 31 dose commitment from              location having the        of the year in the current sample                lowest calculated dose      which Land Use location identified in            or dose commitment, via    Census was Table 6.3.1-1.                    the same exposure          conducted pathway, from the Radiological Monitoring Program.
AND C.3    Identify the new          Upon submittal location(s) in the          of the current Radioactive Effluent        calendar year Release Report, and the    Radioactive revised figures and        Effluent Release tables for the ODCM        Report reflecting the new location.
SURVEILLANCE  REQUIREMENTS SURVEILLANCE                                FREQUENCY SR  6.3.2.1        Conduct a Land Use Census during the            Each calendar growing season by a door-to-door survey,      year during the aerial survey, or by consulting local          growing season agriculture authorities and report the results in the Annual Radiological Environmental Operating Report.
Columbia Generating Station            6.3.2-2              Amendment 35, MAR 2002
 
Interlaboratory Comparison Program 6.3.3 6.3 RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.3    Interlaboratory    Comparison Program RFO 6.3.3            Analyses shall be performed on all radioactive materials supplied as part of an Interlaboratory Comparison Program approved by the NRC.
APPLICABILITY:      At all times.
COMPENSATORY    MEASURES CONDITION                REQUIRED COMPENSATORY MEASURE      COMPLETION TIME A. Requirements of RFO          A.1  Prepare and submit, in        Upon submittal 6.3.3 not met.                    the Annual Radiological      of current Environmental Operating      calendar year Report, corrective          Annual actions to prevent            Radiological recurrence.                  Environmental Operating Report SURVEILLANCE    REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR  6.3.3.1          Perform analysis on all radioactive              Annually material supplied as part of the Interlaboratory Comparison Program as described in the OCDM and submit the results in the next annual Radiological Environmental Operating Report.
Columbia Generating Station              6.3.3-1                Amendment 35, MAR 2002
 
Annual Radiological  Environmental  Operating Report 6.4.1 6.4  REPORTING REQUIREMENTS 6.4.1 Annual  Radiological  Environmental  Operating Report Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 15 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 Requirement for OPERABILITY 6.3.2.
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the 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* covering all sampling locations keyed to a table giving distances and directions from the centerline of the reactor; the results of license participation in the Interlaboratory Comparison Program, required by Requirement for OPERABILITY 6.3.3; discussion of all deviations from the sampling schedule of Table 6.3.1-1; and discussion of all analyses in which the LLD required by Table 6.3.1-3 was not achievable.
* One map shall cover stations near the SITE BOUNDARY; a second shall      include the more distant stations.
Columbia Generating Station            6.4.1-1                Amendment 35, MAR 2002
 
Radioactive Effluent Release Report 6.4.2 6.4  REPORTING REQUIREMENTS 6.4.2  Radioactive Effluent Release Report The routine Radioactive Effluent Release Report covering the operation of the unit shall be submitted within 60 days after January I of each year in accordance with 10 CFR 50.36a(a)(2).
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 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
The 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 on 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. This same 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. This same 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 (ODCM Figure 3-1) 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 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 inaccordance with the methodology and parameters in the ODCM.
(continued)
Columbia Generating Station            6.4.2-1              Amendment 35,  MAR 2002
 
Radioactive Effluent Release Report 6.4.2 6.4.2  Radioactive Effluent Release Report    (continued)
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 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 Standards for Nuclear Power Operation. 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 Report shall include a report of solid waste transported from the site during the year as specified by Regulatory Guide 1.2.1, Revision 1, June 1974, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants."
The Radioactive Effluent Release Reports shall include a list  and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to RFO 6.3.2.
Columbia Generating Station          6.4.2-2              Amendment 35, MAR 2002
 
Major Changes To Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems 6.4.3 6.4  REPORTING REQUIREMENTS 6.4.3  Major Changes To Radioactive Liquid,    Gaseous,  and Solid Waste Treatment Systems*
Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):
: a. Shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the POC.
The discussion of each change shall contain:
: 1. A summary of the evaluation that led to the determination that the change could be made without prior NRC approval;
: 2. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
: 3. A detailed description of the equipment, components, and processes involved and the interface with other plant systems;
: 4. 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;
: 5. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; (continued)
Columbia Generating Station            6.4.3-1              Amendment 35, MAR 2002
 
Major Changes To Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems 6.4.3 6.4.3  Major Changes To Radioactive Liquid,    Gaseous,  And Solid Waste Treatment Systems*  (continued)
: 6. 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;
: 7. An estimate of the exposure to plant operating personnel as a result of the change; and
: 8. Documentation of the fact that the change was reviewed and found acceptable by the POC.
: b. Shall become effective upon review and acceptance by the POC.
* Licensees may choose to submit the information called for in this specification as part of the periodic FSAR update.
Columbia Generating Station            6.4.3-2              Amendment 35, MAR 2002
 
ODCM APPENDIX COLUMBIA GENERATING STATION Bases For Radiological  Effluent And Radiological  Environmental  Monitoring Controls
 
Radioactive Liquid Effluent Monitoring    Instrumentation B 6.1.1 B 6.1  INSTRUMENTATION B 6.1.1  Radioactive    Liquid Effluent Monitoring Instrumentation BASES BACKGROUND          The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, releases of radioactive material in liquid effluents during normal plant operation. They are also provided to monitor potential releases from anticipated operational transients.
These instruments are required by General Design Criteria 60, 63, and 64 of Ref. 1. Setpoints for these instruments are established by the methodology of ODCM Section 2.0 to ensure compliance with the limits of Ref. 2.      Specific requirements for this instrumentation are provided in the Radiological Effluent Technical Specifications for BWRs (Ref. 3). Additional guidance is provided in Refs. 4 and 5.
APPLICABLE          The' radioactive liquid effluent monitoring instrumentation SAFETY ANALYSES      provides assurance that radioactive releases resulting from normal operation and any potential radioactive releases caused by Anticipated Operational Transients are monitored.
Chapter 15, Accident Analysis, of the FSAR (Ref. 6) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased radioactive liquid effluent releases. Additional information on the Radioactive Liquid Process and Effluent Radiation Monitoring System can be found in Ref. 7.        The liquid process and effluent radiation monitors are listed in Ref. 8.
REQUIREMENTS        1. Liquid Radwaste Effluent Line Gross Radioactivity FOR OPERABILITY    Monitor This monitor measures the radioactivity in the liquid effluent prior to its entering the cooling tower blowdown line. All radioactive liquid effluent passes through this four inch line. The detector has seven decades of range and has a sensitivity of 1OE-6 yCi/cc for Cs-137.      This monitor will automatically close a valve stopping the effluent release if its setpoint is exceeded.      Ref. 9 provides additional information on this monitor.
(continued)
Columbia Generating Station            B 6. 1. 1-1          Amendment  35,  MAR 2002
 
Radioactive Liquid Effluent Monitoring    Instrumentation B 6.1.1 BASES REQUIREMENTS        2. Deleted FOR OPERABILITY (continued)      3. Turbine Service Water (TSW)    System Gross Radioactivity Monitor The TSW system provides for the removal of heat from balance of plant auxiliary equipment (Ref. 11).      It is normally a non-radioactive system but it has the potential to become contaminated because it provides cooling to some reactor auxiliary systems. This radioactivity monitor is provided to detect that off-normal condition by monitoring for radioactivity in the TSW return header to the Circulating Water System.
The monitor is located on the 441' level of the turbine building. The readout meter and recorder is located in the main control room panel BD-RAD-24.
If the setpoint is exceeded, an alarm will activate in the main control room. The control room operator can then evaluate and take action to terminate the discharge and mitigate any uncontrolled release of radioactive material.
Ref. 12 provides additional information on this subsystem.
: 4. Standby Service Water (SW)    System Gross Radioactivity Monitor The Standby Service Water System provides cooling to the reactor during normal shutdown conditions.      It also provides cooling of the reactor during emergency conditions (Ref.
13). The SW system has the potential to become contaminated because of its interface with the heat exchangers that cool the reactor. This radiation monitor is provided to detect any increase in radioactivity in the system.
The Standby Service Water Monitors are located on the 501 and 522 foot level of the reactor building.      The meter is located in the main control room on panel P-604.
If the setpoint is exceeded, an alarm will activate in the main control room. The control room operator can then evaluate and take action to terminate the discharge and mitigate any uncontrolled release of radioactive material.
Reference 14 provides further information on this subsystem.
(continued)
Columbia Generating Station          B 6.1.1-2            Amendment 36,  June 2002
 
Radioactive  Liquid Effluent Monitoring      Instrumentation B 6.1.1 BASES REQUIREMENTS        5. Liquid Radwaste Effluent Line Flow Rate Monitor FOR OPERABILITY (continued)      This flow rate monitor measures the flow of effluent from the Liquid Waste Management System.        This flow is variable from zero to 190 gpm.      The flow rate recorder is located in the radwaste control room.        Two displays of flow are provided:  the red pen spans a flow rate from zero to 270 gpm and the green pen from zero to 35 gpm.
                  .6. Plant Discharge Blowdown Line Flow Rate Monitor This instrument measures the flow in the circulating water (CW) blowdown line.      Water from the CW system is returned to the Columbia River in order to prevent the buildup of dissolved solids in the cooling water.          This flow is recorded (range 0 - 7500 gpm) in the main control room.          The radioactive liquid effluent from the plant is normally mixed with this blowdown flow to provide dilution before release into the river (Ref. 15).
: 7. Alarm/Trip Setpoints For function 1 the set point is adjusted based on the material to be released.        The monitor is set to ensure the concentration limits of RFO 6.2.1.1 are not exceeded during the release. The setpoint is adjusted to make sure the release into the Columbia River is below the required limits taking into account the amount of radioactivity in.the batch being released and the dilution flow in the blowdown line.
For Functions 2 through 4 the alarm/trip setpoints are based on Ref. 2 limits. To ensure these limits are never exceeded, the alarm setpoint shall be established at 80% or less of the maximum setpoint plus background.          This setpoint corresponds to a Cs-137 concentration of 8E-6 kCi/ml.
Functions 5 and 6 are record only and have no setpoints.            A record of the flow is important to verify dilution parameters.
(continued)
Columbia Generating Station          B 6 . 1.1-3              Amendment 36,  June 2002
 
Radioactive Liquid Effluent Monitoring    Instrumentation B 6.1.1 BASES APPLICABILITY      Gross radioactivity monitors on the Standby Service Water (SSW) System and Turbine Service Water (TSW) System are required to be in service when there is flow in the respective system.      The SSW monitors are not required to be operable when there is no flow in the SSW system.
Similarly, when there is no flow in the TSW system, the TSW radioactivity monitors are not required to be operable.
Without flow, the monitors are not able to detect leakage.
The Liquid Radwaste Effluent Line Gross Radioactivity Monitor and its corresponding Liquid Radwaste Effluent Line Flow Rate Monitor and the Plant Discharge Blowdown Line Flow Rate Monitor are only required when radioactive effluents are being discharged through this pathway.
COMPENSATORY      A.I,  B..1  B.2,  B.3. and E.1 MEASURES These Compensatory Measures give the actions required to be taken if the Liquid Radwaste Effluent Line Gross Radioactivity Monitor (Function 1) is inoperable and batch releases are to continue.      Two independent samples must be taken and analyzed from the batch to be released.          The calculations and valve lineups must be verified by two qualified members of the technical. staff.      These actions are necessary to provide assurance that a controlled release can occur with an inoperable monitor.        If the monitor is not restored to operable status in 30 days this must be reported in the Radioactive Effluent Release Report.        This provides the NRC with information on the status of inoperable equipment.
(continued)
Columbia Generating Station          B 6 .1. 1-4        Amendment 37, January 2003
 
Radioactive Liquid Effluent Monitoring    Instrumentation B 6,1.1 BASES COMPENSATORY        A.I. C.I. C.2. and E.I MEASURES (continued)      These Compensatory Mea'sures are required to be taken if the TSW, or SW Gross Radioactivity Monitors (Functions 3, and 4) are inoperable. Grab samples are required to be taken and I
analyzed every 12 hours. The 12 hour frequency provides an adequate indication of the trend of radioactivity in these locations when the monitor is not OPERABLE.      The channel instrument must also be restored to OPERABLE status within 30 days. The 30 day limit provides an acceptable risk for this loss of monitoring condition.      If the monitor is not restored to OPERABLE status in 30 days this must be reported in the Radioactive Effluent Release Report.
A.I. D.I. D.2. and E.1 These Compensatory Measures are associated with the Liquid Radwaste Effluent Line and Plant Discharge Line Flow Rates.
These flow rates are required to determine the concentration of the effluents released.      If the Liquid Radwaste Effluent Flow Monitor is not OPERABLE and a relea.se is to occur the flow must be estimated.      This is done by observing the level in the tank being pumped initially and every 4 hours.        The combination of tank level and time can be used to estimate the flow. If the Plant Discharge Blowdown Line Flow Rate Monitor is not OPERABLE during a radioactive release this must also be estimated.      Methods of estimating this flow rate using pump differential pressure are given in plant procedures. A release can occur, with no dilution from the blowdown line if the concentration of radioactivity being releases is low enough to meet limits without dijution.
SURVEILLANCE      SR 6.1.1.1 REQUIREMENTS A CHANNEL CHECK is required every 24 hours when a monitor is required to be functional.      The 24 hour time interval provides assurance that the instrument will perform its function if needed.
(continued)
Columbia Generating Station          B 6.1.1-5            Amendment 36, June 2002
 
Radioactive    Liquid Effluent Monitoring Instrumentation B 6 1..1 BASES SURVEILLANCE        SR 6.1.1.1    (continued)
REQUIREMENTS A CHANNEL CHECK for-the flow instruments associated with a batch release is required before a radioactive release and every 24 hours thereafter until the release is complete.
This ensures the flow instruments are operational prior to and during a batch release.
SR 6.1.1.2 The Liquid Radwaste Effluent Line Monitor is used for batch releases and a SOURCE CHECK must be performed prior to each release. This provides increased confidence that the instrument will operate as required to prevent further releases if its trip setpoint is exceeded.
SR 6.1.1.3 A SOURCE CHECK is required monthly to ensure all gross radioactivity monitors are functional. The monthly interval provides adequate reliability of instrument operation if needed to perform its function.
SR 6.1.1.4 A CHANNEL FUNCTIONAL TEST is required on all Radioactive Liquid Monitoring Instrumentation listed on Table 6.1.1-1 quarterly. This test interval is adequate to ensure the instrument will perform its complete function. The CHANNEL FUNCTIONAL TEST for the Liquid Radwaste Effluent Line must demonstrate automatic isolation of the effluent pathway.
(continued)
Columbia Generating Station          B 6 .1. 1-6          Amendment 36, June 2002
 
Radioactive Liquid Effluent Monitoring    Instrumentation B 6.1.1 BASES SURVEILLANCE        SR 6.1.1.4    (continued)
REQUIREMENTS The CHANNEL FUNCTIONAL TEST for all gross radioactivity monitors must demonstrate that control room annunciation occurs if the:
a)  instrument indicates measured levels above the alarm setpoint, b)    high voltage abnormally low,    and c)    instrument indicates downscale failure.
These tests are required to provide assurance of channel operation. In addition, the Liquid Radwaste Effluent Line and SW Gross Radioactivity Monitors must demonstrate control room alarm annunciation if the instrument controls are not set in the operate mode. This test provides increased assurance that the plant operator are aware of an inoperable instrument.
SR 6.1.1.5 A CHANNEL CALIBRATION shall be performed on all Radioactive Liquid Monitoring Instrumentation listed on Table 6.1.1-1 every 18 months. This time interval provides adequate protection against drift and other changes in the instrument performance. For the gross radioactivity monitors the initial CHANNEL CALIBRATION shall be performed using one or more reference standards certified by the National Institute of Science and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST.      These standards shall permit calibrating the system.      For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. These standards provide assurance of instrument accuracy.
(continued)
Columbia Generating Station          B 6.1.1- 7            Amendment 36,  June 2002
 
Radioactive Liquid Effluent Monitoring Instrumentation B 6.1.1 BASES REFERENCES          1. 10 CFR 50,  Appendix A.
: 2. 10 CFR 20.1001-20.2402.
: 3. NUREG-0473,  Revision 2.
: 4. NUREG-0133.
: 5. NUREG-1302
: 6. FSAR Section 15.0.3.
: 7. FSAR Section 11.5.2.2.2.
: 8. FSAR Table 11.5-2.
: 9. FSAR Section 11.5.2.2.2.3.
: 10. FSAR Section 11.5.2.2.2.5.
: 11. FSAR Section 9.2.1.
: 12. FSAR Section 11.5.2.2.2.4.
: 13. FSAR Section 9.2.7.
: 14. FSAR Section 11.5.2.1.4.
: 15. FSAR Section 11.2.3.1.
Columbia Generating Station          B 6 .1. 1-8        Amendment 36, June 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 B 6.1  INSTRUMENTATION B 6.1.2  Radioactive Gaseous Effluent Monitoring Instrumentation BASES BACKGROUND          The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, releases of radioactive material in gaseous effluents.        These instruments are required by General Design Criteria 60, 63, and 64 of Ref. 1. Setpoints for these instruments are established by the methodology of ODCM Section 3.0 to ensure compliance with the limits of Ref. 2.        Specific requirements for this instrumentation are provided in the Radiological Effluent Techni.cal Specifications for BWRs (Ref. 3).
Additional guidance is provided in Refs. 4 and 5.
APPLICABLE          The radioactive gaseous effluent monitoring instrumentation SAFETY ANALYSES    providesrassurance that radioactive releases resulting from normal operation and any potential radioactive releases caused by Anticipated Operational Transients are monitored.
Chapter 15, Accident Analysis, of the FSAR (Ref. 6) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent releases.      The FSAR (Ref. 7) provides additional information on the Gaseous Process and Effluent Radiation Monitoring System.      Ref. 8 provides additional detail on the Gaseous Process and Effluent Radiation Monitors.
REQUIREMENTS        1. Main Condenser Offgas Post-Treatment Radiation Monitors FOR OPERABILITY Two monitors measure the radioactivity in the offgas piping downstream of the off gas system charcoal vessels and upstream of the offgas system discharge valve.        The purpose of these radiation monitors is to provide information on the amount of beta radiation in the offgas process stream after treatment and to automatically isolate the offgas system or I
to automatically initiate offgas treatment if the system is being operated in the bypass mode.      Radiation is measured by a scintillation detector and countrate meters display the results in the main control room.      An auxiliary trip unit I
uses the signal  from  these two  detectors  to isolate the charcoal system if the logic is satisfied by a combination (continued)
Columbia Generating Station            B 6.1. 2-1              Amendment 44,  MAY 2005
 
Radioactive Gaseous Effluent Monitoring    Instrumentation B 6.1.2 BASES REQUIREMENTS        1. Main Condenser Offqas Post-Treatment Radiation Monitors FOR OPERABILITY      (continued) of high-high-high radiation and downscale trips.        This is the normal Mode:of operation with the charcoal beds in service. Another trip signal from these monitors initiates opening of the charcoal absorber treatment valve and closure of the bypass valve if the system is being run in the bypass mode during plant startup. Alarms are provided in the main control room that show monitor system status.      Further information on the subsystem can be found in Ref. 9.
: 2. Main Condenser Offqas Pre-Treatment Radiation Monitor This monitor is located at the outlet of the water separator downstream of the catalytic recombiners.      The monitor measures the radioactivity due to fission gases and activation products produced in the reactor and transported with the steam through the turbine to~the condenser.        The purpose of this monitor is to provide the main control room with information on the amount of gamma radioactivity in the untreated offgas stream.      The detector is a gamma-sensitive ion chamber. Trip outputs are used for alarm only.      Further information on this subsystem can be found in Ref. 10.
2a. Mechanical  Vacuum Pump Discharge Radiation Monitor This Geiger Mueller type radiation monitor is designed to alarm, stop, and isolate the mechanical vacuum pumps in the event of high levels of radioactive gases in the air being exhausted from the condenser.      The mechanical vacuum pumps are operated during plant start-ups to remove bulk air from the condenser.      They are stopped when the steam jet air ejectors are put into service.      Further information on this radiation monitoring system can be found in Ref. 11.
: 3. Reactor Building Elevated Discharge Radiation Monitor This monitoring system measures radioactivity in the reactor building elevated release duct from the a)    Offgas effluent, b)    the gland seal system, (continued)
Columbia Generating Station            B 6.1.2-2            Amendment  44,  MAY 2005
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES REQUIREMENTS      3. Reactor Building Elevated Discharge Radioactivity-Monitor FOR OPERABILITY    (continued) c)    Mechanical vacuum pump, d)    the Standby Gas Treatment System, and e)    the exhaust air from the Reactor Building Ventilation System.    -
The purpose of this monitoring subsystem is to characterize the radioactivity being released through the reactor building elevated discharge vent and to provide alarms in the event of abnormal operation. The monitor has no control function but an alarm is provided in the control room if the setpoint is exceeded.
A continuous representative sample is extracted from the elevated release duct through an isokinetic probe. The sample is filtered to collect particulates and passes through an impregnated charcoal cartridge to collect iodine.
The sample travels through a flow indicator that provides flow rate and totalized flow. Both the sample flow rate and the exhaust duct flow rate and their totalized flow are recorded and alarmed in the control room. These samples are collected and analyzed per SR 6.2.2.1.2 and Table 6.2.2.1-1.
Further information on this subsystem can be found in Ref. 12.
Footnote (b) is intended to address the condition when neither the Reactor Building Ventilation System nor the Standby Gas Treatment System is operating. If there is no flow through the elevated discharge point this monitor is not OPERABLE and the Compensatory Measures are also not effective. The reactor building is a confinement structure which acts as a secondary containment-so leakage of radioactive material should be limited. Under these no effluent flow conditions building ambient air samples, building equipment status, including the identification of possible sources of radioactivity, and meteorological data (continued)
Columbia Generating Station          B 6 . 1. 2-3 B6123Aedet3,MR20    Amendment 35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES REQUIREMENTS        3. Reactor Building Elevated Discharge Radioactivity Monitor FOR OPERABILITY      (continued) can be used to estimate the out-leakage from the building.          _
The Technical Specifications place severe limits on the amount of time that can be spent without secondary containment so this condition is not likely to persist for any length of time.
: 4. Turbine Building Exhaust Radiation Monitor This monitoring system measures radioactivity in the turbine building exhaust. The monitor detects fission and activation products from the steam which may leak from the turbine or other equipment in the building.      The purpose of this monitoring subsystem is to characterize the radioactivity being released through the turbine building discharge vent and to provide alarms in the event of abnormal operation.      The monitor is read and recorded in the main control room.      An alarm is provided in the control room if the setpoint is exceeded.
A continuous representative sample is extracted from the exhaust duct through an multi-ported isokinetic probe.            The sample is filtered to collect particulates and passes through an impregnated charcoal cartridge to collect iodine.
These samples are collected and analyzed per SR 6.2.2.1.2 and Table 6.2.2.1-1.
Further information on this subsystem can be found in Ref.
13.
Footnote (b) is intended to address the condition when the Turbine Building Ventilation System is not operating.          If there is no flow through the turbine building exhaust this monitor is not OPERABLE and the Compensatory Measures are also not effective.      The turbine building is a conventional steel structure so leakage of radioactive material could be significant. Under these conditions building ambient air samples, building equipment status including potential radioactivity sources, and meteorological data can be used to estimate the out-leakage from the building.
(continued)
Columbia Generating Station          B 6. 1. 2-4            Amendment  42,  AUG 2004
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES REQUIREMENTS        5. Radwaste Buildinq Ventilation Exhaust Radiation Monitor FOR OPERABILITY (continued)      This monitoring system measures radioactivity in the radwaste building ventilation air exhaust.      Radioactivity originates from radwaste tank vents, processing equipment, sampling hoods, as well as from process treatment systems within the building.      The purpose of this monitoring subsystem is to characterize the radioactivity being released through radwaste building discharge vent and to provide alarms in the event of abnormal operation.        The monitor is read and recorded in the main control room.          An alarm is provided in the control room if the setpoint is exceeded.
A continuous sample is extracted from the exhaust duct, through a multi-ported isokinetic, probe.      The sample is filtered to collect particulates and passes through an impregnated charcoal cartridge to collect iodine.        These samples are collected and analyzed per SR 6.2.2.1.2 and Table 6.2.2.1-1.
Further information on this subsystem can be found in Ref.
14.
Footnote (b) is intended to address the condition when the Radwaste Building Ventilation System is not operating.          If there is no flow through the radwaste building exhaust this monitor is not OPERABLE and the Compensatory Measures are also not effective.      Under these conditions building ambient air samples, building equipment status including potential radioactive sources, and meteorological data can be used to estimated the out-leakage from the building.
: 6. Alarm/TriD Setnoints The Function 1 monitors are normally operated in the mode where they, are monitoring the output of the gaseous effluent stream from the charcoal beds.      Under these conditions a High-High-High Radiation signal from both monitors initiates closure of OG-V-60 terminating the effluent release.        The setpoints are established to assure the valve will be closed before the site boundary dose rate criteria is exceeded.            A downscale trip of both monitors or a combination of downscale trip and High-High-High radiation will also cause (continued)
Columbia Generating Station          B 6 .1. 2-5            Amendment.35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring      Instrumentation B 6.1.2 BASES REQUIREMENTS        6. Alarm/Trip Setpoints      (continued)
FOR OPERABILITY isolation of the effluent release.      The Function I monitors can also be operated in the mode where the charcoal beds are being bypassed. In this case a High radiation trip from either monitor will cause closure of the bypass valve, OG-V-
: 45. Prior to placing the gaseous radwaste treatment system in the charcoal bypass mode, the alarm setpoints on the main plant vent release monitor shall be set to account for the increased percentages of short-lived noble gases.        Noble gas percentages shall be based either on actual measured values or on primary coolant design base noble gas concentration percentages adjusted for 30-minute decay.      Table 3-15 lists the percentage values for 30-minute decay.
The setp~oint for this action is well below the site boundary dose rate criteria. Control room alarms are provided for a downscale condition, High Radiation, and High-High-High Radiation.
The Function 2, High Radiation alarm setpoint-is required to be set at 1.5 times the steady state full power nominal radiation level per Ref. 15.      This provides assurance that the control room will be aware of increasing radiation levels in the untreated gaseous effluent stream from the main condenser. A second High-High Radiation alarm is provided giving the control room indication if the level continues to rise.
The Function 2a, 3a, 4a, and 5a monitors have their setpoints established in a manner that assures the instantaneous dose rate limits to the unrestricted area are less than the limits given in RFO 6.2.2.1.        ODCM Section 3.4 provides the methodology for establishing these setpoints.
The calculations take into account the partition of the limiting dose rate between the three release points.          The maximum normalized diffusion coefficient (X/Q) at and beyond the site boundary is used to model the decrease in radionuclide concentration between the point of release and the receptor location.        Both whole body and skin dose setpoints are calculated and the lower setpoint is used for the instrument. The dose rate limits are provided in Ref. 2.
(continued)
Columbia Generating Station          B 6. 1. 2-6              Amendment 35,  MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES REQUIREMENTS        6. Alarm/Trip Setpoints      (continued)
FOR OPERABILITY Functions 3b, 3c, 4b, 4c, 5b, and 5c do not have established setpoints. These functions collect iodine and particulate samples for further analysis and no setpoints are associated
                  -with these activities.      It is not considered to be practicable to apply instantaneous alarm/trip setpoints to integrating radiation monitors sensitive to radioiodines or radioactive materials in particulate form.
Functions 3d, 4d, and 5d are associated with the flow rate of effluent from the three buildings.        Design ventilation rates are given in Ref. 16.      These setpoints are established to provide the plant operators with an alarm indicating abnormal flow conditions.
Functions 3e, 4e, and 5e are associated with flow rate in the iodine and particulate sampling devices.        The flow through these filters must be maintained to obtain a representative sample.      Setpoints are established to provide plant operators with an alarm if the sample flow fails to register the required value.
Further guidance on setpoint determination can be found in Ref. 4.
APPLICABILITY      The Main Condenser Offgas Post-Treatment and Pre-Treatment Monitors are required to be OPERABLE when the steam jet air ejectors are put into operation.        Prior to this time condenser offgas is exhausted by the mechanical vacuum pump.
The Mechanical Vacuum Pump Radiation Monitor is required to be OPERABLE when the Mechanical Vacuum Pumps are being used to create and maintain a condenser vacuum and there is a potential for radioactive release from the condenser area.
During startup and shutdown when vacuum is being maintained with a Mechanical Vacuum Pump and the Steam Seal System is operational the Mechanical Vacuum Pump Radiation Monitor must be OPERABLE.
All other instruments are required to be OPERABLE when there is exhaust flow through the building exhaust vent.          If there is no flow through the exhaust, the monitors do not provide any useful information.      In this case footnote (b) of Table 6.1.2-1 shall be followed to prevent an unmonitored release condition.
(continued)
Columbia Generating Station          B 6 .1.2- 7              Amendment 35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES COMPENSATORY        Separate entry is allowed for each channel. The note allows MEASURES            a Condition to be entered separately for each inoperable instrument and completion times tracked on a per instrument basis.
A.1 and B.1 This Compensatory Measure gives the action required if the Mechanical Vacuum Pump Radiation Monitor is not OPERABLE when it is required. The Reactor Building Elevated Discharge Radiation Monitor monitors the effluent pumped from the condenser by the Mechanical Vacuum Pump. If the Reactor Building Elevated Discharge Radiation Monitor is OPERABLE this provides adequate assurance that potential releases will be monitored. If this monitor is also INOPERABLE then sampling is required.
A.I. C.I. C.2. C.3. and F.1 The Compensatory Measures give the actions required to be taken if the Main Condenser Offgas Post-Treatment Radiation Monitor, the Reactor Building Elevated Discharge Low Range Activity Monitor, the Turbine Building Ventilation Exhaust Noble Gas Activity Monitors, or the Radwaste Building Ventilation Exhaust Noble Gas Activity Monitors are inoperable. A grab sample must be taken at the associated sample location within B hours and every 8 hours thereafter.
The grab sample must be analyzed for noble gas gamma emmiters within 24 hours after the sample is taken. This sampling and analysis provides adequate assurance that the noble gases being released are characterized. If the channel is not restored to OPERABLE status in 30 days, the event must be reported to the NRC in the calendar year Radioactive Effluent Release Report. This allows the NRC to monitor equipment that has been inoperable for an extended time period.
(continued)
Columbia Generating Station          B6128Aedet3,MR20 B 6.1.2-8            Amendment 35, MAR 2002
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES COMPENSATORY      A.1. D.I. D.2. and F.1 MEASURES (continued). These COMPENSATORY MEASURES are required to be taken if the Effluent Flow Rate Monitors in the Reactor Building Elevated Discharge, the Turbine Building Ventilation Exhaust, or the Radwaste Building Ventilation Exhaust are INOPERABLE. An estimate of the flow rate from the associated building is required every 4 hours. An estimate of the flow c-an be obtained by noting the number of fans in operation.
These same COMPENSATORY MEASURES are also required to be taken if associated building Iodine and Particulate Sampler Flow Rate Monitors are INOPERABLE. Sample line flow rates are needed to determine the concentration (activity in a given volume of air) of radioactive material in gaseous effluents. The flow rate at each release point is necessary to determine the total activity of radioactive material released in a given time period. There is flexibility in the method by which these estimates can be made for the purpose of obtaining best-available estimates. *For example, if a flow rate or alarm signal is not being received at the control room, but the installed, calibrated, mechanical flow rate device is working properly, the best-available estimate would come from the mechanical flow rate device as opposed to using auxiliary sampling equipment. On the other hand, if the local flow rate indication is electronic, as is the case with the reactor building sample flow rate, and out of service, then auxiliary flow rate instruments may be used.
The operation of the sampling equipment and sample flow is recorded every 4 hours.
If any flow related channel above is not restored to OPERABLE status in 30 days the event must be reported to the NRC in the calendar year Radioactive Effluent Release Report. This allows the NRC to monitor equipment that has been inoperable for an extended time period.
A.1. E.1.1. E .1.2. E.2. and F'.1 These Compensatory Measures are required to be taken if the Iodine and Particulate Samplers associated with the Reactor Building Elevated Discharge, the Turbine Building (continued)
Columbia Generating Station          B6129Aedet4,FB20 8 6. 1.2-9        Amendment 45, FEB 2006
 
Radioactive Gaseous Effluent Monitoring        Instrumentation B 6.1.2 BASES COMPENSAFORY        A.I,  E.I.I. E.1.2,  E.2,    and F.1  (continued)
MEASURES Ventilation Exhaust, or the Radwaste Building Ventilation Exhaust are inoperable.        Auxiliary sampling equipment is used to collect samples from the associated effluent release pathway within 4 hours.          This auxiliary sampler is to be used continuously to provide a sample for later analysis.
Compensatory Measure E.I.2 can be used if auxiliary sampling is not available.      Relevant information should be collected to characterize the particulate and iodine effluent releases during the period of time the auxiliary sampling equipment was not available when required by Compensatory Measure E.1.1. If Compensatory Measure E.I.2 is used this must be reported to the NRC upon submittal of the current calendar year Radioactive Effluent Release Report.
A.1,  G.1,  G.2,  G.3 and H.1 If the normal sampling equipment is not restored to OPERABLE status in 30 days the event must be reported to the NRC in the calendar year Radioactive Effluent Release Report.            This allows the NRC to monitor equipment that has been inoperable for an extended time period.
These Compensatory Measures are associated with the Main Condenser Offgas Pre-Treatment Radiation Monitor.            If this monitor is inoperable immediate action must be taken to verify the Offgas Treatment System (RECHAR) is in operation.
To maintain continuous monitoring, within 4 hours place a temporary Area Radiation Monitor (ARM) to monitor pre-treatment radiation levels and verify the operation of the ARM every 4 hours,. To verify that the noble gas 30-minute decay gross gamma activity rate is < 332 mCi/second (Reference TS 3.7.5), take and analyze an Offgas grab sample within 8 hours and every 24 hours thereafter.            Efforts must be made to restore the Main Condenser Offgas Pre-Treatment Radiation Monitor to OPERABLE status in 30 days.            If this is not done, the plant must be in Hot Shutdown within the following 12 hours.        Plant shutdown is required because continuous monitoring of the pre-treatment radiation level provides a direct indication of the condition of the reactor fuel. Continued power operation beyond 30 days is not allowed.
(continued)
Columbia Generating Station            B 6. 1 .2-I10              Amendment 45, FEB 2006
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6. 1. 2 BASES  (continued)
SURVEILLANCE        SR 6.1.2.1 REQU IREMENTS A CHANNEL CHECK is required every 24 hours for the radiation monitoring systems and flow rate monitors. The 24 hour time interval provides assurance that the instrument will perform its function if needed. A CHANNEL CHECK for the Main Condenser Offgas Post-Treatment and Pre-Treatment Radiation Monitors is required every 24 hours when the Steam Jet Air Ejectors are in operation. A CHANNEL CHECK of the building noble gas gadmma monitors is required every 24 hours at all times. For the building flow rate instruments, a CHANNEL CHECK is required every 24 hours when the associated building. exhaust system is operational. The CHANNEL CHECK for the Radwaste Building Effluent System Flow Rate Monitor shall be performed by comparing a computer reading or power signal comparing each fan's local amperage reading with pre-established baseline values..
SR 6.1.2.2 A daily SOURCE CHECK is required for the Main Condenser Offgas Post-Treatment Radiation Monitors. This provides increased confidence that the instruments will operate as required to detect excess radioactivity in the treated gas
                  *process stream. This test is required whenever the steam jet air ejectors are in operation.
SR 6.1.2.3 A CHANNEL CHECK for the iodine and particulate samplers is required weekly. The change-out of the filters is considered the equivalent of a CHANNEL CHECK. The weekly check of sampler operation is sufficient to provide assurance of a quality sample for analysis. This CHANNEL CHECK is required at all times.
SR 6.1.2.4
                  -The monthly SOURCE CHECK is required for the Main Condenser Offgas Pre-Treatment Monitor whenever the steam jet air ejectors are in operation. A monthly SOURCE CHECK is also (cont inoued)
B 6 . 1.2- 11 Columbia Generating StationB61.11Aednt                                4,MY20 Amendment 44,  MAY 2005
 
Radioactive Gaseous Effluent Monitoring Instrumentation 8 6.1.2 BASES SURVEILLANCE        SR 6.1.2.4  (continued)
REQUIREMENTS required for the Reactor Building Elevated Discharge Low Range Activity Monitor, the Turbine Building Ventilation Exhaust Noble Gas Activity Monitor, and the Radwaste Building Ventilation Exhaust Noble Gas Activity Monitor.
The building effluent noble gas gamma monitors require a source check at all times. The monthly SOURCE CHECK provides adequate assurance of channel OPERABILITY.
SR 6.1.2.5 A CHANNEL FUNCTIONAL TEST is required for all instruments except the iodine and particulate samplers and the Mechanical Vacuum Pump Discharge Radiation Monitor quarterly or every 92 days. This test frequency provides adequate assurance of the channel function. A CHANNEL FUNCTIONAL TEST is not required for the samplers as the routine analysis of the samples provides assurance of sample function. The CHANNEL FUNCTIONAL TEST for the Main Condenser Offgas Post and Pre-Treatment Monitors, the Reactor Building Elevated Discharge Low Range Noble Gas Monitor require a demonstration that control room alarm annunciation occurs if:
a)    the instrument indicates measured levels above the alarm setpoint,  and b)    circuit failure occurs.
The CHANNEL FUNCTIONAL TEST for the Turbine Building Ventilation Exhaust and Radwaste Building Ventilation Exhaust Noble Gas Monitors must demonstrate that circuit failures or instrument controls when set in the off position produce control room alarm annunciation. The CHANNEL FUNCTION TEST for the Radwaste Building Ventilation Exhaust Effluent System Flow Rate Monitor shall be performed by measurement of the ,phase currents for each fan.
(continued)
Columbia Generating Station          B 6.1.2-12            Amendment 42, AUG 2004
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES SURVEILLANCE        SR 6.1.2.5a REQUIREMENTS (continued)    A CHANNEL CHECK of the Mechanical Vacuum Pump Radiation Monitor is required prior to its use.      This check is required only if it has not been performed during the past 30 days. During shutdowns this check can be performed by verifying that the instrument was on-scale prior to shutdown.
SR 6.1.2.6 A CHANNEL CALIBRATION is required for all instruments except the iodine and particulate samplers, the Offgas Post Treatment Monitor, and the Mechanical Vacuum Pump-Discharge Radiation Monitor every 18 months. This test frequency provides adequate protection against drift or other instrument inaccuracies. The CHANNEL CALIBRATION is not required for the samplers as the routine analysis of the samples provides assurance of sample integrity.      The CHANNEL CALIBRATION of the Main Condenser Offgas Pre-Treatment Monitors, the Reactor Building Elevated Discharge Low Range Activity Monitor, and the Turbine and Radwaste Building Low Range Noble Gas Activity Monitors have special requirements.
For these gross radioactivity monitors the initial CHANNEL CALIBRATION shall be performed using one or more reference standards traceable to the National Institute of Science and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and monitor response.      Subsequent CHANNEL CALIBRATION shall be performed using the equivalent quality and radioactive sources that have been related to the initial calibration.
The CHANNEL CALIBRATION for the Radwaste Building Ventilation Effluent System Flow Rate Monitor shall be performed by using a flow measurement device to determine the fan current to flow relationship.
(continued)
Columbia Generating Station          B 6.1.2-13            Amendment 46, OCT 2006
 
Radioactive Gaseous Effluent Monitoring Instrumentation B 6.1.2 BASES SURVEILLANCE        SR 6.1.2.7 REQUIREMENTS (continued)      The CHANNEL CALIBRATION for the Offgas Post Treatment Monitor and the Mechanical Vacuum Pump Discharge Radiation        I Monitor is required during refueling outages.      This longer frequency is acceptable since this monitor is backed up by the Reactor Building Elevated Discharge Radiation Monitor.
The initial calibration of the Offgas Post Treatment Monitor shall be performed using sufficient NIST traceable sources to permit calibrating the system over its intended range of energy and monitor response.
Subsequent channel calibration shall be performed using equivalent quality and radioactive sources that have been related to the primary calibration.
REFERENCES        1. 10 CFR 50,  Appendix A.
: 2. 10 CFR 20.1-20.602.
: 3. NUREG-0473,  Revision 2.
: 4. NUREG-0133.
: 5. NUREG-1302.
: 6.      FSAR Section 15.0.3.
: 7.      FSAR Section 11.5.2.2.1.
: 8.      FSAR Table 11.5-1.
: 9.      FSAR Section 11.5.2.2.1.2.
: 10. FSAR Section 11.5.2.2.1.1.
: 11. FSAR Section 11.5.2.2.1.4.
: 12. FSAR Section 11.5.2.2.1.5.
: 13. FSAR Section 11.5.2.2.1.6.
: 14. FSAR Section 11.5.2.2.1.7.
: 15. Technical  Specification Amendment 112 (G12-93-028).
: 16. FSAR Table 11.3-8.
Columbia Generating Station            B 6. 1.2-14          Amendment 46, OCT 2006
 
Liquid Concentration B 6.2.1.1 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.1  Liquid Effluents B 6.2.1.1    Liquid Concentration BASES BACKGROUND          This RFO is focused on the concentration of radionuclides released to the Columbia River. The radionuclide concentrations shall not exceed 10 times those given in Appendix B, Table 2, Column 2 of Ref. 1. The RFO also implements the requirements of Technical Specification 5.5.4.b.
APPLICABLE          The radionuclide concentration limits provide assurance that SAFETY ANALYSES    any liquid radioactive releases caused by normal operation and Anticipated Operational Transients result in very low exposures to the general public.      A hypothetical person using water at the outlet of the blowdown line at the RFO limits would receive a whole body dose of 500 mrem/year.
Chapter 15, Accident Analysis, of the FSAR (Ref. 2) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased liquid effluent release concentrations.      Further information on liquid radioactive releases can be found in Chapter 11 of the FSAR (Ref. 3). Ref. 4 provides initial estimated isotopic annual releases of liquid effluent.
REQUIREMENTS        This RFO provides assurance that the concentration of FOR OPERABILITY      radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within the limits of Part 20.1302(b)(2)(i) of Ref. 1. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in Ref. 6.
APPLICABILITY      This RFO is applicable at all times since liquid effluents can be released at any time.
(continued)
Columbia Generating Station          B 6 .2. 1.1-1            Amendment 35, MAR 2002
 
Liquid Concentration B 6.2.1.1 BASES COMPENSATORY        The concentration of radionuclides being released can be MEASURES            decreased by decreasing the amount of activity being released or by increasing the dilution.      The amount of radioactivity will decrease with time due to decay.        The recycling of water within the plant prior to release will, therefore, decrease the amount of activity available.        The dilution of the waste can occur within the plant prior to release or during blowdown to the river.
SURVEILLANCE      This SR defines the radioactive liquid waste sampling and REQUIREMENTS      analysis needed to determine the radionuclide concentrations in the effluent. This program meets the requirements of General Design Criteria 64 in Appendix A of Ref. 5.        It also uses the guidance provided in References 7 and 8.
Liquid Release Type At Columbia Generating Station batch releases are the only release type. 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 by a method described in plant procedures to assure representative sampling. SR 6.2.1.1.1 requires a pre-release analyses of each batch release.
Sample and Analysis Frequency A sample is required to be taken prior to the release of each batch. This provides assurance of a representative sample of the radioactivity to be released.      The samples must be analyzed prior to the release of each batch.        This allows for a review of the results of the analysis prior to release. SR 6.2.1.1.2 requires a post-release analysis of samples composited from the batch releases.      For tritium and gross alpha a composite sample is collected prior to the release of each batch. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.      This may be (continued)
Columbia Generating Station        B 6.2.1.1-2.            Amendment 35,  MAR 2002
 
Liquid Concentration B 6.2.1.1 BASES SURVEILLANCE        Samole and Analysis Freouency        (continued)
REQUIREMENTS accomplished through composites of grab samples obtained prior to discharge after the tanks have been recirculated.
Every 31 days the composite sample made up of all the releases during the prior 31 days must be analyzed.            If no releases are made during the previous 31 day period no analysis is required.        This 31 day composite analysis for tritium and gross alpha is considered adequate to characterize the liquid release during this time period.
For Fe-55, Sr-89, and Sr-90 a composite sample is also required for each batch released.          In this case the composite samples may be collected for 92 days prior to analysis. If no releases occur during the previous 92 day period no analysis is required.
Sample Lower Limit of Detection The LLD is defined in the definitions section of the ODCM Appendix. The principle gamma emitters for which the LLD definition applies in this RFO includes the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141.          Ce-144 shall also be measured, but with an LLD of 5 x 10-.          This list does not mean that only these nuclides are to be considered.          Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.
Detailed discussions of the LLD and other detections limits can be found in References 9, 10, and 11.
REFERENCES          1. 10 CFR 20.1001 - 20.2402.
: 2.      FSAR Section 15.0.3.
: 3.      FSAR 11.2.3.
: 4.      FSAR Table 11.2-14.
: 5.      10 CFR 50.
(continued)
Columbia Generating Station          B 6 . 2.1 .1-3              Amendment  35,  MAR 2002
 
Liquid Concentration B 6.2.1.1 BASES REFERENCES        6. ICRP Publication 2.
(continued)
: 7. Reg. Guide 1.21.
: 8. Reg. Guide 4.15.
: 9. HASL Procedures Manuals,  HASL-300 (revised annually).
: 10. Currie, L.A. "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).
: 11. Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
Columbia Generating Station        B 6.2.1.1-4          Amendment 35, MAR 2002
 
Liquid Dose B 6.2.1.2 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.1  Liquid Effluents B 6.2.1.2  Liquid Dose BASES BACKGROUND          This RFO requires a ca.lculation of the dose resulting from the routine release of liquid radioactive effluents.      It implements the requirements of Sections II.A, III.A, and IV.A of Appendix I of Ref. 1. It should be noted that these RFO limits apply to a MEMBER OF THE PUBLIC. The dose calculation models are, therefore, allowed to take into account dilution and other factors which decrease the dose between the point of release and the person's location.
APPLICABLE          This dose calculation provides assurance that any liquid SAFETY ANALYSES    radioactive releases caused by normal operation and Anticipated Operational Transients are evaluated to determine their impact on a MEMBER OF THE PUBLIC.      Chapter 15, Accident Analysis, of the FSAR (Ref. 2) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased liquid effluent releases. The FSAR (Ref. 3) and the Columbia Generating Station Environmental Report (Ref. 4) calculate estimated annual exposure due to liquid effluent release.
The estimated whole body dose of 2.3 mrem per year and the largest calculated single organ dose of 1.6 mrem per year to I
the bone are well below the guidelines of Appendix I of Ref. 1.
REQUIREMENTS        The dose calculation methodology and parameters in the ODCM FOR OPERABILITY    implement the requirements of Section III.A of Appendix I, Conformance with the guidance of Appendix I must be shown by calculational procedures based on models and data, such that the actual exposure of the MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the dose due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Ref. 5 and Ref. 6.      For fresh water sites with drinking water supplies that can be (continued)
Columbia Generating Station          B 6.2. 1 .2-1        Amendment 38, APRIL 2003
 
Liquid Dose B 6.2.1.2 BASES REQUIREMENTS      potentially affected by plant operations, there is FOR OPERABILITY    reasonable assurance that the operation of the facility (continued)    will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of Ref. 7.
APPLICABILITY      This RFO-is applicable when radioactive liquid effluents are released. During the periods of time when no radioactive liquids are released this RFO is not applicable.
COMPENSATORY      The general note is provided to implement the guidance in MEASURES          Ref. 8. This NUREG states that as long as the calculated dose does not exceed twice the limits of this RFO no extra analysis is required to demonstrate compliance with RFO 6.2.3.1. If the dose levels exceed twice those cited in this RFO additional analysis must be performed to determine if additional limitations on plant operation will be necessary to ensure compliance with Ref. 9.
The COMPENSATORY MEASURES statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I of Ref. 1 to ensure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."
SURVEILLANCE      This SR for this RFO requires a dose calculation every REQUIREMENTS      31 days to provide assurance that the radioactive liquid effluents released do not result in a dose above the limits stated. The dose to the whole body is calculated for the quarter and year prior to the date of the calculation.        The dose to each organ is calculated for the quarter and year prior to the date of the calculation.      All calculated doses must be below the limits given in this RFO or the COMPENSATORY MEASURES must be implemented.      When no liquid radioactive releases occur during the time period no new calculation is required since the RFO is not applicable.
(continued)
Columbia Generating Station          B 6.2.1.2-2          Amendment 38,  APRIL 2003
 
Liquid Dose B 6.2.1.2 BASES REFERENCES          1. 10 CFR 50.
: 2. FSAR Section 15.0.3.
: 3. FSAR 11.2.3.3.
: 4. Columbia Generating Station Environmental Report, Operating License Stage, Section 5.2.
: 5. Regulatory Guide 1.109.
: 6. Regulatory Guide 1.13.
: 7. 40 CFR 141.
: 8. NUREG-0543.
: 9. 40 CFR 190.
Columbia Generating Station        B 6.2.1.2-3        Amendment 35, MAR 2002
 
Liquid Waste Management System B 6.2.1.3 B 6.2 RADIOACTIVE EFFLUENTS B 6.2.1  Liquid Effluents B 6.2.1.3  Liquid Waste Management System BASES BACKGROUND          This RFO requires the Liquid Waste Management System to be OPERABLE at all times.      In addition it requires the Liquid Waste Management System to be in service whenever the projected dose calculated prior to release indicates a value above the limit given in the RFO.        The requirement that the appropriate portions of this system be used, when specified, provides assurance that the release of radioactive materials in liquid effluent wil.l be kept "as low as is reasonably achievable."  This RFO implements the requirements of Part 50.36a, General Design Criterion 60 of Appendix A and the design objective given in Section II.D of Appendix I to Ref. 1. It also implements the requirements of Technical Specification Section 5.5.4.f.      Further guidance on the implementation of the RFO is provided in Refs. 2, 3, and 4.
APPLICABLE          This RFO provides assurance that liquid radioactive releases SAFETY ANALYSES    caused by normal operating and Anticipated Operational Transients are treated if they are above the given limits.
Chapter 15, Accident Analysis, of the FSAR (Ref. 5) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased liquid effluent releases.
REQUIREMENTS      The OPERABILITY of the Liquid Radioactive Waste Management FOR OPERABILITY    System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment.      The term "Liquid Waste Management System" involves all of the installed and available Liquid Radioactive Waste Management System equipment, as well as their controls, power, instrumentation, and services that make the system function.      Equipment that is considered standby or redundant is also included, since their function is to assure OPERABILITY.      The RFO also permits alternate treatment paths using alternate subsystems and equipment to (continued)
Columbia Generating Station        B 6 .2.,1.3- 1            Amendment 35, MAR 2002
 
Liquid Waste Management System B 6.2.1.3 BASES REQUIREMENTS        be used in the event that the normal treatment equipment is FOR OPERABILITY    inoperable. Further information on the design bases, system (continued)      description, and process description of the Liquid Waste Management System is given in Ref. 6.
This RFO requires the Liquid Waste Management System to be in use when the projected doses due to the liquid effluent released to the UNRESTRICTED AREA would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.
Thus, the Liquid Waste Management System is required to be I
in use well before the limits of RFO 6.2.1.2 are reached as required by Section II.A of Appendix I to Ref. 1. If the system is in use radioactive liquid waste can be released up to the limits of RFO 6.2.1.2.
APPLICABILITY      This RFO is applicable at all times since the Liquid Waste Management System must be OPERABLE at all times.
COMPENSATORY      A Special Report to the NRC is required if the Liquid Waste MEASURES          Management System is not OPERABLE for more than 31 days or if the systemwas not in use and the calculated dose values were found to be above those given in the RFO.      This allows the NRC to be kept informed of any inoperable or unused equipment needed to reduce the dose and the corrective actions being taken to restore it to OPERABLE status.
SURVEILLANCE      SR 6.2.1.3.1 REQUIREMENTS To determine if use of the installed equipment is necessary, the cumulative liquid effluent releases over the ensuing 31 days must be projected.      These releases should include all plant effluents from all Liquid Radioactive Waste Management and Liquid Waste Disposal System components that are planned to be operated.      These releases should include a margin, based on operating data, for anticipated and unplanned operational occurrences.      The ODCM methodology used to project the dose shall be the same as that used for SR 6.2.1.2.1. The impact of this projected cumulative release is to be compared to 0.06 mrem for the whole body and 0.2 Icontnued (conti nued)
Columbia Generating Station          B 6.2.1.3-2          Amendment 38, APRIL 2003
 
Liquid Waste Management System B 6.2.1.3 BASES SURVEILLANCE      SR 6.2:1.3.1      (continued)
REQUIREMENTS mrem for any organ.      If the projection indicates these values will be exceeded, then the installed Liquid Waste Management System components that will reduce these radioactive materials in the liquid effluent and their projected dose impact must be used.
SR 6.2.1.3.2 The Liquid Waste Management System is considered OPERABLE if the requirements of RFO 6.2.1.1, Liquid Concentration and RFO 6.2.1.2, Liquid Dose are met.      RFO 6.2.1.1 limits the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS.      These low concentration limits are an indicator of a properly functioning Liquid Waste Management System.      In a similiar manner, RFO 6.2.1.2 limits the dose from liquid effluents.      A low dose from liquid effluents is an indication of an OPERABLE Liquid Waste Management System.
REFERENCES        1. 10 CFR 50.
: 2. NUREG-0473,  Revision 2.
: 3. NUREG-0133.
: 4. NUREG-1302.
: 5. FSAR Section 15.0.3.
: 6. FSAR Section 11.2.
Columbia Generating Station          B 6.2.1.3-3              Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 B 6.2  RADIOACTIVE    EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.1  Dose Rate BASES BACKGROUND          This ODCM Requirement for OPERABILITY (RFO) is provided to ensure the dose rate at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits of Ref. 1. The annual dose limits are the dose rates associated with the concentrations of Appendix B, Table II, Column 1 of Ref. 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposures at or beyond the SITE BOUNDARY, that exceed the limits specified.
For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, are provided in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the whole body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also restrict, at all times the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.
APPLICABLE          The gaseous effluent dose rate limits provide assurance SAFETY ANALYSES    that any gaseous effluent releases caused by normal operation and Anticipated Operational Transients result in very low exposures to the general public. A hypothetical person living at or beyond the SITE BOUNDARY would receive a whole body dose of <= 500 mrem/year. Chapter 15, Accident Analysis, of the FSAR (Ref. 2) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent dose rates. Further information on gaseous radioactive releases can be found in Chapter 11 of the FSAR (Ref. 3).
Estimated gaseous releases from the plant are given in Ref. 4.
(continued)
Columbia Generating Station          B 6. 2.2 .1-1      Amendment 38, APRIL 2003
 
Dose Rate B 6.2.2. 1 BASES REQUIREMENTS      This RFO provides the dose rate limit.s for FOR OPERABILITY    gaseous radioactive materials released from the plant.
APPLICABILITY      This RFO is applicable at a)) times since gaseous effluents can be released at any time.
COMPENSATORY      The dose rate resulting from the release of gaseous MEASURES            radionuclides can be decreased by decreasing the amount of radioactivity being released. The amount of radioactivity will decrease with time due to decay. The use of the charcoal beds greatly decreases the Iodine and Particulates
                    .released and will delay the release of Noble Gases. This wil~l decrease the amount of activity available for release.
Other actions that would result in a decreased source of radioactivity should be considered. The Reactor Building Sump Vent Exhaust Filter System can be used to decrease the amount of radioactivity released.
SURVEILLANCE      These surveillance requirements define the radioactive REQUIREMENTS      gaseous waste sampling and analysis needed to determine the radionuclide concentrations and dose rates in the effluent.
This program meets the requirements of General Design Criteria 64 in Appendix A of Ref. 5. It also conforms to the Refs. 6 and 7.
Table 6.2.2.1-1 defines the radioactive gaseous waste sampling and analysis program. The table is divided into four sections: 1) those sampling and analysis requirements that are required to be in place at the plant at all times,
: 2) those requirements applicable during purge and vent operations, 3) additional sampling and analysis required during startup and shutdown, and 4) additional requirements that must be followed during significant reactor power changes.
Radioactive Gaseous Sampling and Analysis Regui red at All Times (Table 6.2'.2.1-1. Part 1)
A primary containment grab sample for Tritium is required to be taken and analyzed every 31 days. This analysis provides information on any releases that might be occurring within the primary containment structure. 'The 31 day sampling (continued)
Columbia Generating Station        B62212Aedet3,MR20 B 6. 2 . 2. 1-2      Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 BASE S SURVEILLANCE        period is sufficient to identify trends during power REQUIREMENTS        operation. Footnote (a) recognizes the plant configuration (continued)      when the primary containment is open and the reactor building becomes the primary confinement space. In this case the grab sampling being done at the Reactor Building Elevated discharge is representative of all space inside the Reactor Building including primary containment.
A reactor building elevated discharge grab sample *for tritium is required every 7 days. This analysis determines the tritium release in the ventilation exhaust from the spent fuel pool area. It is required whenever spent fuel is in the fuel pool.
A reactor building elevated discharge noble gas grab sample and analysis *for gamma emitters is required every 31 days.
The principal gamma emitters for noble gas effluents are defined under the LLD discussion in this bases. This analysis identifies the gamma emitters in the gaseous effluent stream that are~needed to perform the dose rate calculations. A 31 day sampling and analysis period is sufficient to characterize this release.
Turbine and radwaste building vent noble gas and tritium grab samples and analysis for principal gamma emitters and tritium are required every 31 days. The principal gamma emitters for noble gas effluents are defined under the LLD discussion in this bases. This analysis identifies the gamma emitters and tritium in the gaseous effluent stream that is needed to perform the dose rate calculations. The tritium analysis provides information on the trend of steam leaks in these two areas. A 31 day analysis period is sufficient to characterize this release.
Continuous charcoal filter sample s are colle~cted from all three building vent gaseous effluent streams. Every seven days the charcoal filters are to be removed and analyzed.
This analysis provides information on the halogen (Iodine) isotopes being released from the plant and is used to calculate the halogen dose rate due to gaseous effluents.
(continued)
Columbia Generating Station        B62213Aedet3,MR20 B 6.2.2.1-3            Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 BASES SURVEILLANCE        Continuous particulate filter samples are collected from all REQUIREMENTS        three building vent gaseous effluent streams. Every seven (continued)      days these particulate filters are removed and analyzed.
This analysis provides information on the gamma emitters collected in the sample.
A composite of the continuous particulate filter sample is collected from all building vents. Every 31 days these samples are totalized for gross alpha. Every 92 days these composite samples are analyzed for strontium. These analyses provide information on possible trends in fuel performance.
The noble gas monitor provides a continuous record of noble gas release from the plant. The gross gamma and beta count rates along with the periodic grab samples are used to characterize the noblegas release and calculate the dose rate.
Footnote (b) notes the importance of the ratio of sample flow rate to the sampled effluent stream flow. This provides one of several values needed for estimating the dose rate.
Radioactive Gaseous Sampling and Analysis Required During PURGE or VENT (Table 6.2.2.1-1. Part 2)
Primary Containment grab sampling and analysis is not required if a VENT release (by the two inch exhaust lines) is through the Standby Gas Treatment System and the noble gas monitors in containment are reading below their setpoint as stated in footnote (d). Under these conditions, grab sampling is not required as the gaseous effluent will be released at a slow rate, will be filtered, and will be measured by the Reactor Building Elevated Discharge noble gas, iodine, and particulate monitors. Additional requirements on VENTING and PURGING are found in RFO 6.2.2.6.
During other PURGE or VENT conditions, primary containment noble gas, iodine, and particulate grab samples are required prior, to each PURGE or VENT. The sample must be taken no more than 8 hours before the PURGE or VENT. The sample must be analyzed for gamma emitters prior to the PURGE or VENT.
This analysis determines the status of radioactivity in the primary containment atmosphere prior to the release.
(continued)
Columbia Generating Station        B 6.2.2.1-4            Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 BASES SURVEILLANCE        If the Reactor Building Elevated Discharge Low Range REQUIREMENTS        Radiation Monitor is not OPERABLE, sampling and analysis of (continued)        the containment shall be completed at least once per 12 hours during PURGE AND VENT. With an inoperable Reacto.r Building Elevated Discharge Low Range Radiation Monitor this additional sampling provides additional data to characterize the release from containment. This additional sampling is required for a VENT release through the Standby Gas Treatment System even when the noble gas monitors in containment are reading below their setpoint if the VENT exceeds 12 hours.
Radioactive Gaseous Sampling and Analysis Required During Startup and Shutdown (Table 6.2.2.1-1. Part 3)
Noble gas, iodine, and particulate grab samples are required from the Primary Containment and the Reactor Building following startup and shutdown. These grab samples are analyzed for gamma emitters to provide information about any changes in gaseous activity during large power changes.
Grab samples in the reactor building vent are normally taken during startup following the start of the Steam Jet Air Ejectors and shutdown of the Mechanical Vacuum Pump.      During shutdown, reactor building vent samples are taken after the Steam Jet Air Ejector is turned off and transition is made to the Mechanical Vacuum Pump. Taking grab samples at this time allows a better definition of the radioactivity being released during this transition.
Additional steps must be taken during Startup and Shutdown to determine if additional sampling is required. The first step is taken between 2 to 6 hours after the beginning of a startup or shutdown. At this point, the DOSE EQUIVALENT 1-131 concentration in the primary coolant is evaluated.      If it is less than I.OE-03 blCi/cc or if it is above I.OE-03 pCi/cc but has not changed more than a factor of three, additional sampling is not required because of high iodine concentration in the primary coolant. The low concentration or lack of change of concentration in the primary coolant is an indication of stable fuel performance that does not require additional effluent sampling.
The second step to determine if additional sampling is required involves the building vent Noble Gas Monitors. If all of the building vent monitors are less than or equal (continued)
Columbia Generating    Station      B 6.2.2.1-5            Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 BASES SURVEILLANCE        to two percent of their setpoint or if they are greater than REQUIREMENTS        two percent of the setpoint but have not increased more than (continued)      a factor of three then additional tests are not required.      A high and changing noble gas radioactivity reading from the building monitors during startup or shutdown operation could be an indication of an unusual condition within that building and additional sampling is required.
If the high and changing primary coolant radioactivity or high and changing building vent activity triggers additional testing, filters shall be removed from the continuous sampler at least once per 24 hours. This additional sampling must take place for at least seven days following shutdown or startup. The analyses shall be completed within 48 hours of filter removal. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. Once this condition is entered it must be followed to completion for the seven day time period even if the primary coolant or building vent radioactivity levels decrease below the trigger point.
Footnote (b)  notes the importance of the ratio of sample flow rate to  the sampled effluent stream flow. This provides one  of several values needed for estimating the dose rate.
Radioactive Gaseous Sampling and Analysis Required During Reactor Power Changes (Table 6.2.2.1-1. Part 4)
Following THERMAL POWER changes exceeding 15% of RATED THERMAL POWER within a 1-hour period a determination must be made to see if radioactive grab sampling is required from the Containment and Reactor Building Elevated Discharge.
If the main condenser offgas pre-treatment monitor is indicating less than or equal to 15,000 pCi/sec, no sampling is required. To convert the pretreatment monitor reading from mr/hr to A.Ci/sec multiply the average monitor reading from OG-RR-604 in mr/hr by the Offgas System flow rate from OG-FR-620 in cfm. A low pre-treatment monitor reading is an indication of stable fuel performance during Reactor Power Changes that does not require sampling. If the main (continued)
Columbia Generating Station          B 6.2.2.1-6,          Amendment 35,  MAR 2002
 
Dose Rate B 6.2.2.1 BASES SURVEILLANCE        condenser offgas pre-treatment monitor is greater than REQUIREMENTS        15,000 pCi/sec, Noble Gas, Iodine, and Particulate Grab (continued)      samples are required from the Primary Containment and the the Reactor Building.
Additional steps must also be taken during Reactor Power changes to determine if additional continuous filter sampling is required. Between 2 to 6 hours after the beginning of a reactor power change, with the off-gas pre-treatment monitor greater than 15,000 4Ci/sec, the DOSE EQUIVALENT 1-131 concentration in the primary coolant is evaluated. If it is less than the I.OE-03 pCi/cc or if it is above I.OE-03 pCi/cc but has not changed more than a factor of three additional sampling is not required because of high iodine concentration in the primary coolant.        The low concentration or lack of change of concentration in the primary coolant is an indication of stable fuel performance that does not require additional effluent sampling.
The second step to determine if additional continuous filter sampling is required involves the building vent Noble Gas Monitors. If all of the building vent monitors are less than or equal to two percent of their setpoint or if they are greater than two percent of the setpoint but have not increased more than a factor of three, then additional tests are not required.      A high and changing noble gas radioactivity reading from the building monitors during startup or shutdown operation could be an indication of an unusual condition within that building and additional sampling is required.      If the DOSE EQUIVALENT 1-131 concentration and the building vent monitors meet the criteria above, no additional continuous filter sampling is required even if the main condenser off-gas pre-treatment monitor reading is greater than 15,000 MCi/sec.
However, a high and changing primary coolant radioactivity or high and changing building vent activity requires additional continuous filter testing,      Filters shall be removed from the continuous samplers in all building vents at least once per 24 hours.      This additional sampling must take place for at least seven days following THERMAL POWER changes as defined above.      The analyses shall be completed (continued)
Columbia Generating Station          8 6.2.2. 1-7            Amendment  35,  MAR 2002
 
Dose Rate B 6.2.2.1 BASES SURVEILLANCE        within 48 hours of filter removal.      When samples collected REQUIREMENTS          for 24 hours are analyzed, the corresponding LLDs may be (continued)        increased by a factor of 10. Once this condition is entered it must be Followed to completion for the seven day time period even if the primary coolant or building vent radioactivity levels decrease below the trigger point.
Footnote (b) notes the importance of the ratio of sample flow rate to the sampled effluent stream flow.        This provides one of several parameters needed for estimating the dose rate.
Sample Lower Limit of Detection The LLD is    defined in the definitions section of the ODCM.
The principal gamma emitters for which the LLD specification applies in this RFO includes the following radionuclides:
Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, 1-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases.        Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.
Detailed discussion of the LLD and other detections limits can be found in References 8, 9 and 10.
REFERENCES          1.      10 CFR 20.1-20.602.
: 2.      FSAR Section 15.0.3.
: 3.      FSAR Section 11.3.3.
: 4.      FSAR Section 11.3.3.3.
: 5.      10 CFR 50.
: 6.      Regulatory Guide 1.21.
: 7.      Regulatory Guide 4.15.
(continued)
Columbia Generating Station            B 6.2.2..1-8            Amendment 35, MAR 2002
 
Dose Rate B 6.2.2.1 BASES REFERENCES        8. HASL Procedures Manuals,  HASL-300 (revised annually).
(continued)
: 9. Currie, L.A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).
: 10. Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
Columbia Generating Station        B 6.2.2.1-9          Amendment 35, MAR 2002
 
Dose -*Noble Gases B 6.2.2.2 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.2  Dose  -  Noble Gases BASES BACKGROUND            This ODCM Requirement for OPERABILITY (RFO) requires a calculation of the noble gas air dose resulting from gaseous effluent release. It implements the requirements of Sections II.B, III.A, and IV.A of Appendix I, of Ref. 1. It also implements Technical Specification Sections 5.5.4.e and 5.5.4.h. These RFO limits apply at any location at or beyond the site boundary.
APPLICABLE            This air dose calculation provides assurance. that any SAFETY ANALYSES      gaseous noble gas radioactive releases caused by normal operation and Anticipated Operational Transients are evaluated to determine their impact on a MEMBER OF THE PUBLIC. Chapter 15, Accident Analysis, of the FSAR (Ref. 2) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent releases. The FSAR (Ref. 3) lists a number of events that result in steam beihg discharged to the suppression pool. The non-condensible noble gases present in this steam will migrate from the primary containment to the Reactor Building and be released through the Reactor Building Elevated Release vent. The Columbia Generation Station Environmental Report (Ref. 4) calculates the estimated annual exposure due to gaseous effluent release.
REQUIREMENTS          The dose calculation methodology and parameters in the ODCM FOR OPERABILITY      implement the requirements of Section III.A of Appendix I of Ref. 1. Conformance with the guidance of Appendix I must be shown by calculational procedures based on models and data,
                    ,such that the actual exposure at locations at or beyond the SITE BOUNDARY is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of noble gases in gaseous effluents are consistent with the methodology provided in Refs. 5 and 6.
(continued)
Columbia Generating Station            B 6.2.2.2-1            Amendment 35, MAR 2002
 
Dose - Noble Gases B 6.2.2.2 BASES APPLICABILITY      This RFO is applicable at all    times since gaseous effluents can be released at any time.
COMPENSATORY        The general note is provided to implement the guidance in MEASURES            Ref. 7. This NUREG states that as long as a nuclear plant operates at a level below twice the reporting requirement of this RFO no extra analysis is required to demonstrate compliance with RFO 6.2.3.1. If the dose levels cited in this RFO are exceeded by a factor of two, additional analysis must be performed to determine if additional limitations on plant operation will be necessary to ensure compliance with Ref. 8.
The COMPENSATORY MEASURES statements provide the required operating flexibility and at the same time implement the guides.set forth in Section IV.A of Appendix I of Ref. 1 to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."
SURVEILLANCE      The surveillance requirement for this RFO requires a dose REQUIREMENTS      calculation every 31 days to provide assurance that the air dose due to noble gases does not result in a dose above the limits stated. The dose is calculated for the quarter and year based on actual and projected effluents for the 31 day period being analyzed.      All calculated doses must be below the limits given in this RFO or the COMPENSATORY MEASURES must be implemented.
REFERENCES          1. 10 CFR 50.
: 2.      FSAR Section 15.0.3.
: 3. FSAR Table 11.3-10 Section 11.3.3.3.
: 4. Columbia Generating Station Environmental    Report, Operating License Stage, Table 5.2-14.
(continued)
Columbia Generating Station          B 6.2.2.2-2            Amendment 35, MAR 2002
 
Dose - Noble Gases B 6.2.2.2 BASES REFERENCES        5. Regulatory Guide 1.109.
(continued)
: 6. Regulatory Guide 1.111.
: 7. NUREG-0543.
: 8. 40 CFR 190.
Columbia Generating Station        B 6.2.2.2-3  Amendment 35, MAR 2002
 
Dose  -    1-131,  1-133,  Tritium, and Radionuclides in Particulate Form B 6.2.2.3 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.3  Dose 131,    1-133,, Tritium, and Radionuclides  in Particulate Form BASES BACKGROUND            This ODCM Requirement for OPERABILITY (RFO) requires a calculation of the dose resulting from gaseous effluent release containing 1-131, 1-133, Tritium, and Radionuclides in Particulate Form (I-T-P). It implements the requirements of Sections II.C, III.A, and IV.A of Appendix I, of Ref&#xfd; 1.
It also implements Technical Specification Sections 5.5.4.e and 5.5.4.i. These RFO limits apply to a MEMBER OF THE PUBLIC. The dose calculation models are, therefore, allowed to take into account the additional diffusion and other source depletion between the point of release and locations occupied by individuals in the UNRESTRICTED AREA.
APPLICABLE            This organ dose calculation provides assurance that any SAFETY ANALYSES      I-T-P radioactive releases caused by normal operation and Anticipated Operational Transients are evaluated to determine their impact on a MEMBER OF THE PUBLIC. Chapter 15, Accident Analysis, of the FSAR (Ref. 2) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased I-T-P gaseous effluent releases. The Columbia Generating Station Environmental Report (Ref. 3) calculates the estimated annual exposure due to gaseous effluent release.      Gaseous Waste Management System radioactive releases are discussed in Ref. 4.
REQUIREMENTS          The dose calculation methodology and parameters in the ODCM FOR OPERABILITY      implemen't the requirements of Section III.A of Appendix I of Ref. 1. Conformance with the guidance of Appendix I must 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 equations specified in the ODCM for calculating the doses due to the actual release rates of I-T-P in gaseous effluents are consistent with the methodology provided in Refs. 5 and 6. The release rate (continued)
Columbia Generating Station            B 6.2.2.3-1            Amendment 35, MAR 2002
 
Dose 131,  1-133, Tritium,  and Radionuclides in Particulate Form B 6.2.2.3 BASES REQUIREMENTS        specifications for I-T-P are dependent upon the existing FOR OPERABILITY    radionuclide pathways to man in UNRESTRICTED AREAS.      The (continued)      pathways that were examined in the development of these calculations were:    (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animal's 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.
APPLICABILITY      This RFO is applicable at all times since I-T-P effluents can be released at any time.
COMPENSATORY        The general note is provided to implement the guidance in MEASURES            Ref. 7. This NUREG states that as long as a nuclear plant operates at a level below twice the reporting requirement of this RFO no extra analysis is required to demonstrate compliance with RFO 6.2.3.1. If the dose levels cited in this RFO are exceeded by a factor of two, additional analysis must be performed to determine if additional limitations on plant operation will be necessary to ensure compliance with Ref. 8.
The COMPENSATORY MEASURES statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I of Ref I to assure that the releases of I-T-P radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."
SURVEILLANCE        The surveillance requirement for this RFO requires a dose REQUIREMENT        calculation every 31 days to provide assurance that the organ dose due to I-T-P does not result in a dose above the limits stated. The dose is calculated for the quarter and year based on actual and projected effluents for the 31 day period being analyzed. All calculated doses must be below the limits given in this RFO or the COMPENSATORY MEASURES must be implemented.
(continued)
Columbia Generating Station          B 6.2.2.3-2            Amendment 35,  MAR 2002
 
Dose  -  1-131,  1-133,  Tritium,  and Radionuclides in Particulate Form B 6.2.2.3 BASES  (continued)
REFERENCES          I  . 10 CFR 50.
: 2.      FSAR Section 15.0.3.
: 3. Columbia Generating Station Environmental      Report, Operating License Stage, Table 5.2-14.
: 4. FSAR Section 11.3.3.
: 5. Regulatory Guide 1.109.
: 6. Regulatory Guide 1.111.
: 7. NUREG-0543.
: 8. 40 CFR 190.
Columbia Generating Station              B 6.2.2.3-3            Amendment 35, MAR 2002
 
GASEOUS OFFGAS TREATMENT SYSTEM B 6.2.2.4 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.4  GASEOUS OFFGAS TREATMENT SYSTEM BASES BACKGROUND          This ODCM Requirement for OPERABILITY (RFO) requires the GASEOUS OFFGAS TREATMENT SYSTEM be available for use in either the normal or charcoal bypass mode of operation.      The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."    This Requirement for OPERABILITY implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A of Ref. 1. It also implements the design objectives given in Section 1I.D of Appendix I of Ref I and Technical Specification Section 5.5.4.f. Further guidance on the implementation of this RFO is provided in Refs. 2, 3 and 4.
A GASEOUS OFFGAS TREATMENT SYSTEM is defined as 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.
APPLICABLE          This RFO provides assurance that gaseous radioactive SAFETY ANALYSES    releases from the main condenser when the steam jet air ejectors are in service are monitored or processed. These radioactive releases occur as a result of normal operation or could be caused by Anticipated Operational Transients.
Chapter 15, Accident Analysis, of the FSAR (Ref. 5) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent releases.
(continued)
Columbia Generating Station          B 6.2.2.4-1          Amendment 35, MAR 2002
 
GASEOUS OFFGAS TREATMENT SYSTEM B 6.2.2.4 BASES REOUIREMENTS        The OPERABILITY of the GASEOUS OFFGAS TREATMENT SYSTEM FOR OPERABILITY    ensures that this system will be available for use whenever these gaseous effluents require treatment prior to release to the environment. The term "gaseous offgas treatment system" involves all of the installed and available gaseous radioactive waste management system equipment, as well as their controls, power, instrumentation, and .services that make the system functional. Equipment that is considered standby or redundant is also included, since their function is to assure OPERABILITY. The RFO also permits alternate treatment paths using alternate subsystems and equipment to be used in the event that the normal treatment equipment is inoperable. In the bypass mode of operation the offgas post-treatment monitors must be operational and capable of automatically changing to normal mode of operation if their setpoint is exceeded. Further information on the design bases, system description, and process description of the Gaseous Waste Management System is given in Ref. 6.
APPLICABILITY      This RFO is applicable when the main condenser steam jet air ejector (evacuation) system is in operation.      During plant startup, prior to the use of the steam jet air ejector, the mechanical vacuum pump is used to remove air from the condenser.
COMPENSATORY        The normal mode of operation of the GASEOUS OFFGAS TREATMENT MEASURES            SYSTEM is with the charcoal beds in operation. This provides for the removal of most iodine and particulate radioactive isotopes and delays the release of noble gases.
If the system is not being operated in the normal mode compensatory measure A.1 requires it to be placed in the normal mode in 7 days.
If the required compensatory measure A.1 and the associated 7 day completion time is not met a Special Report must be submitted to the NRC that identifies the required inoperable equipment and the reasons for the inoperability, corrective actions taken to restore the required inoperable equipment to OPERABLE status, and the corrective actions to prevent recurrence.
(continued)
Columbia Generating Station          B 6.2.2.4-2            Amendment 35, MAR 2002
 
GASEOUS OFFGAS TREATMENT SYSTEM B 6.2.2.4 BASES COMPENSATORY        The NRC requires the report once the GASEOUS OFFGAS MEASURES            TREATMENT SYSTEM is not in the normal mode of operation for (continued)        more than 7 days. The report must be completed 30 days after condition B is entered.
SURVEILLANCE        SR 6.2.2.4.1 REQUIREMENTS This surveillance provides verification of GASEOUS OFFGAS TREATMENT SYSTEM operation in either the normal mode or the bypass mode provided the offgas post-treatment radiation monitor is OPERABLE every 7 days.      The main purpose of this surveillance is to provide assurance that the system is being operated in the normal mode with the charcoal beds in service.
REFERENCES        1. 10 CFR 50.
: 2. NUREG-0473,  Rev. 2.
: 3. NUREG-0133.
: 4. NUREG-1302.
: 5. FSAR Section 15.0.3.
: 6. FSAR Section 11.3.
Columbia Generating Station          B 6.2.2.4-3            Amendment 35, MAR 2002
 
VENTILATION EXHAUST TREATMENT SYSTEM B 6.2.2.5 I
B 6.2  RADIOACTIVE EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.5  VENTILATION EXHAUST TREATMENT SYSTEM                                    I BASES BACKGROUND          The Gaseous Waste Management System is designed to ensure that gaseous radioactive effluents are as low as reasonably achievable (ALARA). This ODCM Requirement for OPERABILITY-(RFO) for the VENTILATION EXHAUST TREATMENT SYSTEM is in accordance with the Standard Radiological Effluent Technical Specifications for BWRs (Refs. 1 and 2). Examples of the VENTILATION EXHAUST TREATMENT SYSTEM would include the following components: TG 441 Sample Hood Exhaust HEPA; Reactor Building Sump Vent Exhaust Filter System; CRD Rebuild Room Exhaust HEPA; Sampler & Analyzer Room 1A Exhaust HEPA; Radwaste Building Exhaust Ventilation HEPAs; and Chemistry Lab Sample Hood Exhaust HEPA (Ref. 3). The system(s) appropriate for reducing the gaseous radioactive effluents must be in service if the project 31 day organ dose is greater 0.3 mrem. The determination of the appropriate system to be in service is based on the source of the gaseous radioactive effluent.
The OPERABILITY of the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever
                                                                                    .I gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable."
This RFO implements the requirements of Part 50.36a, General Design Criterion 60 of Appendix A and the design objective given in Section II.D of Appendix I of Ref. 4. The specified limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I of Ref. 4 for gaseous effluents.
(continued)
Columbia Generating Station        B 6.2.2.5-1I          Amendment 47, AUG 2008
 
VENTILATION EXHAUST TREATMENF SYSTEM B 6.2.2.5 I
BASES APPLICABLE          The Gaseous Waste Management System is designed to ensure SAFETY ANALYSES    that gaseous radioactive effluents are as low as reasonable achievable (ALARA). This RFO requires that the VENTILATION EXHAUST TREATMENT SYSTEM be maintained OPERABLE with the appropriate systems in service to provide assurance that gaseous radioactive releases caused by normal operation and Anticipated Operational Transients are treated.      Chapter 15, Accident Analysis, of the FSAR (Ref. 5) analyzes a number of      I anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent releases.
REQUIREMENTS      This RFO has two parts. The first requirement is for FOR OPERABILITY    the system to be operable and the second is for the system to be used. The VENTILATION EXHAUST TREATMENT SYSTEM          I involves all of the installed and available system equipment, as well as their controls, power, instrumentation, and services that make the system functional. Equipment that is considered standby or redundant is also included, since their function is to assure OPERABILITY.
The appropriate VENTILATION EXHAUST TREATMENT SYSTEM is to.
be used whenever the projected gaseous effluent 31 day organ dose is above 0.3 mrem. This dose is calculated for all gaseous releases, from all buildings. The 0.3 mrem organ dose specified in this RFO is a fraction of the quarterly limit of RFO 6.2.2.3.
APPLICABILITY      This RFO is applicable at all times as the VENTILATION EXHAUST TREATMENT SYSTEM may be required at any time.
COMPENSATORY      If the VENTILATION EXHAUST TREATMENT SYSTEM is not in MEASURES          service when the projected 31 day organ dose is above 0.3 mrem a Special Report must be written to the NRC in the next 30 days. This 31 day organ dose calculation is required by SR 6.2.2.5.1.
(continued)
Columbia Generating Station        B 6.2.2.5-2            Amendment 47, AUG 2008
 
VENTILATION EXHAUST TREATMENT SYSTEM B 6.2.2.5 I
BASES SURVEILLANCE      SR 6.2.2.5.1 REQUIREMENTS This surveillance requires a projection of the 31 day organ dose every 31 days. This dose calculation is used to determine if the VENTILATION EXHAUST TREATMENT SYSTEM must be used during that time period.
I SR 6.2.2.5.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by complying with RFOs 6.2.2.1 and 6.2.2.3. These RFOs provide the dose and dose rate limits applicable to the iodine and particulate gaseous radioactive releases that the VENTILATION EXHAUST TREATMENT SYSTEM is designed to treat.
REFERENCES        1. NUREG-0473,  Rev. 2.
: 2. NUREG-1302.
: 3. FSAR Section 9.4.
I
: 4. 10 CFR 50.
: 5. FSAR Section 15.0.3.
Columbia Generating Station        B 6.2.2.5-3            Amendment 47, AUG 2008
 
VENTING or PURGING B 6.2.2.6 B 6.2  RADIOACTIVE EFFLUENTS B 6.2.2  Gaseous Effluents B 6.2.2.6  VENTING or PURGING BASES BACKGROUND          This ODCM Requirement for OPERABILITY (RFO) provides assurance that releases from primary containment during VENTING or PURGING operations will be as low as reasonably achievable. This RFO implements the requirement of Appendix I to Ref. I and Technical Specification' Section 5.5.4.k.        A description of the PURGE and VENT System is provided in Ref.
: 2. Sampling requirements during PURGE and VENT operation are given in RFO 6.2.2.1, Gaseous Effluents-Dose Rate.
APPLICABLE          This RFO provides assurance that gaseous radioactive SAFETY ANALYSES      releases from the containment caused by normal operation and Anticipated Operational Transients are filtered during the first 24 hours of release. Chapter 15, Accident Analysis, of the FSAR (Ref. 3) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased gaseous effluent releases to containment.
REQUIREMENTS        This RFO provides three different conditions that must be FOR OPERABILITY    considered during VENTING and PURGING. The first two state.
the requirements that must be met during the first 24 hours of VENTING and PURGING operation when the Standby Gas Treatment (SGT) must be used to filter the releases from the Primary Containment. The SGT train(s) being used for filtration must be "functional for filtration."        SGT functionality as used in this RFO shall include the ability to accept the exhaust gases from primary containment, process the gases through the filtration unit, and discharge the gases to either the reactor building stack or the reactor building.
In Modes 1, 2, or 3, the SGT train not being used for filtration is required to be OPERABLE.      The requirement to have the second train OPERABLE addresses the fact that when the 24-inch or 30-inch exhaust valves are used the (continued)
Columbia Generating Station          B 6.2.2.6-1            Amendment 35,  MAR 2002'
 
VENTING or, PURGING B 6.2.2.6 BASES REQUIREMENTS        in-service standby gas treatment system train used for FOR OPERABILITY    filtration of the exhaust gases is inoperable.. This is due (continued)      to the potential for post-LOCA over-pressurization of that train. The over-pressurization could be caused by the postulated rapid pressurization of containment during LOCA conditions before the qualified containment isolation valves close. This condition does not exist if the 2-inch valves are being used as the SGT and the valve closure are designed to maintain SGT integrity in response to post-LOCA pressurization. There are also control considerations that impact SGT operability.      The train being used for filtration requires its SGT controller to be placed in the manual mode which makes that train inoperable since it cannot automatically perform its safety related function if required. The use of manual control is required to initiate all VENTING and PURGING configurations including the use of the 2 inch lines. The assurance of an OPERABLE second SGT train avoids the potential of having the plant enter LCO 3.0.3 as a result of two inoperable SGT trains.
In Mode 4 when VENTING or PURGING for deinerting it is acceptable to have both standby gas treatment system trains in service supporting filtration for the first 24 hours of any VENTING and PURGING operation.      In Mode 4 there are no postulated accidents that could result in over-pressurization of the standby gas treatment system trains.
However, if manual control is used for both SGT trains resulting in both trains being INOPERABLE Technical Specification LCO 3.6.4.3 is applicable.      It requires no movement of irradiated fuel assemblies in secondary containment, suspension of CORE ALTERATIONS, and suspension of OPDRVs.
Following    the first 24 hours of any VENTING and PURGING operation    filtration through SGT is no longer required.      The potential    for contamination of the primary containment atmosphere    is highest during this initial time period.
APPLICABILITY      This RFO is applicable in MODES 1, 2, and 3 when VENTING and PURGING. It is also applicable when deinerting the containment.      This RFO is not APPLICABLE during primary (continued)
Columbia Generating Station          B 6.2.2.6-2            Amendment 35, MAR 2002
 
VENTING or PURGING B 6.2.2.6 BASES APPLICABILITY      containment depressurization following Type A testing (continued)      required by Ref. 4. However, during primary containment depressurization following Type A testing appropriate radioactive sampling shall be performed.
COMPENSATORY        If RFO 6.2.2.6 is not met VENTING and PURGING of the MEASURES          containment must be discontinued immediately.
SURVEILLANCE      SR 6.2.2.6.1 REQUIREMENTS This surveillance requires verification that the containment VENT and PURGE system is aligned in the correct manner.
This verification is required no more than 4 hours before the start of VENT and PURGE operations and every 12 hours thereafter. This provides assurance that equipment will function in the manner intended.
REFERENCES        1. 10 CFR 50.
: 2. FSAR Section 6.2.1.1.8.2.
: 3. FSAR Section 15.0.3.
: 4. 10 CFR 50,  Appendix J.
Columbia Generating Station        B 6.2.2.6-3            Amendment 35, MAR 2002
 
Total Dose B 6.2.3 B 6.2  RADIOACTIVE    EFFLUENFS B 6.2.3  Total Dose BASES BACKGROUND          This RFO is provided to ensure conformance to the U. S. Environmental Protection Agency (EPA) radiation protection standard contained in Ref. 1. In Ref. 2 the NRC issued guidance on how the EPA standard is to be implemented and interfaced with its own regulations. The NRC guidance states that compliance with the EPA regulations is assured if the calculated dose from liquid and gaseous effluents remains below twice the limits of Ref. 3. This RFO implements this guidance.
APPLICABLE          This dose calculation provides assurance that all SAFETY ANALYSES      radioactive releases and direct radiation caused by normal operation and Anticipated Operational Transients are evaluated to determine their impact on a MEMBER OF THE PUBLIC. Chapter 15, Accident Analysis, of the FSAR (Ref. 4) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased effluent releases and direct radiation.      The Columbia Generating Station Environmental Report (Ref. 5) provides estimated calculated annual doses for Columbia Generating Station.
REQUIREMENTS        This RFO requires a calculation of the dose or dose FOR OPERABILITY    commitment from all Uranium fuel cycle sources.      The method of calculation is given in the ODCM and uses the standard methodology required to satisfy the requirements of Ref. 6.
The dose is to be calculated for a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA beyond the SITE BOUNDARY.      This RFO adds the additional requirement that direct radiation must I
be added to the total dose. While the direct radiation from Turbine Building shine can be measurable at distances near the building it is not expected to be significant in the UNRESTRICTED AREA beyond the SITE BOUNDARY.      This calculation must also consider other direct radiation, sources such as spent fuel storage facilities. A direct measurement of the direct radiation can be obtained from Environmental Monitoring devices.
(continued)
Columbia Generating Station            B 6.2.3-1          Amendment 38, APRIL 2003
 
Total Dose B 6.2.3 BASES APPLICABILITY        This RFO is applicable at all times since effluents can be released at any time.
COMPENSATORY        This RFO is related to'RFOs 6.2.1.2, (Dose-Liquid MEASURES            Effluent), 6.2.2.2, (Dose-Noble Gases), and 6.2.2.3, (Dose-l-T-P). Condition A is entered when the dose calculated by the surveillance requirements associated with any of these RFOs reaches twice the limit specified. The compensatory measure requires a calculation of the total dose including the contribution from direct radiation. The direct radiation calculation shall include the contribution from the plant, outside storage tanks, and any "outside fuel storage facilities.
If the calculated total dose exceeds the limits of this RFO compensatory measure B.1 must be implemented. This requires a Special Report to the NRC within 30 days. The Special Report shall be submitted pursuant to Appendix I, Section IV.A of Ref. 3. The Special Report shall define the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the RFO limits including a schedule for achieving conformance with the above limits.
This Special Report must also conform to the requirements of Part 20.2203.A of Ref. 7. The special report shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources including all effluent pathways and direct radiation within a radius of eight km, for the calendar year that includes release(s) covered by the report. 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.      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 limits of thi's RFO, and if the release condition resulting in violation of Ref. I has not already been corrected, the Special Report shall include a request for a variance in accordance with the provision of Ref. 1. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.        The variance only relates to the limits of Ref. 1, and does not apply in any way to the other requirements for dose limitation of Refs. 3 and 7.
(continued)
Columbia Generating Station          B 6.2.3-2            Amendment 35,  MAR 2002
 
Total Dose B 6.2.3 BASES SURVEILLANCE      This surveillance requires a determination of the REQUIREMENTS      dose or dose commitment to any MEMBER OF THE PUBLIC in the UNRESTRICTED AREA beyond the SITE BOUNDARY from liquid and gaseous effluents. This surveillance requirement is related to the surveillance requirements 6.2.1.2.1, (Dose-Liquid Effluent), 6.2.2.2.1, (Dose-Noble Gases), and 6.2.2.3.1, (Dose-I-T-P). When the calculated dose in these surveillances reach twice the limit specified it triggers a calculation of the total dose.
REFERENCES        1. 40 CFR Part 190.
: 2. NUREG-0543.
: 3. 10 CFR 50,  Appendix I.
: 4. FSAR Section 15.0.3.
: 5. Columbia Generating Station Environmental    Report, Operating License Stage, Table 5.2-14.
: 6. Regulatory Guide 1.109.
: 7. 10 CFR Part 20.
Columbia Generating Station          B 6.2.3-3            Amendment 38, APRIL 2003
 
Radiological  Environmental  Monitoring Program B 6.3.1 B 6.3  RADIOLOGICAL ENVIRONMENTAL MONITORING B 6.3.1  Radiological  Environmental  Monitoring Program BASES BACKGROUND          The Radiological Environmental Monitoring Program (REMP) required by this Requirement for OPERABILITY (RFO) provides representative measurements of radiation and of radioactive materials in the environment near Columbia Generating Station. It includes all exposure pathways for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the plant operation. This monitoring program implements Section IV.B.2 of Appendix I to Ref. 1 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.      Guidance for this monitoring program is provided by Ref. 2.
APPLICABLE          This Radiological Environmental Monitoring Program provides SAFETY ANALYSES    assurance that radiation levels in the environment caused by normal operation and Anticipated Operational Transients are evaluated to determine their impact on a MEMBER OF THE PUBLIC. Chapter 15, Accident Analysis, of the FSAR (Ref. 3) analyzes a number of anticipated process disturbances and component failures that have the potential of causing increased environmental impact.
REQUIREMENTS        This RFO, through Table 6.3.1-1, specifies sample locations, FOR OPERABILITY    analyses, and frequencies for several different exposure pathways. The RFO is a comprehensive radiological environmental surveillance program to assess the impact of plant operation on the environment.      The results of this program are intended to supplement the results of the Radiological Effluent Monitoring Program.      The REMP verifies that the measurable concentrations of radioactive materials and levels of radiation are no higher than expected based on the effluent measurements and modeling of the exposure pathways. The program uses the guidance of Refs. 2, 4, 5, 6, and 7. The following describes the environmental sampling categories of Table 6.3.1-1.
(continued)
Columbia Generating Station            B 6 .3.1 -1            Amendment 35, MAR 2002
 
Radiological  Environmental  Monitoring Program B 6.3.1 BASES REQUIREMENTS        Direct Radiation Pathway FOR OPERABILITY (continued)      Thirty-four direct radiation monitoring stations are required. One or two of these are to serve as control stations located in area(s) not impacted by direct radiation from the plant. The other stations are to be located at various distances from the plant (as noted to Table 6.3.1-1) in order to obtain a representative measurement of possible direct radiation in the environment surrounding Columbia Generating Station.      These measurements are normally made with thermoluminescent dosimeter (TLDs).        However, one or more instruments, such as a 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 Table 6.3.1-1, a TLD is considered to be one phosphor card with multiple readout areas; a phosphor card in a packet is considered to be equivalent to two or more, dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation.      The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and selected to obtain optimum dose information with minimal fading.      TLDs are to meet the requirements of Ref. 6 except for specified energy-dependence response.
Airborne Pathway-Particulate/Iodine Samples are required from 6 locations.        One of these locations is to be a control location located in a direction not expected to be impacted by plant airborne effluents.
The other five locations are to be indicator stations located in a direction that are most likely to be impacted by plant gaseous effluents.      Samples are obtained through the use of constant flow-rate sampling units that collect both particulate and iodine samples.        Gross beta, gamma isotopic and radioiodine analysis are required.
Waterborne Pathway-Surface Water An upstream and a downstream sample are required.        At Columbia Generating Station the upstream sample is considered the control sample.      It is taken from the plant (continued)
Columbia Generating Station          B 6. 3 .1-2            Amendment 35, MAR 2002
 
Radiological  Environmental  Monitoring Program B 6.3.1 BASES REQUIREMENTS        Waterborne Pathway-Surface Water      (continued)
FOR OPERABILITY water intake which is :located upstream of the plant discharge into the river. The downstream sample requirement allows for dilution in the mixing zone associated with the plant discharge line into the Columbia River.      However, it is not feasible to perform flow-proportional composite sampling in the mixing zone area of the river downstream from the plant discharge point.      Therefore, the downstream sample is taken from the plant discharge prior to river dilution. This provides a very conservative measurement.
Gamma isotopic, gross beta, and tritium analyses are required for these samples.
Waterborne Pathway-Ground Water Groundwater samples from one or two sources are required when the source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.      At Columbia Generating Station the deep well just north of the Reactor Building is sampled as it is a backup source of drinking water at the plant.      Two wells at WNP-1 that provide drinking water for that site are also sampled.        Gamma isotopic and tritium analysis are required.
Waterborne Pathway-Drinking Water One sample is required of each of one to three of the nearest water supplies that could be affected.        A sample is also required from a control location.        The primary indicator station for drinking water is the City of Richland Water Treatment Plant that obtains its drinking water from the Columbia River.      The control location is the plant intake water sampler that provides a sample of the Columbia River water upstream of the discharge line.        Gross beta, gamma isotopic, and tritium analysis is required periodically. If the calculated dose from the consumption of water is greater than I mrem per year 1-131 analysis is also performed.
(continued)
Columbia Generating Station          B 6.3 .1-3              Amendment 35,  MAR 2002
 
Radiological  Environmental  Monitoring Program B 6.3.1 BASES REQUIREMENTS        Waterborne Pathway  - Sediment From Shoreline FOR OPERABILITY (continued)      One sample is required from a downstream area with existing or potential recreational value. This sample is collected approximately one mile downstream from the plant effluent line. Gamma isotopic analysis is required of this sample.
InQestion-Milk Samplesare required from milking animals in three locations within five kilometers having the highest dose potential.
At Columbia Generating Station there are no milking animals at this distance as the area inside a five kilometer radius is all on the Hanford Site. In this case, a sample is required from milking animals in each of three areas between five and sixteen kilometers distant where doses are calculated to be greater than one mrem per year to the maximum organ and age group. Even at this distance, the availability of samples is limited and may not be available at all times. A sample from milking animals is also required at a control location, 30-50 kilometers distant and in the least prevalent wind direction.      Indicator samples are obtained from cattle across the Columbia-River in Franklin County. The control sample is obtained from the Sunnyside/Grandview area. However, it is sometimes difficult to obtain a control sample because most of the cattle in this area use feed from the area east of the plant. This invalidates this milk as a control sample.
Gamma isotopic and 1-131 analysis are required.        In addition, if Cs-134 or Cs-137 is measured in an individual milk sample in excess of 30 pCi/liter, then Sr-90 analysis shall be performed.
Ingestion-Fish One sample of each of three recreationally important species (one anadromous and two resident) is required from the vicinity of the plant discharge.      A control sample, of the same three species is required from an area not influenced by the plant discharge. Resident samples are taken from the vicinity of the plant discharge.      Anadromous species are obtained from the Ringold Fish Hatchery.      Control samples for resident fish are collected from the Snake River.
(continued)
Columbia Generating Station          B 6.3.1-4              Amendment 35,  MAR 2002
 
Radiological  Environmental Monitoring Program B 6.3.1 BASES REQUIREMENTS        Ingestion-Fish    (continued)
FOR OPERABILITY Anadromous control samples are obtained from the Lyon's Ferry Fish Hatchery on the Snake River. Gamma isotopic analysis is required on the edible portions.
Ingestion-Food  Products One sample of each principal class of food products is required from any area that is irrigated by water in which liquid plant wastes have been discharged.      In'addition samples of three different kinds of broad leaf vegetation grown nearest each of two different o ffsite locations of highest predicted annual average ground-level D/Q are required if milk sampling is not performed.      A control sample of broad leaf vegetation grown 30-50 kilometers in the least prevalent wind direction if milk sampling is not performed is also required. Garden produce is routinely obtained from farms or gardens using Columbia River water for irrigation. One sample of a root crop, leafy vegetable, and a fruit should be collected each sample period if available. Gamma isotopic analysis is required for all samples. In addition, broad leaf vegetation requires an 1-131 analysis if milk sampling is not performed.
APPLICABILITY      This RFO is applicable at all times since effluents can be released at any time.
COMPENSATORY        Required Compensatory Measure A.1 requires any deviations MEASURES            from the program to be reported in the next annual Radiological Environmental Operating Report. This allows the NRC to be aware of any samples not collected and analyzed in compliance with the program detailed in Table 6.3.1-1.
The Condition B note allows separate condition entry for each milk and fresh leafy vegetable sample.      If one or more of the milk and leafy vegetable samples are unavailable, a location for replacement samples must be identified within 30 days. The new sample locations along with revised tables and figures must be submitted in the next Radiological Environmental Operating Report.      The NRC considers the sampling of milk and fresh leafy vegetables important and thus requires replacement sample locations to be identified within 30 days.
(continued)
Columbia Generating Station            B 6.3.1-5            Amendment 35,  MAR 2002
 
Radiological Environmental Monitoring Program B 6.3.1 BASES COMPENSATORY        The Condition C note allows separate condition entry MEASURES            for each sample. The first condition involves the detection (continued)      of the concentration of any single radionuclide in the environment exceeding the value given in Table 6.3.1-2. The concentrations given approximate the dose levels allowed by Appendix I or Ref. 1. The detection of this level of radioactivity in the environment requires a Special Report to the NRC within 30 days. If more than one radionuclide is detected, the concentration level that requires a report is decreased. The formula given in the footnote of-Table 6.3.1-2 requires a report whenever the sum of the detected concentrations divided by the concentration limits is greater than one. The Special Report must describe the corrective actions being taken to reduce the dose below those given in the Appendix I of Ref. 1 RFOs (RFOs 6.2.1.2, 6.2.2.2, and 6.2.2.3). The methodology and parameters used to estimate the potential annual dose to a member of the public shall be indicated in this report.
Condition D involves the detection of radionuclides other than those listed on Table 6.3.1-2. If these radionuclides are'detected, a dose calculation is required to determine the annual dose that would result if this concentration were present in the environment for a year. If the dose is equal to or greater than the limits given by RFOs 6.2.1.2, 6.2.2.2, or 6.2.2.3, the condition is reportable unless it can be shown that the radionuclides were not the result of Columbia Generating Station operation. The Special Report must state the corrective actions being taken to reduce the dose below the limits and must be written in 30 days. The NRC requires timely information on unexpected' radionuclides in the environment. The detection of these non Table 6.3.1. -2 radionuclides must also be reported in the Annual Radiological Environmental Operating Report. In this report, the detection of any non Table 6.3.1-2 radionuclides regardless of the calculated dose and source of the radiation is to be reported. The methodology and parameters used to estimate the potential annual dose to a member of the public shall be indicated in this report.
(continued)
Columbia Generating Station        B6316Aedet3,MR20 B 6. 3. 1 -6          Amendment 35, MAR 2002
 
Radiological  Environmental    Monitoring Program B 6.3.1 BASES  (continued)
SURVEILLANCE        This surveillance requires sampling and analysis presented REQUIREMENTS        in Table 6.3.1-1 be performed to verify all environmental radiological conditions are within limits.        All sampling and analysis is to be performed according to the frequency given in the Table.
The detection capabilities for the analysis are given in Table 6.3.1-3. The required detection capabilities for environmental sample analyses are tabulated in terms of the Lower Limits of Detection (LLDs).        The LLDs required by Table 6.3.1-3 are considered optimum for routine environmental measurements in industrial laboratories.          It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.          The formula to be used for LLD is given in the definition section of this ODCM Appendix. Detailed discussions of the LLD and other detections limits can be found in References 8, 9, and 10.
REFERENCES          1. 10 CFR 50.
: 2. NUREG-1302.
: 3. FSAR 15.0.3.
: 4. Regulatory Guide 4.1,    Rev. 1.
: 5. Regulatory Guide 4.8.
: 6. Regulatory Guide 4.13,    Rev. 1.
: 7. Regulatory Guide 4.15,    Rev. 1.
: 8. HASL Procedures Manuals,      HASL-300 (revised annually).
: 9. Currie, L.A. "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).
: 10. Hartwell, J.K. "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
Columbia Generating Station            B 6 .3. 1-7              Amendment 35, MAR 2002
 
Land Use Census B 6.3.2 B 6.3  RADIOLOGICAL ENVIRONMENTAL MONITORING B 6.3.2  Land Use Census BASES BACKGROUND          The Land Use Census required by this RFO provides assurance that changes in the use of land in the unrestricted area 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 a door-to-door survey, from an aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to Ref. 1. Guidance for the Land Use Census is provided by Ref. 2.
APPLICABLE          This Land Use Census provides assurance that changing SAFETY ANALYSES    conditions in the environment surrounding Columbia Generating Station are identified. This allows more accurate evaluation of the impact of radionuclide releases caused by normal operation and Anticipated Operational Transients. Chapter 15, Accident Analysis, of the FSAR (Ref. 3) analyzes a number of anticipated process disturbances and component failures that have the potential of causing measurable environmental impact.
REQUIREMENTS      This RFO provides for a periodic update of the location of FOR OPERABILITY    the nearest milk animal, residence, and garden greater than 500 square feet. At Columbia Generating Station the closest residence is across the Columbia River at a distance of approximately 4 miles. Because of the restrictions of the Hanford Site, the only land use area that has the potential for milk animals, residences, or gardens within 5 miles is a narrow strip of land just east of the Columbia River.
However, the Radiological Environmental Monitoring Program evaluates land use beyond the 5 mile limit. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified arid monitored since a garden of this size is the minimum required to produce the quantity
                                                                    .(continued)
Columbia Generating Station          B6321Aedet3,MR20 B 6.3.2-1            Amendment 35, MAR 2002
 
Land Use Census B 6-.3.2 BASES REQUIREMENTS        (26 kg/year) of leafy vegetables assumed in Ref. 4 for FOR OPERABILITY    consumption by a child. To determine this minimum garden (continued)      size, the following assumptions were made:
a)  20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage, and b)    a vegetation yield of 2 kg/in2 .
APPLICABILITY      The RFO is applicable at all times.
COMPENSATORY      If a Land Use Census is not conducted or if all required.
MEASURES          locations are not identified Compensatory Measure A.1 requires this condition to be reported in the next Annual Radiological Environmental Operating Report.
Condition B requires notification in the next submission of the Radioactive Effluent Release Report if the new Land Use Census identifies a new location of a milk animal, residence, or garden greater than 500 square feet and the calculated dose of dose commitment at this new location is
                  'greater than that calculated by the current census at the current locations.
Condition C is an escalation of Condition B. It occurs if one of the newly identified census locations has a calculated dose greater than 120 percent of the dose from the current locations via the same exposure pathway. In this case the new location must be added to the Radiological Environmental Monitoring Program within 30 days. The compensatory measures also state that the sample location having the lowest calculated dose can be deleted from the Radiological Monitoring Program after October 31 of the year the newly discovered location was found. In addition, the NRC is to be notified in the next Effluent Release Report.
This notification must include a revised figure(s) and table for the ODCM reflecting the new location(s).
The Compensatory Measures associated with Conditions B and C are structured to assure action is taken to evaluate and implement the results of the Land Use Census.
(continued)
Columbia Generating Station        B6322Aedet3,MR20 8 6.3.2-2              Amendment 35, MAR 2002
 
Land Use Census B 6.3.2 BASES  (continued)
SURVEILLANCE        This surveillance requires a Land Use Census to be conducted REQUIREMENTS        every year. The results of the Land Use Census are to be included in the annual Radiological Environmental Operating Report.
REFERENCES          1. 10 CFR 50.
: 2. NUREG-1302.
: 3. FSAR 15.0.3.
: 4. Regulatory Guide 1.109.
Columbia Generating Station          B 6.3.2-3          Amendment 35, MAR 2002
 
Interlaboratory  Comparison Program B 6.3.3 B 6.3  RADIOLOGICAL ENVIRONMENTAL MONITORING B 6.3.3  Interlaboratory    Comparison Program BASES BACKGROUND          The requirement for participation in an approved 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. This comparison is part of the quality assurance program for environmental monitoring to demonstrate that the sampling results are valid for the purposes of Section IV.B.2 of Appendix I to Ref. 1.
APPLICABLE          This Interlaboratory Comparison Program contributes to SAFETY ANALYSES      safety by providing assurance that environmental sampling will be accurate and precise.      This allows more accurate evaluation of the impact of radionuclide releases caused by normal operation and Anticipated Operational Transients.
REQUIREMENTS        The Interlaboratory Comparison Program is a major component FOR OPERABILITY    of the quality assurance program for the Radiological Environmental Monitoring Program.      It involves the comparison of Columbia Generating Station analytical results obtained and compared to samples containing known concentrations of various radionuclides.        The program participates in the Environmental Measurements Laboratory (EML)  intercomparison program. It also participates in local and regional intercomparison studies.
APPLICABILITY        This RFO is    applicable at all times.
COMPENSATORY        Failures associated with the Interlaboratory Comparison MEASURES            Program must be reported to the NRC in the next Radiological Environmental Operating Report.
conti nued Columbia Generating Station            B 6.3.3-1              Amendment 35, MAR 2002
 
Interlaboratory Comparison Program B 6.3.3 BASES  (continued)
SURVEILLANCE        This surveillance requires analysis of\ radioactive material REQUIREMENTS        supplied as part of the Interlaboratory Comparison Program.
REFERENCES          1. 10 CFR 50,  Appendix A.
Columbia Generating Station          B 6.3.3-2            Amendment 35, MAR 2002}}

Latest revision as of 12:50, 14 January 2025

Independent Spent Fuel Storage Installation - Offsite Dose Calculation Manual, Including Revision Made During 2008
ML090720893
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Issue date: 02/27/2009
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