Text

Offsite Dose Calculation Manual, Revision 21
	ML071220307
Person / Time
Site:	Palo Verde
Issue date:	09/15/2006
From:	Bungard J, Drinovsky L, Gaffney A, Hesser J; Arizona Public Service Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML071220307 (134)
	v • d • e

OFFSITE DOSE CALCULATION MANUAL PALO VERDE NUCLEAR GENERATING STATION UNITS 1, 2 AND 3 REVISION 21 Digitally signed by Drinovsky. Louis J Drinovsky, (Z336,)

DN: CN = Drinovsky, Louis J(Z3 99)

SReason: I m the authorof this Louis J(Z33699I dlocument

u.

rftl Date: 2006.07.25 07.0:11.0700 Originator Digitally signed by Bungard, Jaes P (Z18012),

w P(Z1 012)Date:

2006.08.11 13:51-.05 -07W0 Technical Reviewer ff Joh Digitaily signed by Gaffney, John P Gaffney, Joh n:

(Z5)

DN: CN - Gaffney. John PlZ3645g)

P(3659Reason:,I have wiewe thi Director, Radiation Pn(Z36459) doumert.

Protection Date: 20OM.08.1711:45..M.070 Digitally signed by Hesser, John H Hesser, John (Z46708)

ON: CN - Hesser, John H(Z46708)

Reason: I am approving this docum'ent.

(Z46708)

Dale: 2006.08.30 15:24:06 -07W0 Effective Date: September 15, 2006

TABLE OF CONTENTS TITLE PAGE

1.0 INTRODUCTION

1 1.1 Liquid Effluent Pathways 1

1.2 Gaseous Effluent Pathways 2

1.3 Nuisance Pathways 2

1.4 Meteorology 4

2.0 GASEOUS EFFLUENT MONITOR SETPOINTS 5

2.1 Requirements

Gaseous Monitors 5

2.1.1 Surveillance Requirements 5

2.1.2 Implementation of the Requirements 12 2.1.2.1 Equivalent Dose Factor Determination 13 2.1.2.2 Site Release Rate Limit (QspT) 14 2.1.2.3 Unit Release Rate Limits (QUNrr) 15 2.1.2.4 Setpoint Determination 15 2.1.2.5 Monitor Calibration 16 3.0 GASEOUS AND LIQUID EFFLUENT DOSE RATES 17

3.1 Requirements

Gaseous Effluents 17 3.1.1 Surveillance Requirements 17 3.1.2 Implementation of the Requirements 18

3.2 Requirements

Secondary System Liquid Waste Discharges To Onsite Evaporation Ponds or Circulating Water System - Concentration 26 3.2.1 Surveillance Requirements 26 3.2.2 Implementation of the Requirements 26 4.0 GASEOUS & LIQUID EFFLUENTS - DOSE 31

4.1 Requirements

Noble Gases 31 4.1.1 Surveillance Requirements 31 4.1.2 Implementation of the Requirement: Noble Gas 32

4.2 Requirement

Iodine-131, Iodine-133, Tritium, and All Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 33 4.2.1 Surveillance Requirements 33 4.2.2 Implementation of the Requirement 34

4.3 Requirements

Gaseous Radwaste Treatment 36 4.3.1 Surveillance Requirements 36 4.3.2 Implementation of the Requirement 37

4.4 Requirements

Liquid Effluents 57 4.4.1 Surveillance Requirements 57 4.4.2 Implementation of the Requirements 57 i

ODCM Rev. 21

TABLE OF CONTENTS TITLE 5.0 TOTAL DOSE AND DOSE TO PUBLIC ONSITE

5.1 Requirement

Total Dose 5.1.1 Surveillance Requirements 5.1.2 Implementation of the Requirement 6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

6.1 Requirement

REMP 6.1.1 Surveillance Requirements 6.1.2 Implementation of the Requirements

6.2 Requirement

Land Use Census 6.2.1 Surveillance Requirements 6.2.2 Implementation of the Requirements

6.3 Requirement

Interlaboratory Comparison Program 6.3.1 Surveillance Requirements 6.3.2 Implementation of the Requirements 7.0 RADIOLOGICAL REPORTS

7.1 Requirement

Annual Radioactive Effluent Release Report

7.2 Requirement

Annual Radiological Environmental Operating Report PAGE 58 58 58 58 62 62 63 63 71 71 71 72 72 72 83 83 85 86 APPENDIX A APPENDIX B APPENDIX C APPENDIX D DETERMINATION OF CONTROLLING LOCATION BASES FOR REQUIREMENTS 87 2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 87 3.1 GASEOUS EFFLUENT - DOSE RATE 87 3.2 SECONDARY SYSTEM LIQUID WASTE DISCHARGE TO ONSITE EVAPORATION PONDS - CONCENTRATION 88 4.1 GASEOUS EFFLUENT - DOSE, Noble Gases 88 4.2 GASEOUS EFFLUENT - DOSE - Iodine-131, Iodine-133, Tritium, and All Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days89 4.3 GASEOUS RADWASTE TREATMENT 89 4.4 SECONDARY SYSTEM LIQUID WASTE DISCHARGE TO ONSITE EVAPORATION PONDS - DOSE 90 5.1 TOTAL DOSE AND DOSE TO PUBLIC ONSITE 90 6.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) 91 6.2 LAND USE CENSUS 91 6.3 INTERLABORATORY COMPARISON PROGRAM 91 DEFINITIONS REFERENCES 92 96 ii ODCM Rev. 21

LIST OF TABLES TABLE TITLE PAGE 1-1 NUISANCE PATHWAYS 3

2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 6

2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 10 3-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 20 3-2 DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE SITE BOUNDARY 23 3-3 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS 24 3-4 P. VALUES FOR THE INHALATION PATHWAY 25 3-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 27 3-6 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 30 3-7 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 30 4-1 Ri DOSE CONVERSION FACTORS FOR THE GROUND PLANE PATHWAY 39 4-2 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - ADULT RECEPTOR 40 4-3 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - TEEN RECEPTOR 41 4-4 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - CHILD RECEPTOR 42 4-5 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - ADULT RECEPTOR 43 4-6 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY -TEEN RECEPTOR 44 4-7 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - CHILD RECEPTOR 45 4-8 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - ADULT RECEPTOR 46 4-9 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - TEEN RECEPTOR 47 4-10 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - CHILD RECEPTOR 48 iii ODCM Rev. 21

LIST OF TABLES TABLE TITLE PAGE 4-11 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - INFANT RECEPTOR 49 4-12 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - ADULT RECEPTOR 50 4-13 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - TEEN RECEPTOR 51 4-14 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - CHILD RECEPTOR 52 4-15 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - INFANT RECEPTOR 53 4-16 PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 1 54 4-17 PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 2 55 4-18 PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 3 56 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 64 6-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES 68 6-3 DETECTION CAPABILITIES FOR ENVIRONMENTAL ANALYSIS 69 6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS 73 C-1 FREQUENCY NOTATION 95 iv ODCM Rev. 21

LIST OF FIGURES FIGURE TITLE PAGE 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLE SITES 0-10 MILES 77 6-2 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLE SITES 0-35 MILES 78 6-3 DELETED 79 6-4 SITE EXCLUSION AREA BOUNDARY 80 6-5 GASEOUS EFFLUENTS RELEASE POINTS 81 6-6 LOW POPULATION ZONE 0-5 MILES 82 V

ODCM Rev. 21

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) implements the program elements which are required by the Administrative Controls section of the Technical Specifications. The ODCM contains the operational requirements, the surveillance requirements, and actions required if the operational requirements are not met for the Radioactive Effluent Controls Program and the Radiological Environmental Monitoring Program to assure compliance with 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50. The Technical Specifications, Section 3.0, also apply to the ODCM. Substitute the word "Requirements" for "Limiting Condition for Operation." It should be noted that the hot and cold shutdown and operability requirements in Technical Specification 3.0.3 and 3.0.4 do not apply to any of the requirements contained in this ODCM. The ODCM also contains descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report required by the Technical Specifications.

The ODCM provides the parameters and methodology to be used in calculating offsite doses resulting from radioactive effluents, in the calculation of gaseous effluent monitor Alarm/Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. Included are methods for determining air, whole body, and organ dose at the controlling location due to plant effluents to assure compliance with the regulatory requirements detailed in the ODCM. Methods are included for performing dose projections to assure compliance with the gaseous treatment system operability sections of the ODCM. The ODCM utilizes information from NRC Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," October 1977, and NRC NUREG 0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,"

October 1978. NUREG 0133 utilizes some of the key information in Regulatory Guide 1.109 to provide methods which were used in the preparation of the radiological effluent Technical Specifications and which have now been transferred to the ODCM in accordance with NRC Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program," January 31, 1989, and NUREG 1301, "Offsite Dose Calculation Manual Guidance:

Standard Radiological Effluent Controls for Pressurized Water Reactors," Generic Letter 89-01, Supplement No. 1, April 1991. Further guidance for the implementation of the new 10 CFR Part 20, effective January 1, 1994, was obtained from the Federal Register, Vol. 58, December 23, 1993. It is recognized that this is only draft guidance, however, it is the only guidance for referencing the new 10 CFR 20 in the ODCM.

1.1 Liquid Effluent Pathways Dose calculation methodology for radioactive liquid effluents is not included in this manual due to the desert location of the plant, the hydrology of the area, and the fact that there are no liquid releases to areas at or beyond the SITE BOUNDARY during normal operation. All liquid discharges to the onsite evaporation ponds are controlled by Section 3.2. The impact of postulated accidental seepages on the groundwater system, and in particular on the existing wells located in the 5-mile zone around the site area has been calculated and analyzed in Section 2.4.13.3 of the PVNGS FSAR.

If plant operating conditions become such that the likelihood of a liquid effluent pathway is created, then dose calculation methodology for this pathway will be added to this manual.

I ODCM Rev. 21

1.2 Gaseous Effluent Pathways All gaseous effluents are treated as ground level releases and are considered to be "long-term" as discussed in NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants." This includes the containment purge and Waste Gas Decay Tank releases as well as the normal ventilation system and condenser vacuum exhaust releases. All releases are either greater than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in duration or are made at random, not depending upon atmospheric conditions or time of day. The releases are lumped together and calculated as an entity. Historical annual average X/Q values are used throughout this manual for all gaseous effluent setpoint and dose calculations. Airborne releases are further subdivided into two subclasses:

1.2.1 Iodine-131, Iodine-133, Tritium and Radionuclides In Particulate Form with Half-lives Greater than Eight Days In this model, a controlling location is identified for assessing the maximum exposure to a MEMBER OF THE PUBLIC for the various pathways and to critical organs. Infant exposure occurs through inhalation and any actual milk pathway. Child, teenager and adult exposure derives from inhalation, consumed vegetation pathways, and any actual milk and meat pathways. Dose to each of the seven organs listed in Regulatory Guide 1.109 (bone, liver, total body, thyroid, kidney, lung and GI-LLI) are computed from individual nuclide contributions in each sector. The largest of the organ doses in any sector is compared to 10 CFR 50, Appendix I design objectives. The release rates of these nuclides will be converted to instantaneous dose rates for comparison to the limits of 10 CFR 20.

1.2.2 Noble Gases The air dose from both the beta and gamma radiation component of the noble gases will be assessed and compared to the 10 CFR 50, Appendix I design objectives. The noble gas release rate will be converted to instantaneous dose rates for comparison to the limits of 10 CFR 20.

Section 2.0 of this manual discusses the methodology to be used in determining effluent monitor alarm/trip setpoints to assure compliance with the 10 CFR Part 20 limits as implemented in Section 3.0. Section 4.0 discusses the methods to assure releases are As Low As Reasonably Achievable (ALARA) in accordance with Appendix I to 10 CFR Part 50.

Methods are described in Section 5.0 for determining the annual cumulative dose to a MEMBER OF THE PUBLIC from gaseous effluents and direct radiation to assure compliance with 40 CFR Part 190.

The requirements for the Annual Radiological Effluent Release Report and the Radiological Environmental Monitoring Program, including the Annual Land Use Census and the Interlaboratory Comparison Program, and the Annual Environmental Report are described in Sections 6.0 and 7.0 of this manual.

1.3 Nuisance Pathways This section addresses the potential release pathways which should not contribute more than 10% of the doses evaluated in this manual. Table 1-1 lists examples of potential release pathways. The ODCM methodology for calculation of doses will be applied to an applicable release pathway if a likely potential arises for contributing more than 10% of the doses evaluated in this manual.

2 ODCM Rev. 21

TABLE 1-1 NUISANCE PATHWAYS (EXAMPLES)

Evaporation Pond Cooling Towers Laundry/Decon Building Exhaust Unmonitored Secondary System Steam Vents/Reliefs Turbine Building Ventilation Exhaust Unmonitored Tank Atmospheric Vents Dry Active Waste Processing and Storage (DAWPS) Building Respirator Cleaning Facility Secondary Side Decontamination Equipment Low Level Radioactive Material Storage Facility 3

ODCM Rev. 21

1.4 Meteorology Historical annual average atmospheric dispersion (X/Q) and deposition (D/Q) data, based on nine years of meteorological data, and given in Table 3-2 for each of the three nuclear generating units are used to demonstrate compliance with the ODCM Requirements. These Requirements include:

Section 2.0 Gaseous Effluent Monitor Setpoints; Section 3.0 Gaseous and Liquid Effluent - Dose Rate Section 4.0 Gaseous and Liquid Effluent - Dose Section 5.0 Total Dose and Dose to Public Onsite Sections 2.0 and 3.0 specify utilizing the highest X/Q or D/Q meteorological dispersion parameter at the Site Boundary for any of the three units as applicable. Using the highest dispersion parameter for any of the units provides a conservative assumption to assure compliance with the higher 10 CFR Part 20 limits.

Section 4.0 specifies utilizing the highest X/Q at the Site Boundary for the particular unit, from Table 3-2 for noble gases. The highest X/Q and D/Q are utilized for the particular unit's releases as applicable for gases other than noble gases (iodines, particulates, and tritium) for the controlling pathway's location (site boundary using Table 3-2 or other controlling locations using Table 4-16, 4-17, or 4-18).

Section 5.0 specifies utilizing the highest X/Q for the particular unit's releases at the controlling location from Table 4-16, 4-17, or 4-18, for noble gases. The highest X/Q and DIQ are utilized for the particular unit's releases as applicable for gases other than noble gases at the controlling pathway's location using Table 4-16, 4-17, or 4-18.

Section 7.0 requires that 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.

4 ODCM Rev. 21

2.0 GASEOUS EFFLUENT MONITOR SETPOINTS

2.1 Requirements

Gaseous Monitors The radioactive gaseous effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the dose requirements in Section 3.0 are not exceeded. The alarm/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Section 2.1.2.

Applicability:

As shown in Table 2-1.

Action:

a. With the low range radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Requirement, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-1. Restore the inoperable instrumentation to OPERABLE status within 30 days or, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

2.1.1 Surveillance Requirements

a. Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2.

5 ODCM Rev. 21

TABLE 2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT

1.

GASEOUS RADWASTE SYSTEM

a.

Noble Gas Activity Monitor -Providing Alarm and Automatic Termination of Release #RU-12

b.

Flow Rate Monitor

2.

NOT USED

3.

DELETED

4.

PLANT VENT SYSTEM A.

Low Range Monitors

a.

Noble Gas Activity Monitor #RU-143

b.

Iodine Sampler

c.

Particulate Sampler

d.

Flow Rate Monitor

e.

Sampler Flow Rate Measuring Device B.

High Range Monitors

a.

Noble Gas Activity Monitor #RU-144

b.

Iodine Sampler

c.

Particulate Sampler

d.

Sampler Flow Rate Measuring Device Ila OPERABLE APPLICABILITY ACTION 35 36 37 40 40 36 36 42 42 42 42

TABLE 2-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT

5.

FUEL BUILDING VENTILATION SYSTEM A.

Low Range Monitors

a.

Noble Gas Activity Monitor #RU-145

b.

Iodine Sampler

c.

Particulate Sample

d.

Flow Rate Monitor

e.

Sampler Flow Rate Measuring Device B.

High Range Monitors

a.

Noble Gas Activity Monitor #RU-146

b.

Iodine Sampler

c.

Particulate Sample

d.

Sampler Flow Rate Measuring Device OPERABLE APPLICABILITY ACTION I

I 1

1 1

I 1

37,41 40 40 36 36

-j 42 42 42 42 0

.0

Table 2-1 (Continued)

TABLE NOTATION At all times.

- During GASEOUS RADWASTE SYSTEM operation

- Whenever the condenser air removal system is in operation, or whenever turbine glands are being supplied with steam from sources other than the auxiliary boiler(s).

During waste gas release.

0#

In MODES 1, 2, 3, and 4 or when irradiated fuel is in the fuel storage pool.

ACTION 35 -

ACTION 36 -

ACTION 37 -

ACTION 38 -

ACTION 39 -

ACTION 40 -

ACTION 41 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a.

At least two independent samples of the tanks contents are analyzed, and

b.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

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

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the actions of (a) or (b) or (c) are performed:

a.

Initiate the Preplanned Alternate Sampling Program to monitor the appropriate parameter(s).

b.

Place moveable air monitors in-line.

c.

Either take grab samples at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , OR obtain gas channel monitor readings locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the channel is functional locally but inoperable due to loss of communication with the minicomputer. The surveillance requirements of Section 2. 1.1 must be performed at the required frequencies for the channel to be functional locally.

NOT USED NOT USED With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the effected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 3-1 within one hour after the channel has been declared inoperable.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, comply with Technical Requirements Manual TLCO 3.3.108.

8 8 ODCM Rev. 21

Table 2-1 (Continued)

TABLE NOTATION ACTION 42-With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement restore the channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or:

a.

Initiate the Preplanned Alternate Sampling Program to monitor the appropriate parameter(s) when it is needed.

b.

Prepare and submit a Special Report to the Commission within 30 days following the event outlining the action(s) taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

9 ODCM Rev. 21

TABLE 2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0

INSTRUMENT

1.

GASEOUS RADWASTE SYSTEM

a.

Noble Gas Activity Monitor -Providing Alarm and Automatic Termination of Release RU-12

b.

Flow Rate Monitor

2.

DELETED

3.

DELETED

4.

PLANT VENT SYSTEM (RU-143 and RU-144)

a.

Noble Gas Activity Monitor

b.

Iodine Sampler

c.

Particulate Sampler

d.

Flow Rate Monitor

e.

Sampler Flow Rate Measuring Device

5.

FUEL BUILDING VENTILATION SYSTEM (RU-145 and RU-146)

a.

Noble Gas Activity Monitor

b.

Iodine Sampler

c.

Particulate Sample

d.

Flow Rate Monitor

e.

Sampler Flow Rate Measuring Device CHANNEL CHECK P

P D(5)

N.A.

D(6)

D(5)

N.A.

D(6)

SOURCE CHECK P(7)

N.A.

M(7)

N.A.

M(7)

N.A.

CHANNEL CALIBRATION R(3)

R R(3)

N.A.

R R

R(3)

N.A.

R R

CHANNEL MODE IN WHICH FUNCTIONAL SURVEILLANCE TEST IS REQUIRED Q(1),(2),P###

Q,P###

Q(2)

N.A.

Q Q

0 03 Q(2)

N.A.

Q Q

Table 2-2 (Continued)

TABLE NOTATION At all times.

During GASEOUS RADWASTE SYSTEM operation Whenever the condenser air removal system is in operation, or whenever turbine glands are being supplied with steam from sources other than the auxiliary boiler(s).

During waste gas release.

In MODES 1, 2, 3, and 4 or when irradiated fuel is in the fuel storage pool.

Functional test should consist of, but not be limited to, a verification of system isolation capability by the insertion of a simulated alarm condition.

(1)

The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if the instrument indicates measured levels above the alarm/trip setpoint.

(2)

The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Circuit failure.

3.

Instrument indicates a downscale failure.

4.

Instrument controls not set in operate mode.

(3)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards 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 measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used in lieu of the reference standards associated with the initial calibration.

(4)

NOT USED (5)

The channel check for channels in standby status shall consist of verification that the channel is on-line and reachable.

(6)

Daily channel check not required for flow monitors in standby status.

(7)

LED may be utilized as the check source in lieu of a source of increased activity.

11 ODCM Rev. 21

2.1.2 Implementation of the Requirements The general methodology for establishing low range gaseous effluent monitor setpoints is based upon a site release rate limit in ltCi/sec derived from site specific meteorological dispersion conditions, radioisotopic distribution, and whole body and skin dose factors. The high alarm of the low range monitors will alarm/trip when the release rate from an individual vent will result in exceeding the limits in Section 3.1. 80% of Section 3.1 limits is considered to be the site release rate limit. The site release rate limit will be allocated among the licensed units' release points. The unit release rate limit will then be utilized for the determination of gaseous effluent monitor setpoints. A fraction of the unit release rate limit is then allotted to each release point and its monitor alert setpoint (gCi/cc) is derived using actual or fan design flow rates.

Administrative values are used to reduce each setpoint to account for the potential activity in other releases. These administrative values shall be reviewed based on actual release data.

For the purpose of implementation of Section 2.1, the alarm setpoint levels for low range effluent noble gas monitors are established to ensure that personnel are alerted when the noble gas releases are at a rate such that if the releases would continue for the year they would approach the total body dose rate of 500 mrem/yr and 3000 mrem/yr skin dose in Section 3.1.

The equations in Section 3.1 of this manual provide the methodology for calculating the gaseous effluent dose rate.

The evaluation of doses due to releases of radioactive material can be simplified by the use of equivalent dose factors as defined in Section 2.1.2.1.

The equivalent dose factors will be evaluated periodically to assure that the best information on isotopic distribution is being used for the dose equivalent value.

12 ODCM Rev. 21

2.1.2.1 Equivalent Dose Factor Determination The equivalent whole body dose factor is calculated as follows:

Keq = ZI[(Ki)(fi)]

(2-1)

Where:

Keq

= the equivalent whole body dose factor weighted by historical radionuclide distribution in releases in mrem/yr per 54Ci/m 3.

Ki1

= the whole body dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrem/yr per iVCi/m 3 from Table 3-3.

fi

= the fraction of noble gas radionuclide i in the total noble gas radionuclide mix.

The equivalent skin dose factor is calculated as follows:

(L + 1.1M)eq = Fi[(Li + 1.1Mi)(fi)]

(2-2)

Where:

(L+l.lM)cq = the equivalent skin dose factor due to beta and gamma emissions from all noble gases released, weighted by the historical radionuclide distribution in releases in mrem/yr per gLCi/m 3.

= the skin dose factor due to the beta emissions for each identified noble gas radionuclide i, in mrem/yr per g+/-Ci/m 3 from Table 3-3.

Mi

= the air dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrad/yr per t+/-Ci/m3 from Table 3-3.

fi

= the fraction of noble gas radionuclide i in the total noble gas radionuclide mix.

1.1

= unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

13 ODCM Rev. 21

2.1.2.2 Site Release Rate Limit (QsITE)

The release rates corresponding to 80% of the whole body (QwB) and skin (QsK) dose rate limits are calculated using the equivalent dose factors defined in Section 2.1.2.1.

The site release rate limit (QsITE) is the lower of QWB or QSK, thus assuring that the more restrictive dose rate limit will not be exceeded.

The QSrTE is established as follows:

(DWB)(0. 8 )

(2-3)

QSITE,WB -(Keq)(X/Q)sITE Where:

QSnn-EWB the site release rate, in IiCi/sec, that would deliver a dose rate 80% of the whole body dose rate limit, DWB.

DWB

- whole body dose rate limit of 500 mrem/yr.

Keq=

equivalent whole body dose factor, in mrem/yr per jtCi/m3 weighted by the historical radionuclide distribution.

(X/Q)SlTE

= 8.91E-06, the highest calculated annual average dispersion parameter, in sec/m3, at the Site Boundary for any of the 3 units, from Table 3-2.

0.8

= administrative factor to compensate for any unexpected variability in the radionuclide mix and to ensure that Site Boundary dose rate limits will not be exceeded.

(DSK)(O. 8 )

QSITESK = (L + 1.1M)eq(X/Q)sITE (2-4)

Where:

QSITE,SK

= the site release rate limit, in ItCi/sec, that would deliver a dose rate 80% of the skin dose rate limit, DSK.

DSK

= skin dose rate limit of 3000 mremlyr.

(L+1.1M)eq = equivalent skin dose factor, in mrem/yr per IJCi/m 3, weighted by the radionuclide distribution.

(X/Q)sITE

= 8.91E-06, the highest calculated annual average dispersion parameter, in sec/M 3, at the Site Boundary for any of the three units, from Table 3-2.

0.8

= administrative factor to compensate for any unexpected variability in the radionuclide mix and to ensure that Site Boundary dose rate limits will not be exceeded.

After determination of the QSITE whole body and skin dose rates (equations 2-3 and 2-4, respectively), the most conservative result will be used as QsrrE, the site release rate limit.

14 ODCM Rev. 21

2.1.2.3 Unit Release Rate Limits (QuNrr)

Typically QSITE will be divided equally among operating units. If operational history dictates a larger fraction of the QSITE be assigned to a specific unit then a weighted average of each unit's contribution to the QSpTE will be utilized to determine the QuNrr.

QuNrr

= (fuNrr) (QsrrE)

(2-5)

Where:

QuNrr

= unit release rate limit, in pCi/sec.

fuNrr

= the fraction (_< 1) of noble gas historically released from a specific operating unit to the total of all noble gas released from the site.

QsrrE

= the site release rate limit, in jtCi/sec determined in Section 2.1.2.2.

2.1.2A Setpoint Determination To comply with the requirements in Section 2.1, the alarm/trip setpoints can now be established using the unit release rate limit (QuNIT) to ensure that the noble gas releases do not exceed the dose rate limits.

To allow for multiple sources of releases from different or common release points, the effluent monitor setpoint includes an administrative factor which allocates a percentage of the unit release rate limit to each of the release sources. Monitor setpoints will also be adjusted in accordance with Nuclear Administrative and Technical Manual procedures to account for monitor-specific characteristics.

Monitors RU-143 and RU-145 The alarm/trip setpoint for Monitors RU-143 and RU-145 is calculated as follows:

Monitor (QUNIT)(a)

Setpoint (472)(Flow Rate)

(2-6)

Where:

Monitor Setpoint

= the setpoint for the effluent monitor, in g+/-Ci/cc, which provides a safe margin of assurance that the allowable dose rate limits will not be exceeded.

QUNrT

= unit release rate limit, in ItCi/sec, as determined in Section 2.1.2.3.

Flow Rate

= the flow rate, in cfm, from flow rate monitors or the fan design flow rate for the release source under consideration.

472

= conversion factor, cubic centimeter/second per cubic feet/minute.

a

= fraction of QUNrT allocated for a specific release point. The sum of these administrative values shall be less than or equal to one.

15 ODCM Rev. 21

Monitor RU-2 The alarm/trip setpoint for Monitor RU-12, the Waste Gas Decay Tank Monitor, is calculated as follows:

Monitor

[(QUNIT)(a)(0. 9 ) - (H)(PF)(472)]

(2-7) setpoint (Flow Rate)(472)

Where:

Monitor Setpoint

= the setpoint for the monitor, in gtCi/cc at STP, which provides a safe margin of assurance that the allowable dose rate limits will not be exceeded.

QUNIT

= unit release rate limit, in jiCi/sec, as determined in Section 2.1.2.3.

Flow Rate

= flow rate, in cfm at STP at which the tank will be released.

PF

= the current process flow of the plant vent in CFM.

H

= the current plant vent monitor concentration in *tCi/cc.

a

= fraction of QUNIj allocated for a specific release point. This administrative value should be equal to or less than the administrative value used for the Plant Vent.

0.9

= an administrative value to account for potential increases in activity from other contributors to the same release point.

472

= conversion factor, cubic centimeter/second per cubic feet/minute.

If there is no release associated with this monitor, the monitor setpoint should be established as close as practical to background to prevent spurious alarms, and yet assure an alarm should an inadvertent release occur.

2.1.2.5 Monitor Calibration The Radiation Level Conversion Factor (RLF) for each monitor is entered into the Radiation Monitoring System Database and may change whenever the monitor is calibrated. Calibration is performed in accordance with Nuclear Administrative and Technical Manual procedures.

16 ODCM Rev. 21

3.0 GASEOUS AND LIQUID EFFLUENT DOSE RATES

3.1 Requirements

Gaseous Effluents The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 6-4 and Figure 6-5) shall be limited to the following:

a.

For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and

b.

For 1-131 and 1-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

Applicability:

At all times.

Action:

With the dose rate(s) exceeding the above limits, immediately decrease the release rate to within the above limits(s).

3.1.1 Surveillance Requirements

a. The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods contained in Section 3.1.2.

b. The dose rate due to 1-131, 1-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods contained in Section 3.1.2 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3-1.

17 ODCM Rev. 21

3.1.2 Implementation of the Requirements Noble gas activity monitor setpoints are established at release rates which permit corrective action to be taken before exceeding the 10 CFR 20 annual dose limits as described in Section 2.0. The requirements for sampling and analysis of continuous and batch effluent releases are given in Table 3-1. The methods for sampling and analysis of continuous and batch effluent releases are given in the Nuclear Administrative and Technical Manual procedures. The dose rate in unrestricted areas shall be determined using the following equations.

For whole body dose rate:

DWB = Ei[(Ki)(X/Q)S1TE(Qi)]

(3-1)

For skin dose rate:

DSK = X"1[(L1 + l.lMi)(X/Q)SITE(Qi)]

(3-2)

Where:

Ki

= the whole body dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrem/yr per VCi/m 3 from Table 3-3.

Qi

= the release rate of radionuclide i, in pCi/sec.

(X/Q)srrE

= 8.91E-06, the highest calculated annual average dispersion parameter, in sec/m 3, for any of the three units, from Table 3-2.

DWB

= the annual whole body dose rate (mrem/yr.).

Li

= the skin dose factor due to the beta emissions for each identified noble gas radionuclide i, in mrem/yr per gCi/m 3 from Table 3-3.

Mi

= the air dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrad/yr per gCi/m3 from Table 3-3.

DSK

= the annual skin dose rate (mrem/yr).

1.1

= unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

18 ODCM Rev. 21

1-131.1-133. tritium and radionuclides in particulate form with half-lives greater than 8 days The methods for sampling and analysis of continuous and batch releases for 1-131, 1-133, tritium and radionuclides in particulate form with half-lives greater than 8 days, are given in the applicable Nuclear Administrative and Technical Manual procedures. Additional monthly and quarterly analyses shall be performed in accordance with Table 3-1. The total organ dose rate in unrestricted areas shall be determined by the following equation:

Do = -I[(Pi)(X/Q)SITE(Qi)I (3-3)

Where:

Pi

= the dose factor, in mrem/yr per gxCi/m 3, for radionuclide i, for the inhalation pathway, from Table 3-4.

(X/Q)srm

= 8.911E-06, the highest calculated annual average dispersion parameter, in sec/m 3, at the Site Boundary, for any of the three units, Qi

= the release rate of radionuclide i, in VCi/sec D.

= the total organ dose rate (mrem/yr).

19 ODCM Rev. 21

TABLE 3-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM LOWER LIMIT SAMPLING ANALYSIS TYPE OF OF DETECTION GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (LLD) (tCi/ml)a A. Waste Gas Storage P

P Principal Gamma L.OE-04 Each Tank Grab Each Tank Emittersg Sample B. Containment Purge P

P Principal Gamma 1.01E-04 Each Purgeb'c Each Purgebc Emitters5 Grab Sample H-3 1OE-06 C. 1. DELETED Mb'e Mb Principal Gamma L.OE-04

2. Plant Vent Grab Sample Emitters9

3. Fuel Bldg. Exhaust H-3 1.01E-06 Continuousf 4/Md 1-131 1.01E-12 Charcoal Sample 1-133 1.OE-10 Continuousf 4/Md Principal Gamma 1.OE-1 1 Particulate Emittersg Sample (1-131, Others)

Continuousf M

Gross Alpha 1.OE-1 1 Composite Particulate Sample Continuousf Q

Sr-89, Sr-90 1.OE-I 1 Composite Particulate Sample D. All Radwaste Types as listed In A., B., and C.,

above.

Noble Gas Noble Gases Gross Beta Monitor or Gamma 1.OE-06 20 ODCM Rev. 21

Table 3-1 (Continued)

TABLE NOTATION a

The LLD is the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% 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 E

V

2.22E6

Y

exp(-XAt)

Where:

LLD is the a priori lower limit of detection as defined above (as gtCi per unit mass or volume). Current literature defines the LLD as the detection capability for the instrumentation only and the MDC minimum detectable concentration, as the detection capability for a given instrument, procedure and type of sample.

sb 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 transformation),

V is the sample size (in units of mass or volume),

2.22E6 is the number of transformations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

X is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between the midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry the background should include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as an jLpdi (before the fact) limit representing the capability of a measurement system and not as an a posteriri (after the fact) limit for a particular measurement.

21 ODCM Rev. 21

Table 3-1 (Continued)

TABLE NOTATION b

Analyses shall also be performed following SHUTDOWN, STARTUP, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period if 1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has increased more than a factor of 3; and 2) the noble gas activity monitor on the plant vent shows that effluent activity has increased by more than a factor of 3. If the associated noble gas vent monitor is inoperable, samples must be obtained as soon as possible. Analyses shall be performed within a four-hour period. This requirement does not apply to the Fuel Building Exhaust.

c Sampling and analyses shall also be performed at least once per 31 days when purging time exceeds 30 days continuous.

d Samples shall be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

e Tritium grab samples shall be taken at least monthly from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

f The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Requirements 3.1, 4.1 and 4.2 of the ODCM.

g The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides shall also be identified and reported in the Annual Radioactive Effluent Release Report.

22 ODCM Rev. 21

TABLE 3-2 DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE SITE BOUNDARY DISTANCE DIRECTION (METERS)

N 1037 NNE 1057 NE 2206 ENE 1967 E

1927 ESE 1967 SE 2049 SSE 2730 S

3006 SSW 2258 SW 1487 WSW 1251 W

1225 WNW 1244 NW 1254 NNW 1069 UNITI XIQ D/Q (SEC/n 3)

(m"2) 4.93E-06 9.24E-09 4.14E-06 1.19E-08 2.84E-06 6.84E-09 2.51E-06 4.43E-09 2.56E-06 3.24E-09 2.61E-06 2.46E-09 3.56E-06 2.36E-09 3.80E-06 1.58E-09 5.07E-06 1.78E-09 6.52E-06 3.20E-09 7.47E-06 5.65E-09 4.52E-06 5.93E-09 4.73E-06 9.49E-09 3.76E-06 6.76E-09 3.43E-06 5.87E-09 3.70E-06 7.26E-09 UNIT 2 DISTANCE (METERS) 1318 1342 2545 2206 2163 2067 2101 3026 2699 1836 1208 1014 993 1010 1191 1342 XIQ D/Q (SEC/m3)

(M7 2 )

3.85E-06 6.17E-09 3.18E-06 7.93E-09 2.42E-06 5.34E-09 2.22E-06 3.64E-09 2.27E-06 2.66E-09 2.32E-06 2.11E-09 3.47E-06 2.26E-09 3.43E-06 1.32E-09 5.16E-06 1.97E-09 7.90E-06 4.56E-09 7.72E-06 6.88E-09 5.55E-06 8.44E-09 5.86E-06 1.34E-08 4.67E-06 9.60E-09 3.62E-06 6.40E-09 2.85E-06 4.87E-09 UNIT 3 DISTANCE (METERS) 1661 1693 2756 2337 2290 2023 2256 2786 2346 1607 1057 889 871 885 1045 1561 XIQ D/Q (SEC/rn3)

(rn 2) 3.54E-06 4.86E-09 2.86E-06 6.23E-09 2.21E-06 4.65E-09 2.08E-06 3.30E-09 2.14E-06 2.41E-09 2.37E-06 2.10E-09 3.24E-06 2.OOE-09 3.72E-06 1.52E-09 5.90E-06 2.51E-09 8.91E-06 5.73E-09 8.68E-06 8.61E-09 5.34E-06 8.83E-09 6.72E-06 1.67E-08 5.37E-06 1.19E-08 4.17E-06 7.98E-09 2.93E-06 4.58E-09 tO3 0

Reference:

Distances are from the PVNGS ER-OL, Table 2.3-33. Dispersion and Deposition parameters are from a September, 1985, calculation by NUS Corporation based on 9 years of meteorological data; NUS Corporation letter NUS-ANPP-1386, dated October 4, 1985.

TABLE 3-3 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS Whole Body Dose Factor Y4 Radionuclide Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Ar-41 mrem-m3 yr-RCi 7.56E-02 1.17E+03 1.61E+O1 5.92E1+03 1A7E+04 1.66E+04 1.56E+04 9.15E+01 2.511E+02 2.94E+02 3.12E+03 1.81E+03 1.42E3+03 8.83E+03 8.84E+03 Skin Gamma Air Dose Factor Dose Factor Li Mi mrem-m mrad-m3 yr-ACi yr-tCi 1.93E+01 1.46E+03 1.23E3+03 1.34E3+03 1.72E+01 9.73E1+03 6.17E+03 2.37E+03 1.52E+04 1.01E+04 1.73E+04 7.29E+03 1.63E+04 4.76E+02 1.56E+02 9.94E+02 3.27E+02 3.06E3+02 3.53E+02 7.11E+02 3.36E+03 1.86E+03 1.92E+03 1.22E+04 1.51E+03 4.13E+03 9.21E+03 2.69E+03 9.30E+03 Beta Air Dose Factor Ni mrad-m3 yr-RCi 2.88E+02 1.97E+03 1.95E+03 1.03E+04 2.93E+03 1.06E+04 7.83E+03 1.11E+03 1.48E+03 1.05E+03 7.39E+02 2.46E+03 1.27E+04 4.75E+03 3.28E+03

Reference:

Regulatory Guide 1.109, Table B-1.

24 ODCM Rev. 21

TABLE 3-4 P1 VALUES FOR THE INHALATION PATHWAY (mrem/yr/tCi/m3)

NUCLIDE Age Group Organ P1 H-3 TEEN LIVER 1.27E+03 CR-51 TEEN LUNG 2.10E+04 MN-54 TEEN LUNG 1.98E+06 FE-59 TEEN LUNG 1.53E+06 CO-58 TEEN LUNG 1.34E+06 CO-60 TEEN LUNG 8.72E+06 ZN-65 TEEN LUNG 1.24E+06 SR-89 TEEN LUNG 2.42E+06 SR-90 TEEN BONE 1.08E+08 ZR-95 TEEN LUNG 2.69E+06 SB-124 TEEN LUNG 3.85E+06 1-131 CHILD THYROID 1.62E+07 1-133 CHILD THYROID 3.85E+06 CS-134 TEEN LIVER 1.13E+06 CS-137 CHILD BONE 9.07E+05 BA-140 TEEN LUNG 2.03E+06 CE-141 TEEN LUNG 6.14E+05 CE-144 TEEN LUNG 1.34E+07 25 ODCM Rev. 21

3.2 Requirements

Secondary System Liquid Waste Discharges To Onsite Evaporation Ponds or Circulating Water System - Concentration The concentration of radioactive material discharged from secondary system liquid waste to the circulating water system shall be limited to:

5.0E-07 pCi/ml for the principal gamma emitters (except Ce-144) 3.0E-06 VCi/mI for Ce-144 1.0E-06 gCi/ml for 1-131 1.0E-03 pCi/ml for H-3 The concentration of radioactive material discharged from secondary system liquid waste to the onsite evaporation ponds shall be limited to:

2.0E-06 p.Ci/ml for Cs-134 2.0E-06 ItCi/ml for Cs-137 The concentrations specified in 10 CFR Part 20.1001-20.2402, Appendix B, Table 2, Column 2, for all other isotopes Applicability:

At all times.

Action:

When any secondary system liquid waste discharge pathway concentration determined in accordance with the surveillance requirements given below exceeds the above Requirements, divert that discharge pathway to the liquid radwaste system without delay or terminate the discharge.

3.2.1 Surveillance Requirements

a. Secondary system liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 3-5.

3.2.2 Implementation of the Requirements This requirement is implemented by Nuclear Administrative and Technical Manual procedures.

26 ODCM Rev. 21

F -

Y -

I I -

Y -

II, II, 0

9

'I,9 9

rJJ9 9I SI, 9

lo C-9 9

U, 0

9 8

1 SI, O*99 II,.-.~~

C--

0 0

Il, 0

9 0

9 4

4 4

4 4-4-

Sn 0C 0

00 0 I 0

U

-C~) ~

4) 0I I.

4) 0 0

C.)

0 ~

~xJ CC 00 0

4)

I u -

u -

q -

y -

4 -

~ -

I -

P -

S

~

C 0

'~.0

.0 E~~<

.0<

0

~0.

0.

4)

4) 4)

CO CO

.0

.0 0

~

C..)

.0

.0 0.

0

.0 e

.~

.0

.0<

~

r Z

r~

Z

~

U

0.

~

0a CI 0C UI

.0 0

0 0

4) 4)

0 0

,0 ma 0.

5-4) 0 0

0ý 00.0

~CO 00

.0CO 0

t u~

.0CO 0

L)L)

OS 0

Ci) 0.0

~CO 00 0

Cd, 0

.0 00 0

2 A

4) 0 US U) 5-4)z 4)

-a £ 30 E0 A*

UP di 4) 0 C.,

U 4)

V00U

.0 C.)

0 0

-a0 4)0 4)Cd, 0

4)0'I 4

4) 0U,04)

V00 U

V I -

h a -

a -

a m 27 ODCM Rev. 21

TABLE 3-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Of Detection Secondary System Liquid Release Sampling &

Type Of (LLD)a Pathway Destination Analysis Frequency Notes Activity Analysis (jtCi/ml)

5.

North & South Condenser Area retention basin D Grab Sample 3

Principal Gamma Emittersc 5.OE-07 Sumpsd CWNT N.A.

1-131 1.OE-06 H-3 1.OE-05

6.

Steam Generator Blowdown to retention basin P Each Discharge 2 Principal Gamma Emittersc 5.OE-07 Retention Basin through SC-N-V064 1-131 1.OE-06 H-3 1.01E-05

7.

Retention Basin to Evaporation evaporation pond P

Each Batch Principal Gamma Emitters-5.OE-07 Pond 1-131 1.OE-06 H-3 1.OE-05 I

Sampling and analysis are required only when concentration for chemical waste neutralizer tank or steam generator activity exceeds the requirement 2

RU-200 shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 3-6. The Alarm/Trip setpoints for RU-200 are set to ensure that the concentrations in the retention basins do not exceed the Requirement 3

Sampling and analysis are required only when concentration for chemical waste neutralizer tank or condensate activity exceeds the requirement 00 0

U

Table 3-5 (Continued)

TABLE NOTATION a

The LLD is defined 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 E

V

2.22E6

Y

exp(-XAt)

Where:

LLD is the "a priori" lower limit of detection as defined above as microcuries per unit mass or volume, sb 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, 2.22E6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield when applicable, X, is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting.

Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as an a pdiri (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

b A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

c The principal gamma emitters for which the LLD specification applies include 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 3.0E-06. 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 Annual Radioactive Effluent Release Report.

d A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release 29 ODCM Rev. 21

TABLE 3-6 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Channel Mode In which Channel Source Channel Functional Surveillance is Instrument Check Check Calibration Test Required RU-200 P

N. A.

R Q

See Table 3-7 TABLE 3-7 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Mode In which Surveillance Secondary System Liquid Release Pathway is Required Action if RU-200 is Inoperable Obtain grab sample at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyze in accordance Pre-service rinse to retention basins At All Times with section 3.2 Obtain grab sample at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyses in accordance Condensate overboard to retention basins 1-4 with section 3.2 Modes 1-4: Suspend the release Steam Generator Blowdown/Drain to retention At All Times Modes 5,6 & defueled: Obtain grab basins sample at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyze in accordance with sec-tion 3.2 30 ODCM Rev. 21

4.0 GASEOUS & LIQUID EFFLUENTS -DOSE

4.1 Requirements

Noble Gases The air dose due to noble gases released in gaseous effluents, from each reactor unit to areas at and beyond the SITE BOUNDARY (see Figure 6-4 and Figure 6-5) shall be limited to the following:

a.

During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation and,

b.

During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

Applicability:

At all times.

Action:

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

4.1.1 Surveillance Requirements

a. Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology contained in Section 4.1.2 at least once per 31 days.

31 ODCM Rev. 21

4.1.2 Implementation of the Requirement: Noble Gas The air dose in unrestricted areas beyond the site boundary due to noble gases released in gaseous effluents from each unit during any specified time period shall be determined by the following equations:

For gamma radiation:

D yu

-= (3.17E-08) Z [(Mi) (X/Q)UNrT(Qi)]

(4-1)

For beta radiation:

D 0,

= (3.17E-08) 7-i [(Ni) (X/Q)UN(Qi)]

(4-2)

Where:

Mi

= the air dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrad/yr per tCi/m3 from Table 3-3.

Ni

= the air dose factor due to beta emissions for each identified noble gas radionuclide i, in mrad/yr per ilCi/m 3 from Table 3-3.

(X/Q)uNrr

= the highest calculated annual average dispersion parameter, in sec/m 3, at the site boundary for the particular Unit, from Table 3-2. Optionally, the highest value may be used for any Unit calculation.

=7.47E-06 from Unit 1

=7.90E-06 from Unit 2

=8.91E-06 from Unit 3 D yu

= the total gamma air dose, for the particular unit, in mrad, due to noble gases released in gaseous effluents for a specified time period at the SITE BOUNDARY.

D fu

= the total beta air dose, for the particular unit, in mrad, due to noble gases released in gaseous effluents for a specified time period at the SITE BOUNDARY.

Qi

= the integrated release, from the particular unit, in tCi, of each identified noble gas radionuclide i, in gaseous effluents for a specified time period.

3.17E-08

= the inverse of seconds in a year (yr/sec).

The cumulative gamma air dose and beta air dose for a quarterly or annual evaluation shall be based on the calculated dose contribution from each specified time period occurring during the reporting time period.

32 ODCM Rev. 21

4.2 Requirement

Iodine-131, Iodine-133, Tritium, and All Radionudides in Particulate Form With Half-Lives Greater Than 8 Days The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 6-4 and Figure 6-

5) shall be limited to the following:

a.

During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,

b.

During any calendar year: Less than or equal to 15 torems to any organ.

Applicability:

At all times.

Action:

With the calculated dose from the release of iodine-13 1, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

4.2.1 Surveillance Requirements

a. Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters contained in Section 4.2.2 at least once per 31 days.

33 ODCM Rev. 21

4.2.2 Implementation of the Requirement The organ dose to an individual from 1-131, 1-133, tritium, and all radionuclides in particulate form, with half-lives greater than eight days, in gaseous effluents released to unrestricted areas from each reactor unit is calculated using the following expressions:

Dou

= (3.17E-08) T [7-k (Rik Wk) (Qi)]

(4-3)

Where:

Do,

= the total accumulated organ dose from gaseous effluents for a particular unit, to a MEMBER OF THE PUBLIC, in torem, at the SITE BOUNDARY or at the controlling location.

Qi

= the quantity of radionuclide i, in tCi, released in gaseous effluents from a particular unit.

Rik

= the dose factor for each identified radionuclide i, for pathway k (for the inhalation pathway in mren/yr per ILCi/m 3 and for the food and ground plane pathways in m2 - mrem/yr per lpCi/sec, except H-3, which has units of mrem/yr per gtCi/m 3) at the controlling location. The Rik's for each age group are given in Tables 4-1 through 4-15.

3.17E-08

= the inverse of seconds per year (yr/sec).

Wk

= the highest annual average dispersion or deposition parameter for the particular Unit, used for estimating the dose at the site boundary or to a MEMBER OF THE PUBLIC at the controlling location for the particular Unit. Optionally, the highest value may be used for any Unit calculation.

= (X/Q)UNrr, in sec/m 3 for the inhalation pathway and for all tritium calculations, for organ dose at the site boundary, from Table 3-2.

=7A7E-06 from Unit 1

=7.90E-06 from Unit 2

=8.91E-06 from Unit 3

= (X/Q)t~rr, in sec/m 3 for the inhalation pathway and for all tritium calculations, for organ dose at the controlling location, from Table 4-16, 4-17 or 4-18.

=2.92E-06 from Unit 1

=2.19E-06 from Unit 2

=2.31E-06 from Unit 3

= (D/Q)t.rr, in m-2, for the food and ground plane pathways, for organ dose at the site boundary, from Table 3-2.

=1.19E-08 from Unit 1

=1.34E-08 from Unit 2

=1.67E-08 from Unit 3 34 ODCM Rev. 21

= (D/Q)uNr, in m-2, for the food and ground plane pathways, for organ dose at the controlling location, from Table 4-16, 4-17, or 4-18.

=3.25E-09 from Unit 1

=3.88E-10 from Unit 2

=4.21E-10 from Unit 3 Residences, vegetable gardens and milk animals located within 5 miles of the site will be identified during the annual land use census. The controlling pathway and location will be identified and will be used for all MEMBER OF THE PUBLIC dose evaluations.

The R1 values were calculated in accordance with the methodologies in NUREG-0133. The following site specific information was used to calculate Ri:

The length of the grazing season for milk animals (fQ).

Ref. ER-OL, Section 2.1.3.4.3 0.75 The length of the grazing season for meat animals (Q.).

Ref. ER-OL, Section 2.1.3.4.4 0.25 The fraction of daily feed derived from pasture while on pasture for milk animals (fp).

Ref. ER-OL, Section 2.1.3A.3 0.35 The fraction of daily feed derived from pasture while on pasture for meat animals (fy).

Ref. ER-OL, Section 2.1.3.4.3 0.05 The fraction of year vegetables are grown, (fl) approximation.

Ref. ER-OL, Section 2.1.3.4, Table 2.1-8.

0.667 The annual absolute humidity (g/m3), H, Ref. UFSAR, Table 2.3-16 6

35 ODCM Rev. 21

4.3 Requirements

Gaseous Radwaste Treatment The GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, from each reactor unit, from the site (see Figure 6-4 and Figure 6-5) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 6-4 and Figure 6-5) when averaged over 31 days would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

Applicability:

At all times:

Action:

With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, a Special Report which includes the following information:

a.

Identification of the inoperable equipment or subsystems and the reason for inoperability,

b.

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

c.

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

4.3.1 Surveillance Requirements

a. Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the methodology and parameters in Section 4.3.2.

36 ODCM Rev. 21

4.3.2 Implementation of the Requirement Where possible, consideration for expected operational evolutions (i.e., outages, etc.) should be taken in the dose projections.

Dose Proijection - Noble Gases The air dose, in mrads is determined using the methodology described in Section 4.1.2 of this manual. This information is used to determine an air dose projection for the next 31 days using the following equations:

For gamma radiation:

31 day y

= Dy MyCD (4-4)

For beta radiation:

31 day 3

= D3+/- CDI3 (4-5)

Where:

l*

= the total gamma air dose in mrads at the site boundary due to noble gases released in gaseous effluents for the previous 31 days.

DP

= the total beta air dose in mrads at the site boundary due to noble gases released in gaseous effluents for the previous 31 days.

CDy

= any current or projected change in gamma air dose, in mrads, due to noble gases released in gaseous effluents, which could have a significant impact on 31 day y.

CDP

= any current or projected change in beta air dose, in mrads, due to noble gases released in gaseous effluents, which could have a significant impact on 31 day P.

When performing the 31 day dose projection using the Gaseous Radioactive Effluent Tracking System (GRETS), Dy and DP3 will include the dose from any release permits that fall within the selected 31 day time period. As a result, the actual dose projection will often be based on the accumulated dose for a time period greater than 31 days.

37 ODCM Rev. 21

Dose Projection 13 1. 1-133. tritium. and all radionuclides in particulate form with half-lives greater than eight days The organ dose, in mrerm, is determined using the methodology described in Section 4.2.2 of this manual. This information is used to determine an organ dose projection for the next 31 days using the following equation:

31dayo

= Do+/-CDo (4-6)

Where:

Do

= the total organ dose due to 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in mrem, released in gaseous effluents for the previous 31 days.

CDo

= any current or projected change in organ dose, in mrem, which could have a significant impact on 31 dayo.

When performing the 31 day dose projection using the Gaseous Radioactive Effluent Tracking System (GRETS), Do will include the dose from any release permits that fall within the selected 31 day time period. As a result, the actual dose projection will often be based on the accumulated dose for a time period greater than 31 days.

38 ODCM Rev. 21

TABLE 4-1 Ri DOSE CONVERSION FACTORS FOR THE GROUND PLANE PATHWAY NUCLIDE T. BODY SKIN H-3 O.OOE+00 O.OOE+00 CR-51 4.66E+06 5.51E+06 MN-54 1.39E+09 1.63E+09 FE-59 2.73E+08 3.21E+08 CO-58 3.79E+08 4.44E+08 CO-60 2.15E+10 2.53E+10 ZN-65 7.47E+08 8.59E+08 SR-89 2.16E+04 2.51E+04 SR-90 O.OOE+00 O.OOE+00 ZR-95 2.45E+08 2.84E+08 SB-124 5.98E+08 6.90E+08 1-131 1.72E+07 2.09E+07 1-133 2.45E+06 2.98E+06 CS-134 6.86E+09 8.OOE+09 CS-137 1.03E+10 1.20E+10 BA-140 2.05E+07 2.35E+07 CE-141 1.37E+07 1.54E+07 CE-144 6.95E+07 8.04E+07 39 ODCM Rev. 21

TABLE 4-2 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - ADULT RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 CR-51 O.OOE+O0 O.OOE+00 4.OOE+04 2.39E+04 8.82E+03 5.31E+04 1.01E+07 MN-54 O.OOE+O0 2.97E+08 5.66E+07 O.OOE+00 8.83E+07 O.OOE+00 9.09E+08 FE-59 1.14E+08 2.68E+08 1.03E+08 O.OOE+00 O.OOE+00 7.49E+07 8.93E+08 CO-58 O.OOE+00 2.84E+07 6.38E+07 O.OOE+00 O.OOE+00 O.OOE+00 5.76E+08 CO-60 O.OOE+00 1.59E+08 3.51E+08 O.OOE+00 O.OOE+00 O.OOE+00 2.99E+09 ZN-65 3.OOE+08 9.56E+08 4.32E+08 O.OOE+00 6.39E+08 O.OOE+00 6.02E+08 SR-89 9.08E+09 O.OOE+00 2.61E+08 O.OOE+00 O.OOE+00 O.OOE+00 1.46E+09 SR-90 5.76E+11 O.OOE+00 1.41E+11 O.OOE+00 O.OOE+00 O.OOE+00 1.67E+10 ZR-95 1.08E+06 3.47E+05 2.35E+05 O.OOE+00 5.45E+05 O.OOE+00 1.10E+09 SB-124 9.53E+07 1.80E+06 3.78E+07 2.31E+05 O.OOE+00 7.42E+07 2.71E+09 1-131 5.49E+07 7.85E+07 4.50E+07 2.57E+10 1.35E+08 O.OOE+00 2.07E+07 1-133 1.39E+06 2.42E+06 7.38E+05 3.56E+08 4.22E+06 O.OOE+00 2.17E+06 CS-134 4.44E+09 1.06E+10 8.64E+09 O.OOE+00 3.42E+09 1.13E+09 1.85E+08 CS-137 6.06E+09 8.29E+09 5.43E+09 O.OOE+00 2.81E+09 9.36E+08 1.60E+08 BA-140 9.43E+07 1.19E+05 6.18E+06 O.OOE+00 4.03E+04 6.78E+04 1.94E+08 CE-141 1.73E+05 1.17E+05 1.33E+04 O.OOE+00 5.44E+04 O.OOE+00 4.48E+08 CE-144 3.12E+07 1.30E+07 1.67E+06 O.OOE+00 7.73E+06 O.OOE+00 1.05E+10 40 ODCM Rev. 21

TABLE 4-3 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - TEEN RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GIoLLI H-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 SB-124 1-131 1-133 CS-134 CS-137 BA-140 CE-141 CE-144 0.OOE+00 0.OOE+00 0.OOE+00 1.69E+08 0.OOE+00 0.OOE+00 4.11E+08 1.43E+10 7.30E+11 1.64E+06 1.47E+08 5.29E+07 1.29E+06 6.90E+09 9.86E+09 1.07E+08 2.61E+05 5.11E+07 3.36E+03 0.OOE+00 4.41E+08 3.94E+08 4.16E+07 2.42E+08 1.43E+09 0.OOE+00 0.OOE+00 5.17E+05 2.70E+06 7.41E+07 2.19E+06 1.62E+10 1.31E+10 1.3 1E+05 1.74E+05 2.12E+07 3.36E+03 5.60E+04 8.74E+07 1.52E+08 9.59E+07 5.45E+08 6.65E+08 4.10E+08 1.80E+1I 3.56E+05 5.73E+07 3.98E+07 6.68E+05 7.53E+09 4.57E+09 6.88E+06 2.OOE+04 2.75E+06 3.36E+03 3.11E+04 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 3.33E+05 2.16E+10 3.06E+08 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+O0 3.36E+03 1.23E+04 1.31E+08 O.OOE+00 0.OOE+00 0.OOE+00 9.12E+08 0.OOE+00 O.OOE+00 7.60E+05 O.OOE+00 1.28E+08 3.84E+06 5.16E+09 4.46E+09 4.44E+04 8.19E+04 1.26E+07 3.36E+03 7.99E+04 0.OOE+00 1.24E+08 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 1.28E+08 0.OOE+00 O.OOE+00 1.97E+09 1.73E+09 8.80E+04 O.OOE+00 0.OOE+00 3.36E+03 9.41E+06 9.04E+08 9.31E+08 5.74E+08 3.15E+09 6.04E+08 1.70E+09 2.05E+10 1.19E+09 2.96E+09 1.47E+07 1.66E+06 2.02E+08 1.87E+08 1.65E+08 4.98E+08 1.29E+10 41 ODCM Rev. 21

TABLE 4-4 RI DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - CHILD RECEPTOR NUCLIDES BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 SB-124 1-131 1-133 CS-134 CS-137 BA-140 CE-141 O.OOE+00 O.OOE+00 O.OOE+00 3.79E+08 O.OOE+00 O.OOE+00 7.93E+08 3.44E+10 1.22E+12 3.72E+06 3.38E+08 9.95E+07 2.36E+06 1.57E+10 2.34E+10 2.20E+08 6.15E+05 5.23E+03 O.OOE+00 6.49E+08 6.13E+08 6.21E+07 3.70E+08 2.1 1E+09 O.OOE+00 O.OOE+O0 8.17E+05 4.39E+06 1.00E+08 2.91E+06 2.57E+10 2.24E+10 1.93E+05 3.07E+05 5.23E+03 1.08E+05 1.73E+08 3.05E+08 1.90E+08 1.09E+09 1.3 1E+09 9.83E+08 3.09E+1 I 7.27E+05 1.19E+08 5.68E+07 1.10E+06 5.43E+09 3.31E+09 1.28E+07 4.55E+04 5.23E+03 6.02E+04 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 7.47E+05 3.31E+10 5.41E+08 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 5.23E+03 1.64E+04 1.82E+08 O.OOE+00 O.OOE+00 O.OOE+00 1.33E+09 0.00E+O0 O.OOE+00 1.17E+06 O.OOE+00 1.64E+08 4.85E+06 7.98E+09 7.3 IE+09 6.27E+04 1.34E+05 5.23E+03 1.10E+05 0.00E+00 1.78E+08 O.OOE+O0 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 1.88E+08 O.OOE+00 O.OOE+O0 2.86E+09 2.63E+09 1.15E+05 0.00E+O0 5.23E+03 5.75E+06 5.45E+08 6.38E+08 3.62E+08 2.05E+09 3.71E+08 1.33E+09 1.64E+10 8.52E+08 2.12E+09 8.90E+06 1.17E+06 1.39E+08 1.40E+08

1. 1 IE+08 3.83E+08 CE-144 1.24E+08 3.89E+07 6.62E+06 O.OOE+00 2.15E+07 O.OOE+00 1.01E+10 42 ODCM Rev. 21

TABLE 4-5 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - ADULT RECEPTOR NUCLIDE BONE LIVER H-3 O.OOE+00 4.33E+02 CR-51 O.OOE+00 O.OOE+00 MN-54 O.OOE+00 2.71E+06 FE-59 2.60E+07 6.1 1E+07 CO-58 O.OOE+00 2.84E+06 CO-60 O.OOE+00 2.61E+07 ZN-65 9.97E+07 3.17E+08 SR-89 3.41E+07 O.OOE+00 SR-90 4.43E+09 O.OOE+00 ZR-95 2.68E+05 8.58E+04 SB-124 2.67E+06 5.05E+04 1-131 1.36E+05 1.94E+05 1-133 4.56E-03 7.94E-03 CS-134 2.17E+08 5.17E+08 CS-137 3.11E+08 4.25E+08 BA-140 4.35E+05 5.46E+02 CE-141 8.87E+02 6.OOE+02 CE-144 4.23E+05 1.77E+05 T.BODY THYROID KIDNEY LUNG GI-LLI 4.33E+02 3.44E+02 5.18E+05 2.34E+07 6.36E+06 5.76E+07 1.43E+08 9.79E+05 1.09E+09 5.8 1E+04 1.06E+06 1.11E+05 2.42E-03 4.23E+08 2.78E+08 2.85E+04 6.80E+01 2.27E+04 4.33E+02 4.33E+02 4.33E+02 4.33E+02 2.06E+02 7.58E+01 4.57E+02 8.65E+04 O.OOE+00 8.08E+05 O.OOE+00 8.31E+06 O.OOE+00 O.OOE+00 1.71E+07 2.04E+08 O.OOE+00 O.OOE+00 O.OOE+O0 5.75E+07 O.OOE+00 O.OOE+O0 O.OOE+00 4.90E+08 O.OOE+00 2.12E+08 O.OOE+00 2.OOE+08 O.OOE+00 O.OOE+O0 O.OOE+00 5.47E+06 O.OOE+00 O.OOE+00 O.OOE+00 1.28E+08 O.OOE+00 1.35E+05 O.OOE+00 2.72E+08 6.48E+03 O.OOE+00 2.08E+06 7.59E+07 6.37E+07 3.33E+05 O.OOE+00 5.13E+04 1.17E+00 1.39E-02 O.OOE+00 7.14E-03 O.OOE+00 1.67E+08 5.56E+07 9.05E+06 O.OOE+00 1.44E+08 4.79E+07 8.22E+06 O.OOE+00 1.86E+02 3.13E+02 8.95E+05 O.OOE+O0 2.79E+02 O.OOE+00 2.29E+06 O.OOE+00 1.05E+05 O.OOE+00 1.43E+08 43 ODCM Rev. 21

TABLE 4-6 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - TEEN RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 CR-51 O.OOE+00 O.OOE+00 2.75E+02 1.53E+02 6.03E+01 3.93E+02 4.62E+04 MN-54 O.OOE+00 2.07E+06 4.11E+05 O.OOE+00 6.18E+05 O.OOE+00 4.25E+06 FE-59 2.08E+07 4.85E+07 1.87E+07 O.OOE+00 O.OOE+OO 1.53E+07 1.15E+08 CO-58 O.OOE+00 2.19E+06 5.04E+06 O.OOE+00 O.OOE+00 O.OOE+00 3.02E+07 CO-60 O.OOE+00 2.03E+07 4.56E+07 O.OOE+00 O.OOE+00 O.OOE+00 2.64E+08 ZN-65 7.01E+07 2.43E+08 1.14E+08 O.OOE+00 1.56E+08 O.OOE+00 1.03E+08 SR-89 2.88E+07 O.OOE+00 8.24E+05 O.OOE+00 O.OOE+00 O.OOE+00 3.43E+06 SR-90 2.87E+09 O.OOE+00 7.08E+08 O.OOE+00 O.OOE+00 O.OOE+00 8.05E+07 ZR-95 2.14E+05 6.76E+04 4.65E+04 O.OOE+00 9.93E+04 O.OOE+00 1.56E+08 SB-124 2.18E+06 4.02E+04 8.52E+05 4.95E+03 O.OOE+00 1.91E+06 4.40E+07 1-131 1.13E+05 1.58E+05 8.49E+04 4.61E+07 2.72E+05 O.OOE+00 3.13E+04 1-133 3.82E-03 6.48E-03 1.98E-03 9.04E-01 1.14E-02 O.OOE+OO 4.90E-03 CS-134 1.73E+08 4.07E+08 1.89E+08 O.OOE+00 1.29E+08 4.94E+07 5.06E+06 CS-137 2.58E+08 3.43E+08 1.20E+08 O.OOE+00 1.17E+08 4.54E+07 4.88E+06 BA-140 3.59E+05 4.40E+02 2.31E+04 O.OOE+00 1.49E+02 2.96E+02 5.54E+05 CE-141 7.45E+02 4.97E+02 5.71E+01 O.OOE+00 2.34E+02 O.OOE+OO 1.42E+06 CE-144 3.56E+05 1.47E+05 1.91E+04 O.OOE+00 8.80E+04 O.OOE+00 8.96E+07 44 ODCM Rev. 21

TABLE 4-7 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - CHILD RECEPTOR NUCLIDES BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 3.12E+02 3.12E+02 3.12E+02 3.12E+02 3.12E+02 3.12E+02 CR-51 O.OOE+00 O.OOE+00 4.29E+02 2.38E+02 6.51E+01 4.35E+02 2.28E+04 MN-54 O.OOE+00 2.37E+06 6.31E+05 O.OOE+00 6.64E+05 O.OOE+00 1.99E+06 FE-59 3.68E+07 5.96E+07 2.97E+07 O.OOE+00 O.OOE+00 1.73E+07 6.20E+07 CO-58 O.OOE+00 2.55E+06 7.82E+06 O.OOE+00 O.OOE+00 O.OOE+O0 1.49E+07 CO-60 O.OOE+00 2.40E+07 7.09E+07 O.OOE+00 O.OOE+00 O.OOE+00 1.33E+08 ZN-65 1.05E+08 2.80E+08 1.74E+08 O.OOE+00 1.77E+08 O.OOE+00 4.92E+07 SR-89 5A5E+07 O.OOE+00 1.56E+06 O.OOE+00 O.OOE+00 O.OOE+00 2.11E+06 SR-90 3.70E+09 O.OOE+00 9.39E+08 O.OOE+00 O.OOE+00 O.OOE+00 4.99E+07 ZR-95 3.81E+05 8.36E+04 7.45E+04 O.OOE+00 1.20E+05 O.OOE+00 8.73E+07 SB-124 3.95E+06 5.12E+04 1.38E+06 8.72E+03 O.OOE+00 2.19E+06 2.47E+07 1-131 2.09E+05 2.1 1E+05 1.20E+05 6.96E+07 3.A6E+05 O.OOE+00 1.87E+04 1-133 7.09E-03 8.77E-03 3.32E-03 1.63E+00 1A6E-02 O.OOE+00 3.53E-03 CS-134 3.05E+08 5.OOE+08 1.06E+08 O.OOE+00 1.55E+08 5.56E+07 2.70E+06 CS-137 4.75E+08 4.55E+08 6.71E+07 O.OOE+00 1.48E+08 5.33E+07 2.85E+06 BA-140 6.63E+05 5.81E+02 3.87E+04 O.OOE+00 1.89E+02 3.46E+02 3.36E+05 CE-141 1.40E+03 6.99E-02 1.04E+02 O.OOE+00 3.07E+02 O.OOE+00 8.72E+05 CE-144 6.72E+05 2.1 IE+05 3.58E+04 O.OOE+00 1.17E+05 O.OOE+00 5.49E+07 45 ODCM Rev. 21

TABLE 4-8 RI DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - ADULT RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 SB-124 1-131 1-133 CS-134 CS-137 BA-140 CE-141 CE-144 O.OOE+00 O.OOE+00 O.OOE+00 9.69E+06 O.OOE+00 O.OOE+00 6.34E+08 4.90E+08 2.43E+10 3.39E+02 9.1IE+06 7.77E+07 1.02E+06 2.83E+09 3.83E+09 7.1 1E+06 8.73E+03 1.01E+06 1.02E+03 O.OOE+00 3.99E+06 2.28E+07 1.74E+06 8.41E+06 2.02E+09 O.OOE+00 O.OOE+00 1.09E+02 1.72E+05 1.1 1E+08 1.77E+06 6.73E+09 5.24E+09 8.93E1+03 5.90E+03 4.21E+05 1.02E+03 8.28E+03 7.61E+05 8.73E+06 3.90E+06 1.85E+07 9.12E+08 1.41E+07 5.96E+09 7.37E+01 3.61E+06 6.37E+07 5.39E+05 5.50E+09 3A3E+09 4.66E+05 6.70E+02 5.41E+04 1.02E+03 4.95E+03 O.OOE+00 O.OOE+00 O.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 2.21E+04 3.64E+10 2.60E+08 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1.02E+03 1.82E+03 1.19E+06 O.OOE+00 O.OOE+00 O.OOE+00 1.35E+09 O.OOE+00 O.OOE+00 1.71E+02 O.OOE+00 1.91E+08 3.08E+06 2.18E+09 1.78E+09 3.04E+03 2.74E+03 2.50E+05 1.02E+03 1.101E+04 0.00E+O0 6.36E+06 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 7.09E+06 O.OOE+00 O.OOE+O0 7.23E+08 5.91E+08 5.11 E+03 O.OOE+00 O.OOE+00 1.02E+03 2.08E+06 1.22E+07 7.59E+07 3.53E+07 1.58E+08 1.27E+09 7.86E+07 7.02E+08 3.45E+05 2.59E+08 2.93E+07 1.59E+06 1.18E+08 1.01E+08 1.46E+07 2.26E+07 3.41E+08 46 ODCM Rev. 21

TABLE 4-9 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - TEEN RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 1.33E+03 1.33E+03 1.33E+03 1.33E+03 1.33E+03 1.33E+03 CR-51 O.OOE+00 O.OOE+00 1.45E+04 8.03E+03 3.17E+03 2.06E+04 2.43E+06 MN-54 O.OOE+00 6.64E+06 1.32E+06 O.OOE+00 1.98E+06 O.OOE+00 1.36E+07 FE-59 1.69E+07 3.95E+07 1.52E+07 O.OOE+00 O.OOE+00 1.24E+07 9.33E+07 CO-58 O.OOE+00 2.93E+06 6.76E+06 O.OOE+00 O.OOE+00 O.OOE+00 4.04E+07 CO-60 O.OOE+00 1.42E+07 3.21E+07 O.OOE+00 O.OOE+00 O.OOE+00 1.86E+08 ZN-65 9.74E+08 3.38E+09 1.58E+09 O.OOE+00 2.17E+09 O.OOE+00 1.43E+09 SR-89 9.03E+08 O.OOE+00 2.59E+07 O.OOE+00 O.OOE+00 O.OOE+00 1.08E+08 SR-90 3.43E+10 O.OOE+00 8.48E+09 O.OOE+00 O.OOE+00 O.OOE+00 9.64E+08 ZR-95 5.94E+02 1.87E+02 1.29E+02 O.OOE+00 2.75E+02 O.OOE+00 4.32E+05 SB-124 1.62E+07 2.99E+05 6.34E+06 3.69E+04 O.OOE+00 1.42E+07 3.27E+08 1-131 1.41E+08 1.98E+08 1.06E+08 5.76E+10 3.40E+08 O.OOE+00 3.91E+07 1-133 1.86E+06 3.15E+06 9.60E+05 4.39E+08 5.52E+06 O.OOE+00 2.38E+06 CS-134 4.91E+09 1.16E+10 5.36E+09 O.OOE+00 3.67E+09 1.40E+09 1.44E+08 CS-137 6.95E+09 9.24E+09 3.22E+09 O.OOE+00 3.15E+09 1.22E+09 1.32E+08 BA-140 1.28E+07 1.57E+04 8.27E+05 O.OOE+00 5.33E+03 1.06E+04 1.98E+07 CE-141 1.60E+04 1.07E+04 1.23E+03 O.OOE+00 5.03E+03 O.OOE+00 3.06E+07 CE-144 1.86E+06 7.68E+05 9.97E+04 O.OOE+00 4.59E+05 O.OOE+00 4.67E+08 47 ODCM Rev. 21

TABLE 4-10 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - CHILD RECEPTOR NUCLIDES BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 2.09E+03 2.09E+03 2.09E+03 2.09E+03 2.09E+03 2.09E+03 CR-51 O.OOE+00 0.01E+00 2.95E+04 1.64E+04 4.47E+03 2.99E+04 1.56E+06 MN-54 O.OOE+00 9.94E+06 2.65E+06 O.OOE+00 2.79E+06 O.OOE+00 8.34E+06 FE-59 3.92E+07 6.35E+07 3.16E+07 O.OOE+00 O.OOE+00 1.84E+07 6.61E+07 CO-58 O.OOE+00 4.48E+06 1.37E+07 O.OOE+00 O.OOE+00 O.OOE+00 2.61E+07 CO-60 O.OOE+00 2.21E+07 6.52E+07 O.OOE+00 O.OOE+00 O.OOE+00 1.23E+08 ZN-65 1.91E+09 5.09E+09 3.17E+09 O.OOE+00 3.21E+09 O.OOE+00 8.95E+08 SR-89 2.23E+09 O.OOE+00 6.38E+07 O.OOE+00 O.OOE+00 O.OOE+00 8.65E+07 SR-90 5.80E+10 O.OOE+00 1.47E+10 O.OOE+00 O.OOE+00 O.OOE+00 7.81E+08 ZR-95 1.38E+03 3.03E+02 2.70E+02 O.OOE+00 4.34E+02 O.OOE+00 3.16E+05 SB-124 3.84E+07 4.99E+05 1.35E+07 8.49E+04 O.OOE+00 2.13E+07 2.41E+08 1-131 3.42E+08 3.44E+08 1.96E+08 1.14E+II 5.65E+08 O.OOE+00 3.06E+07 1-133 4.5 1E+06 5.57E+06 2.1 1E+06 1.04E+09 9.29E+06 O.OOE+00 2.25E+06 CS-134 1.13E+10 1.86E+10 3.92E+09 O.OOE+00 5.76E+09 2.07E+09 1.00E+08 CS-137 1.67E+10 1.60E+10 2.36E+09 O.OOE+00 5.22E+09 1.88E+09 1.00E+08 BA-140 3.10E+07 2.71E+04 1.81E+06 O.OOE+00 8.83E+03 1.62E+04 1.57E+07 CE-141 3.94E+04 1.97E+04 2.92E+03 0.00E+00 8.62E+03 O.OOE+00 2.45E+07 CE-144 4.57E+06 1.43E+06 2.44E+05 O.OOE+00 7.94E+05 O.OOE+00 3.74E+08 48 ODCM Rev. 21

TABLE 4-11 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - INFANT RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 0.OOE+00 3.18E+03 3.18E+03 3.18E+03 3.18E+03 3.18E+03 3.18E+03 CR-51 0.OOE+00 O.OOE+00 4.67E+04 3.05E+04 6.66E+03 5.93E+04 1.36E+06 MN-54 O.OOE+00 1.85E+07 4.19E+06 O.OOE+00 4.10E+06 O.OOE+00 6.79E+06 FE-59 7.32E+07 1.28E+08 5.04E+07 0.OOE+00 O.OOE+00 3.78E+07 6.1 1E+07 CO-58 O.OOE+00 8.96E+06 2.23E+07 O.OOE+00 O.OOE+00 O.OOE+00 2.23E+07 CO-60 O.OOE+00 4.52E+07 1.07E+08 O.OOE+00 O.OOE+00 O.OOE+00 1.07E+08 ZN-65 2.57E+09 8.81E+09 4.06E+09 O.OOE+00 4.27E+09 O.OOE+00 7.44E+09 SR-89 4.25E+09 O.OOE+00 1.22E+08 O.OOE+00 O.OOE+00 O.OOE+00 8.74E+07 SR-90 6.31E+10 O.OOE+O0 1.61E+10 O.OOE+00 O.OOE+O0 0.OOE+00 7.88E+08 ZR-95 2.45E+03 5.97E+02 4.23E+02 O.OOE+00 6.43E+02 O.OOE+00 2.97E+05 SB-124 7.41E+07 1.09E+06 2.30E+07 1.97E+05 O.OOE+00 4.64E+07 2.29E+08 1-131 7.14E+08 8.42E+08 3.70E+08 2.77E+ 1I 9.83E+08 O.OOE+00 3.OOE+07 1-133 9.52E+06 1.39E+07 4.06E+06 2.52E+09 1.63E+07 O.OOE+00 2.35E+06 CS-134 1.82E+10 3.40E+10 3.44E+09 O.OOE+00 8.76E+09 3.59E+09 9.24E+07 CS-137 2.67E+10 3.13E+10 2.22E+09 O.OOE+00 8.39E+09 3.40E+09 9.78E+07 BA-140 6.37E+07 6.37E+04 3.28E+06 O.OOE+00 1.51E+04 3.91E+04 1.57E+07 CE-141 7.81E+04 4.77E+04 5.61E+03 O.OOE+00 1.47E+04 O.OOE+00 2.46E+07 CE-144 6.55E+06 2.68E+06 3.67E+05 0.OOE+00 1.08E+06 O.OOE+00 3.76E+08 49 ODCM Rev. 21

TABLE 4-12 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - ADULT RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 CR-51 O.OOE+00 O.OOE+00 1.00E+02 5.95E+01 2.28E+01 1.44E+04 3.32E+03 MN-54 O.OOE+00 3.96E+04 6.30E+03 O.OOE+00 9.84E+03 1.40E+06 7.74E+04 FE-59 1.18E+04 2.78E+04 1.06E+04 O.OOE+00 O.OOE+00 1.02E+06 1.88E+05 CO-58 O.OOE+00 1.58E+03 2.07E+03 O.OOE+00 O.OOE+00 9.28E+05 1.06E+05 CO-60 O.OOE+00 1.15E+04 1.48E+04 O.OOE+00 O.OOE+00 5.97E+06 2.85E+05 ZN-65 3.24E+04 1.03E+05 4.66E+04 O.OOE-0 6.90E+04 8.64E+05 5.34E+04 SR-89 3.04E+05 O.OOE+00 8.72E+03 O.OOE+00 O.OOE+00 1.40E+06 3.50E+05 SR-90 9.92E+07 O.OOE+00 6.1OE+06 O.OOE+00 O.OOE+00 9.60E+06 7.22E+05 ZR-95 1.07E+05 3.44E+04 2.33E+04 O.OOE+00 5.42E+04 1.77E+06 1.50E+05 SB-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 O.OOE+00 2.48E+06 4.06E+05 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 O.OOE+00 6.28E+03 1-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 O.OOE+00 8.88E+03 CS-134 3.73E+05 8.48E+05 7.28E+05 O.OOE+00 2.87E+05 9.76E+04 1.04E+04 CS-137 4.78E+05 6.21E+05 4.28E+05 O.OOE+00 2.22E+05 7.52E+04 8.40E+03 BA-140 3.90E+04 4.90E+01 2.57E+03 O.OOE+00 1.67E+01 1.27E+06 2.18E+05 CE-141 1.99E+04 1.35E+04 1.53E+03 O.OOE+00 6.26E+03 3.62E+05 1.20E+05 CE-144 3.43E+06 1.43E+06 1.84E+05 O.OOE+00 8.48E+05 7.78E+06 8.16E+05 50 ODCM Rev. 21

TABLE 4-13 RI DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - TEEN RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 1.27E+03 1.27E+03 1.27E+03 1.27E+03 1.27E+03 1.27E+03 CR-51 O.OOE+00 O.OOE+00 1.35E+02 7.50E+01 3.07E+01 2.10E+04 3.OOE+03 MN-54 O.OOE+00 5.11E+04 8.40E+03 O.OOE+00 1.27E+04 1.98E+06 6.68E+04 FE-59 1.59E+04 3.70E+04 1.43E+04 O.OOE+00 O.OOE+00 1.53E+06 1.78E+05 CO-58 O.OOE+00 2.07E+03 2.78E+03 O.OOE+O0 O.OOE+00 1.34E+06 9.52E+04 CO-60 O.OOE+00 1.51E+04 1.98E+04 O.OOE+00 O.OOE+00 8.72E+06 2.59E+05 ZN-65 3.86E+04 1.34E+05 6.24E+04 O.OOE+00 8.64E+04 1.24E+06 4.66E+04 SR-89 4.34E+05 O.OOE+00 1.25E+04 O.OOE+00 O.OOE+00 2.42E+06 3.71E+05 SR-90 1.08E+08 O.OOE+00 6.68E+06 O.OOE+00 O.OOE+00 1.65E+07 7.65E+05 ZR-95 1.46E+05 4.58E+04 3.15E+04 O.OOE+00 6.74E+04 2.69E+06 1.49E+05 SB-124 4.30E+04 7.94E+02 1.68E+04 9.76E+01 O.OOE+00 3.85E+06 3.98E+05 1-131 3.54E+04 4.91E+04 2.64E+04 1.46E+07 8.40E+04 O.OOE+00 6.49E+03 1-133 1.22E+04 2.05E+04 6.22E+03 2.92E+06 3.59E+04 O.OOE+00 1.03E+04 CS-134 5.02E+05 1.13E+06 5.49E+05 O.OOE+00 3.75E+05 1.46E+05 9.76E+03 CS-137 6.70E+05 8.48E+05 3.1 1E+05 O.OOE+00 3.04E+05 1.21E+05 8.48E+03 BA-140 5.47E+04 6.70E+01 3.52E+03 O.OOE+00 2.28E+01 2.03E+06 2.29E+05 CE-141 2.84E+04 1.90E+04 2.17E+03 O.OOE+00 8.88E+03 6.14E+05 1.26E+05 CE-144 4.89E+06 2.02E+06 2.62E+05 O.OOE+00 1.21E+06 1.34E+07 8.64E+05 51 ODCM Rev. 21

TABLE 4-14 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - CHILD RECEPTOR NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H-3 O.OOE+00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 CR-51 O.OOE+00 O.OOE+00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 MN-54 O.OOE+00 4.29E+04 9.51E+03 O.OOE+00 1.00E+04 1.58E+06 2.29E+04 FE-59 2.07E+04 3.34E+04 1.67E+04 O.OOE+00 O.OOE+00 1.27E+06 7.07E+04 CO-58 O.OOE+00 1.77E+03 3.16E+03 O.OOE+00 O.OOE+O0 1.11E+06 3.44E+04 CO-60 O.OOE+00 1.3 1E+04 2.26E+04 O.OOE+00 O.OOE+00 7.07E+06 9.62E+04 ZN-65 4.26E+04 1.13E+05 7.03E+04 O.OOE+00 7.14E+04 9.95E+05 1.63E+04 SR-89 5.99E+05 O.OOE+00 1.72E+04 O.OOE+00 O.OOE+00 2.16E+06 1.67E+05 SR-90 1.01E+08 O.OOE+00 6.44E+06 O.OOE+00 O.OOE+00 1.48E+07 3.43E+05 ZR-95 1.90E+05 4.18E+04 3.70E+04 O.OOE+00 5.96E+04 2.23E+06 6.1 IE+04 SB-124 5.74E+04 7.40E+02 2.OOE+04 1.26E+02 O.OOE+00 3.24E+06 1.64E+05 1-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 O.OOE+00 2.84E+03 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 O.OOE+00 5.48E+03 CS-134 6.51E+05 1.01E+06 2.25E+05 O.OOE+00 3.30E+05 1.21E+05 3.85E+03 CS-137 9.07E+05 8.25E+05 1.28E+05 0.00E+00 2.82E+05 1.04E+05 3.62E+03 BA-140 7.40E+04 6.48E+01 4.33E+03 O.OOE+00 2.1 IE+01 1.74E+06 1.02E+05 CE-141 3.92E+04 1.95E+04 2.90E+03 O.OOE+00 8.55E+03 5.44E+05 5.66E+04 CE-144 6.77E+06 2.12E+06 3.61E+05 O.OOE+00 1.17E+06 1.20E+07 3.89E+05 52 ODCM Rev. 21

TABLE 4-15 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - INFANT RECEPTOR NUCLIDE H-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 SB-124 1-131 1-133 CS-134 CS-137 BA-140 CE-141 BONE 0.00E+00 O.OOE+00 0.00E+00 1.36E+04 0.00E+o0 0.OOE+00 1.93E+04 3.98E+05 4.09E+07 1.15E+05 3.79E+04 3.79E+04 1.32E+04 3.96E+05 5.49E+05 5.60E+04 2.77E+04 LIVER 6.47E+02 0.OOE+00 2.53E+04 2.35E+04 1.22E+03 8.02E+03 6.26E+04 0.OOE+00 O.OOE+00 2.79E+04 5.56E+02 4.44E+04 1.92E+04 7.03E+05 6.12E+05 5.60E+01 1.67E+04 T.BODY 6.47E+02 8.95E+01 4.98E+03 9.48E+03 1.82E+03 1.18E+04 3.11E+04 1.14E+04 2.59E+06 2.03E+04 1.20E+04 1.96E+04 5.60E+03 7.45E+04 4.55E+04 2.90E+03 1.99E+03 THYROID 6.47E+02 5.75E+01 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 1.01E+02 1.48E+07 3.56E+06 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 KIDNEY 6.47E+02 1.32E+01 4.98E+03 O.OOE+00 0.OOE+00 0.00E+00 3.25E+04 O.OOE+00 0.OOE+00 3.1 1E+04 O.OOE+00 5.18E+04 2.24E+04 1.90E+05 1.72E+05 1.34E+01 5.25E+03 LUNG 6.47E+02 1.28E+04 1.00E+06 1.02E+06 7.77E+05 4.51E+06 6.47E+05 2.03E+06 1.122E+07 1.75E+06 2.65E+06 0.OOE+00 0.OOE+00 7.97E+04 7.13E+04 1.60E+06 5.17E+05 GI-LLI 6.47E+02 3.57E+02 7.06E+03 2.48E+04 1.11E+04 3.19E+04 5.14E+04 6.40E+04 1.31E+05 2.17E+04 5.91E+04 1.06E+03 2.16E+03 1.33E+03 1.33E+03 3.84E+04 2.16E+04 CE-144 3.19E+06 1.21E+06 1.76E+05 0.OOE+00 5.38E+05 9.84E+06 1.48E+05 53 ODCM Rev. 21

0 00

.~t.

0

.t*.

0 w

00qw-CO 0

00 0

0 0

t-0)

Vd.

M-M 0

0q 0

Cl, CO O C

C CO CO Cu CO CO CO O

O C

C i

0 0

CO i Cl, It 9

C-0 00 009

'Ii C4 0

,C4 0 0t C--9 9

'0 0

C--9 Cs'9

'0 0

C--

0 0

C-i 0

-4i 0

C-00 0

0 0

0~

m,~

0e COq

-ISC C

O C

C O

C fn C-i C-'en 04 N

Itr OR c--

00 Cl, w

%n l in W)0C-%

n t-:

0 N49i 00 00 i

0I 0, 0" 0

0)

I U,

4) 4)

0 0.

I-0 4..

4)

CO CO CO 0.

0 CO U,

4) 6~

4) 4)

C.)

CO U,

CO 4)CO 4)

CO 00 00 00 4)

CO 0o CO N0 0~

Cl, 0;

en OR Cs'

"'-00 diCi d V) en 99 N0 4

00 4:

CO CO U,

CO C.)

U, U,

CO 0

CO 00 CO 0

U I

I z

54 ODCM Rev. 21

TABLE 4-17 PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 2 X/Q RESIDENCE(b)

D/Q X/Q GARDEN(b)

D/Q X/Q MILK(b)

DIQ DIRECTION (Sec/n 3)

Dist. Miles (M"2)

(Sec/rn 3)

Dist. Miles (m"2)

(Sec/rn 3)

Dist. Miles (m"2)

N 2.73E-06 1.5 2.92E-09 2.39E-06 1.7 2.35E-09 7.03E-07 (a) 3.48E-10 NNE 2.20E-06 1.5 3.87E-09 2.20E-06 1.5 3.87E-09 4.70E-07 (a) 4.04E-10 NE 1.85E-06 2.0 3.55E-09 1.57E-06 2.3 2.78E-09 5.77E-07 (a) 6.51E-10 ENE 1.03E-06 2.7 1.08E-09 1.03E-06 2.7 1.08E-09 3.86E-07 (a) 2.86E-10 E

8.80E-07 3.0 6.06E-10 3.71E-07 (a) 1.87E-10 3.71E-07 (a) 1.87E-10 ESE 6.25E-07 3.7 2.76E-10 3.96E-07 4.7 1.51E-10 3.96E-07 4.7 1.51E-10 goat SE 9.06E-07 4.0 2.72E-10 9.06E-07 4.0 2.72E-10 5.84E-07 (a) 1.52E-10 SSE 1.34E-06 4.5 2.81E-10 1.09E-06 (a) 2.15E-10 1.09E-06 (a) 2.15E-10 S

2.63E-06 4.5 5.01E-10 2.19E-06 5.0 3.88E-10 2.19E-06 5.0 3.88E-10 cow SSW 3.48E-06 3.2 9.19E-10 2.28E-06 (a) 4.53E-10 2.28E-06 (a) 4.53E-10 SW 2.93E-06 2.7 9.75E-10 1.58E-06 (a) 3.56E-10 1.58E-06 (a) 3.56E-10 WSW 2.01E-06 2.5 1.16E-09 8.55E-07 (a) 3.18E-10 8.55E-07 (a) 3.18E-10 W

7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 WNW 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 NW 7.84E-07 4.0 4.88E-10 7.84E-07 4.0 4.88E-10 6.02E-07 (a) 3.27E-10 NNW 1.46E-06 2.0 1.47E-09 5.20E-07 5.0 3.04E-10 5.20E-07 (a) 3.04E-10 o

(a) 5-mile value used since there is no pathway located within the sector up to five miles.

¢3 (b) Controlling locations are discussed in Appendix A.

e

References:

1984 Land Use Census (letter ANPM-21221-JRM/LEB). NUS Corporation letters NUS-ANPP-1385 and NUS-ANPP-1386.

!2

TABLE 4-18 PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 3 DIRECTION N

NNE NE ENE E

ESE SE SSE S

SSW SW WSW W

WNW NW NNW XIQ (Sec/Ms) 2.58E-06 1.8513-06 1.6613-06 8.75E-07 8.90E-07 6.37E-07 5.84E-07 1.36E-06 2.6513-06 3.6413-06 3.19E-06 2.12E-06 7.54E-07 6.0313-07 6.8313-07 1.3413-06 RESEIENCE(b)

Dist. Miles 1.8 1.7 2.2 2.9 3.0 3.7 (a) 4.4 4.2 3.1 2.5 2.4 (a)

(a) 4.3 2.2 D/Q (m-2) 2.47E-09 2.97E-09 3.OOE-09 8.86E-10 6.17E-10 2.84E-10 1.52E-10 2.88E-10 5.25E-10 9.82E-10 1.11E-09 1.26E-09 4.44E-10 3.25E-10 4.05E-10 1.26E-09 X/Q (Sec/rn) 2.42E-06 1.85E-06 1.48E-06 8.7513-07 4.0613-07 5.8013-07 5.84E-07 1.09E-06 2.25E-06 2.2813-06 1.58E-06 8.55E-07 7.54E-07 6.0313-07 6.8213-07 5.1611-07 GARDEN(b)

Dist. Miles 1.9 1.7 2.4 2.9 4.6 4.0 (a)

(a) 4.9 (a)

(a)

(a) 4.3 5.0 D/Q (m72) 2.22E-09 2.97E-09 2.54E-09 8.86E-10 2.15E-10 2.46E-10 1.52E-10 2.15E-10 4.06E-10 4.53E-10 3.56E-10 3.188E-10 4.44E-10 3.25E-10 4.05E-10 3.011E-10 XIQ (Sed/r3) 7.03E-07 4.7013-07 5.77E-07 3.86E-07 4.2513-07 3.73E-07 5.84E-07 1.09E-06 2.31E-06 2.28E-06 1.58E-06 8.55E-07 7.5413-10 6.03E-07 6.02E-07 5.2013-07 MILK(b)

Dist. Miles (a)

(a)

(a) 4.5 (a)

(a)

(a) 4.8 (a)

(a)

D/Q (m-2) 3.48E-10 4.04E-10 6.51E-10 2.86E-10 2.31E-10 goat 1.37E-10 1.52E-10 2.15E-10 4.21E-10 cow 4.53E-10 3.56E-10 3.188E-10 4.44E-10 3.25E-I0 3.27E-10 3.04E-10 0

¢.2 (a) 5-mile value used since there is no pathway located within the sector up to five miles.

(b) Controlling locations are discussed in Appendix A.

References:

1984 Land Use Census (letter ANPM-21221-JRM/LEB). NUS Corporation letters NUS-ANPP-1385 and NUS-ANPP-1386.

4.4 Requirements

Liquid Effluents The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (See Figure 6-4 and Figure 6-5) shall be limited:

a.

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

b.

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

Applicability:

At all times.

Action:

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

4.4.1 Surveillance Requirements Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

4.4.2 Implementation of the Requirements This Requirement does not require implementation guidance. There are no offsite liquid effluent releases.

57 ODCM Rev. 21

5.0 TOTAL DOSE AND DOSE TO PUBLIC ONSITE

5.1 Requirement

Total Dose The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to direct radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

Applicability:

At all times.

Action:

With the calculated doses from the release of radioactive materials in liquid and gaseous effluents exceeding twice the limidts of Section 4.4a, 4.4b, 4.1a, 4.1b, 4.2a or 4.2b calculations shall be made including direct radiation contributions from the reactor units (including outside storage tanks, etc.) to determine whether the above limits of Section 5.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report as defined in 10 CFR 20.2203(a)(4), 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, for the calendar year that includes the release(s) covered by this report.

It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report within 30 days is considered a timely request, and a variance is granted until staff action on the request is complete.

5.1.1 Surveillance Requirements

a. Cumulative dose contributions from the gaseous effluents shall be determined in accordance with the surveillance requirements of Section 4.4.1, 4. 1.1 and 4.2.1 and in accordance with the methodology and parameters contained in Section 5.1.2.

b. Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in Section 5.1.2. This requirement is applicable only under conditions set forth in Section 5.1, Action.

5.1.2 Implementation of the Requirement Since all other uranium fuel cycle sources are greater than 20 miles away, only the PVNGS site need be considered.

The total dose to any MEMBER OF THE PUBLIC will be determined based on a sum of the doses from all three units' releases and doses from direct radiation from PVNGS.

58 58 ODCM Rev. 21

This dose evaluation is performed annually and submitted with the Annual Radioactive Effluent Release Report to assure compliance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

NUREG-0543, Methods for Demonstrating LWR Compliance With the EPA Uranium Fuel Cycle Standard (40 CFR Part 190), February 1980, provides a discussion on compliance with 40 CFR Part 190 in relation to the Radiological Environmental Technical Specifications for sites of up to four nuclear power reactors. The NUREG concludes that as long as a nuclear plant site operates at a level below the 10 CFR Part 50, Appendix I reporting requirements, and there is no significant source of direct radiation from the site, no extra analysis is required to demonstrate compliance with 40 CFR Part 190. As a result, this dose evaluation will also be performed whenever calculated doses associated with effluent releases exceed twice the limits of Section 4.4a, 4.4b, 4.1a, 4.1b, 4.2a or 4.2b.

Dose Contribution from Liquid and Gaseous Effluents The annual whole body dose accumulated by a MEMBER OF THE PUBLIC for the noble gases released in gaseous effluents is determined by using the following equation:

DWB

= (3.17E-08) I- [(Ki) (X/Q)uNrr (Qi)]

(5-1)

Where:

KV

= the whole body dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrem/yr per JtCi/m3 from Table 3-3.

Qi

= the integrated release of radionuclide i, in gCi for the previous calendar year.

(X/Q)yuNP

-= the highest calculated annual average dispersion parameter, in sec/m 3, for a particular unit, at the controlling location, from Table 4-16, 4-17, or 4-18, or concurrent meteorological data if available.

=2.92E-06 from Unit 1

=2.19E-06 from Unit 2

=2.31E-06 from Unit 3 DWB

= the annual whole body dose in mrem to a MEMBER OF THE PUBLIC at the controlling location due to noble gases released in gaseous effluents.

3.17E-08

= the inverse of seconds in a year (yr/sec).

59 ODCM Rev. 21

The annual dose to any organ accumulated by a MEMBER OF THE PUBLIC for iodine-13 1, iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents is determined by using the following equation:

Do

= (3.17E-08) Y- [Y-k(RikWk) (Qi)]

(5-2)

Where:

Do

= the total annual organ dose from gaseous effluents to a MEMBER OF THE PUBLIC, in mrem, at the controlling location.

Qi

= the integrated release of radionuclide i, in gtCi, for the previous calendar year.

Rik

= the dose factor for each identified radionuclide i, for pathway k (for the inhalation pathway in mrem/yr per ItCi/m3 and for the food and ground plane pathways in m2-nmrem/yr per JiCi/sec) at the controlling location. The Rnk's for each age group are given in Tables 4-1 through 4-15.

WK

= the highest annual average dispersion or deposition parameter for the particular unit, used for estimating the total annual organ dose to a MEMBER OF THE PUBLIC at the controlling location for the particular unit.

= (X/Q)uNrT, in sec/m 3 for the inhalation pathway and for all tritium calculations, for organ dose at the controlling location, from Table 4-16, 4-17, or 4-18, or concurrent meteorological data if available.

=2.92E-06 from Unit 1

=2.19E-06 from Unit 2

=2.31E-06 from Unit 3

= (D/Q)UNT, in m"2, for the food and ground plane pathways, for organ dose at the controlling location, from Table 4-16, 4-17, or 4-18, or concurrent meteorological data if available.

=3.25E-09 from Unit 1

=3.88E-10 from Unit 2

=4.21E-10 from Unit 3 3.17E-08

= the inverse of seconds in a year (yr/sec).

60 ODCM Rev. 21

Dose Due to Direct Radiation The component of dose to a MEMBER OF THE PUBLIC due to direct radiation will be evaluated by first determining the direct radiation dose at the site boundary in each sector, and then extrapolating the site boundary dose to the controlling location by the inverse square law of distance.

Dose from Radioactive Liquid and Gaseous Effluents to MEMBERS OF THE PUBLIC due to their activities within the SITE BOUNDARY.

These activities have been determined to be limited to the vicinity of the Energy Information Center (EIC) located inside the SIT7E BOUNDARY. An assumption was made that no MEMBER OF THE PUBLIC would spend more than eight hours per year at this location.

However this calculation has been historically performed assuming an occupancy factor of one (implying continuous occupancy over the entire year).

A X/Q, determined for the Energy Information Center, will be used for this assessment.

Equations 5-1 and 5-2 in Section 5.1.2 should be used for this assessment.

61 61 ODCM Rev. 21

6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

6.1 Requirement

REMP The radiological environmental monitoring program shall be conducted as specified in Table 6-1, based on locations determined using data from the pre-operational monitoring period; and/or the operational monitoring period indicating a need to make changes in the program.

Applicability:

At all times.

Action:

a. With the radiological environmental monitoring program not being conducted as specified in Table 6-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, as required by Section 7.2, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 6-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose* to A MEMBER OF THE PUBLIC is less than the calendar year limits of Section 4.4, 4.1 and 4.2. When more than one of the radionuclides in Table 6-2 are detected in the sampling medium, this report shall be submitted if-concentration (1) + concentration (2) reporting level (1) reporting level (2)

When radionuclides other than those in Table 6-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose* to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Section 4.4, 4.1 and 4.2. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. W*ith milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 6-1, identify locations for obtaining replacement samples and add them to the Radiological Environmental Monitoring Program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program.

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

62 62 ODCM Rev. 21

6.1.1 Surveillance Requirements

a. The radiological environmental monitoring samples shall be collected pursuant to Table 6-1 from the specific locations given in Table 6-4 and Figure 6-1 and Figure 6-2 and shall be analyzed pursuant to the requirements of Table 6-1, and the detection capabilities required by Table 6-3.

6.1.2 Implementation of the Requirements The results of the radiological environmental monitoring program are intended to supplement the results of the radiological effluent monitorinig program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected based on the effluent measurements and modeling of the environmental exposure pathways.

Thus the specified environmental monitoring program provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures to individuals resulting from station operation.

This requirement is implemented by Nuclear Administrative and Technical Manual procedures.

63 ODCM Rev. 21

TABLE 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Type and Frequency and/or Sample Samples and Sample Locations' Collection Frequency' of Analysisd Airborne Samples from 5 locations: 4 Continuous sampling Gross beta weeklyc, samples at or near the SITE collected weekly, or 1-131 weekly; gamma Radioiodine and BOUNDARIES (#14A, 15, 29,40) more frequently if isotopic analysis of particulates including 3 different sectors of the required by dust composite (by highest calculated annual average

loading, location) quarterly.

ground level D/Q.*

I sample (#40) from areas of special interest, which is from the vicinity of a community having the highest calculated annual average D/Q.

1 sample (#6A) from a control location 15-30 km (9-18 mi) distant and in the least prevalent wind direction.e Direct radiationb Forty (40) routine monitoring Quarterly Gamma dose stations (#5-40, #42, #44, #46, #50) quarterly.

either with two or more 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 (16 locations);

An outer ring of stations, one in each meteorological sector in the 6-8 km (4-5 ml) range from the site (16 locations); and The balance of the stations (8 locations) to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

D/Q refers to average annual relative ground deposition rate.

64 ODCM Rev. 21

TABLE 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Type and Frequency and/or Sample Samples and Sample Locations' Collection Frequency' of Analysisd Waterhorne Surface Water storage reservoir (#60)

Monthly composite of Gamma isotopic Evaporation pond #1 (#59) weekly grab sample.

analysis monthly; Evaporation pond #2 (#63) tritium quarterly.

Ground 2 onsite wellsf (#57, #58)

Quarterly grab sample Tritium and gamma isotopic analysis quarterly.

Drinking (well) 3 wells from surrounding Composite sample of 1-1 31 analysis on each residences (#46, #48, #49) that weekly grab samples composite when the would be affected by its discharge.

over 2-week period dose calculated for when 1-131 analysis is the consumption of performed, monthly the water is greater composite of weekly than I mrem per grab samples otherwise year.g Composite for gross beta and gamma isotopic analyses monthly. Composite for tritium analysis quarterly.

Ingestion Samples from milking animals in Semimonthly for Gamma isotopic and 3 locations within 5 km distant animals on pasture; 1-131 analysis Milk having the highest dose potential.

otherwise, monthly.

semimonthly when If there are none, 1 sample from animals are on pasture milking animals in each of three or monthly at other areas between 5 and 8 km (3-5 mi) times.

distant (#51, #54) where doses are calculated to be greater than 1 mrem per year.5 One sample from milking animals at a control location 15 to 30 km (9-18 mi) distant (#53) and in the least prevalent wind direction!e 65 ODCM Rev. 21

TABLE 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Type and Frequency and/or Sample Samples and Sample Locationsa Collection Frequency" of Analysisd Food Products&

2 samples (#47, #52) of 3 types of Monthly during Gamma isotopic broad leaf vegetation (as available) growing season.

analysis.

from locations identified per the criteria of Section 6.2b. of this manual.

1 control sample (#62) of 3 types Monthly during Gamma isotopic of broad leaf vegetation (as growing season.

analysis.

available) grown 15 to 30 km (9-18 mi) distant in the least prevalent wind direction.!

When broad leaf vegetation samples are not available, reports from 4 existing supplemental airborne radioiodine sample locations will be substituted.

66 ODCM Rev. 21

TABLE 6-1 (Continued)

TABLE NOTATION a

The number, media, frequency, and location of sampling may vary from site to site. It is recognized that, at times, it may not be possible or practical to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question. Actual locations (distance and direction) from the site shall be provided in Table 6-4 and Figure 6-1 or Figure 6-2 in the ODCM. Refer to Regulatory Guide 4.1, "Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants."

b Regulatory Guide 4.13 provides guidance for thermoluminescence dosimetry (TLD) systems used for environmental monitoring. 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 may be considered to be one phosphor, and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges should not be used for measuring direct radiation.

c Particulate sample filters shall be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the individual samples.

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 purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the wind direction and distance criteria, other sites that provide valid background data may be substituted.

f Groundwater samples should 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.

g The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

67 ODCM Rev. 21

TABLE 6-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Airborne Particulate Fresh Milk Food Products Analysis Water (pCi/i) or Gas (pCi/m3)

(pCI/I)

(pCI/kg, wet)

H-3 20,000

Mn-54 1,000 Fe-59 400 Co-58 1,000 Co-60 300 Zn-65 300 Zr-Nb-95 400 1-131 2 **

0.9 3

100 Cs-134 30 10 60 1,000 Cs-137 50 20 70 2,000 Ba-La-140 200 300 For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/i may be used.

If no drinking water pathway exists, a reporting level of 20 pCi/l may be used.

68 ODCM Rev. 21

TABLE 6-3 DETECTION CAPABILITIES FOR ENVIRONMENTAL ANALYSISa Lower Limit of Detection (LLD)b Airborne Particulate Fresh Milk Food Products Analysis Water (pCi/i) or Gas (pCIm 3)

(pCi/I)

(pCi/kg, wet)

Gross Beta 4

0.01 H-3 2000*

Mn-54 15 Fe-59 30 Co-58, -60 15 Zn-65 30 Zr-95 30 Nb-95 15 1-131 1**

0.07 1

60 Cs-134 15 0.05 15 60 Cs-137 18 0.06 18 80 Ba-140 60 60 La-140 15 15 NOTE: This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

If no drinking water pathway exists, a value of 3000 pCi/I may be used.

If no drinking water pathway exists, a value of 15 pCi/l may be used.

69 ODCM Rev. 21

Table 6-3 (Continued)

TABLE NOTATION a

Guidance for detection capabilities for thermoluminescent dosimeters used for environmental measurements is given in Regulatory Guide 4.13.

b Table 6-3 indicates acceptable detection capabilities for radioactive materials in environmental samples.

These detection capabilities are tabulated in terms of the lower limits of detection (LLDs). The LLD is defined, for purposes of this guide, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

466Sb E

V

2.22

Y

exp(-XAt)

Where:

LLD is the a priori lower limit of detection as defined above (as pCi per unit mass or volume),

sb 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),

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

X is the radioactive decay constant for the particular radionuclide, and At for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting.

In calculating the LLD for a radionuclide determined by gamma-ray spectrometry the background should include the typical contributions of other radionuclides normally present in the samples (e.g.,

potassium-40 in milk samples). Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as an a pori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

70 ODCM Rev. 21

6.2 Requirement

Land Use Census A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden* of greater than 50 m2 (500 ft2) producing broad leaf vegetation.

Applicability:

At all times.

Action:

a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Section 4.2. 1, identify the new location(s) in the next Annual Radioactive Effluent Release Report, pursuant to Section 7. 1.

b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Section 6. 1, add the new location(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may then be deleted from the monitoring program.

6.2.1 Surveillance Requirements

a. The land use census shall be conducted during the growing season annually using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Section 7.2.

6.2.2 Implementation of the Requirements The above Requirement is implemented by Nuclear Administrative and Technical Manual procedures.

Broad Leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 6-1 shall be followed, including analysis of control samples.

71 71 ODCM Rev. 21

6.3 Requirement

Interlaboratory Comparison Program Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that correspond to samples required by Table 6-1, as applicable.

Applicability:

At all times.

Action:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Section 7.2.

6.3.1 Surveillance Requirements

a. A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the methodology and parameters in this manual shall be included in the Annual Radiological Environmental Operating Report pursuant to Section 7.2.

6.3.2 Implementation of the Requirements PVNGS laboratories or contract laboratories which perform analyses for the Radiological Environmental Monitoring Program (REMP) participate in an Interlaboratory Comparison Program. The participation includes all of the determinations (sample medium-radionuclide combinations) that are included in the monitoring program.

If deviation from specified limits is identified an investigation is made to determine the reason for the deviation and corrective actions are taken as necessary. The results of all analyses made under this program are included in the Annual Radiological Environmental Operating Report.

72 ODCM Rev. 21

TABLE 6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS LOCATION SAMPLE SAMPLE NOTE DESIGNATION SITE TYPE (d)

(a)

LOCATION DESCRIPTION (c) 1 TLD SUP E30 Goodyear 2

TLD SUP ENE24 Scott-Libby School 3

TLD SUP E21 Liberty School 4

TLD SUP E16 Buckeye 4

Air SUP E16 Same as TLD 5

TLD (b)

SP ESE11 Palo Verde School 6

TLD (b)

Control SSE31 APS Gila Bend substation 6A Air (b)

Control SSE13 Old US 80 7

TLD (b)

SP SE7 Old US 80 and Arlington School Rd.

7A Air SUP ESE3 Arlington School 8

TLD (b)

OR SSE4 Southern Pacific Pipeline Rd.

9 TLD (b)

OR S5 Southern Pacific Pipeline Rd.

10 TLD (b)

OR SE5 355th Ave. and Elliot Rd.

11 TLD (b)

OR ESE5 339th Ave. and Dobbins Rd.

12 TLD (b)

OR E5 339th Ave. and Buckeye-Salome Rd.

13 TLD (b)

IR NI N site boundary 14 TLD (b)

IR NNE2 NNE site boundary 14A Air (b)

NNE2 371st Ave. and Buckeye-Salome Rd.

15 TLD (b)

IR NE2 NE site boundary, WRF access road 15 Air (b)

NE2 Same as TLD 16 TLD (b)

IR ENE2 ENE site boundary 17 TLD (b)

IR E2 E site boundary 17A Air SUP E3 351st Ave.

18 TLD (b)

IR ESE2 ESE site boundary 19 TLD (b)

IR SE2 SE site boundary 20 TLD (b)

IR SSE2 SSE site boundary 73 ODCM Rev. 21

TABLE 6.4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS LOCATION SAMPLE SAMPLE NOTE DESIGNATION SITE TYPE (d)

(a)

LOCATION DESCRIPTION (c) 21 TLD (b)

IR S3 S site boundary 21 Air SUP S3 Same as TLD 22 TLD (b)

IR SSW3 SSW site boundary 23 TLD (b)

OR W5 N of Elliot Rd 24 TLD (b)

OR SW4 N of Elliot Rd.

25 TLD (b)

OR WSW5 N of Elliot Rd.

26 TLD (b)

OR SSW4 Duke Property 27 TLD (b)

IR SWi SW site boundary 28 TLD (b)

IR WSW1 WSW site boundary 29 TLD (b)

IR WI W site boundary 29 Air (b)

WI Same as TLD 30 TLD (b)

IR WNW1 WNW site boundary 31 TLD (b)

IR NW1 NW site boundary 32 TLD (b)

IR NNW1 NNW site boundary 33 TLD (b)

OR NW4 S of Buckeye Rd.

34 TLD (b)

OR NNW5 395th Ave. and Van Buren St.

35 TLD (b)

SP NNW8 Tonopah 35 Air SUP NNW8 Same as TLD 36 TLD (b)

OR N5 Wintersburg Rd. and Van Buren St.

37 TLD (b)

OR NNE5 363rd Ave. and Van Buren St.

38 TLD (b)

OR NE5 355th Ave. and Buckeye Rd.

39 TLD (b)

OR ENE5 343rd Ave. N of Broadway Rd.

40 TLD (b)

SP N2 Wimtersburg 40 Air (b)

N2 SameasTLD 41 TLD SUP ESE3 Arlington School 42 TLD (b)

SP N8 Ruth Fisher School 44 TLD (b)

Control ENE35 El Mirage 74 ODCM Rev. 21

I TABLE 6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS LOCATION SAMPLE SAMPLE NOTE DESIGNATION SITE TYPE (d)

(a)

LOCATION DESCRIPTION (c)

SUP Transit 45 TLD Control ONSITE Central lab, lead pig 46 TLD (b)

SP ENE30 Litchfield Park School 46 Water (b)

WD NNW8 Local residence 47 TLD SUP E35 Littleton School 47 Vegetation (b)

ESE4 Local residence 48 TLD SUP E24 Jackrabbit Trail 48 Water (b)

WD SW1 Local residence 49 TLD SUP ENE11 Palo Verde Rd.

49 Water (b)

WD N2 Local residence 50 TLD (b)

OR WNW5 S of Buckeye-Salome Rd.

51 Milk (b)

NE4 Local residence (goats) 52 Vegetation (b)

NNE2 Local residence 53 Milk (b)

Control NE36 Local residence (goats) 54 Milk (b)

NNE4 Local residence (goats)

WD 55 Water SUP SW3 Local residence Ground Water 57 (b)

WG onsite Well 27ddc Ground Water 58 (b)

WG onsite Well 34abb Surface Water 59 (b)

WS onsite Evaporation Pond #1 Surface Water 60 (b)

WS onsite Reservoir 62 Vegetation (b)

Control ENE26 Commercial produce company Surface Water 63 (b)

WS onsite Evaporation Pond #2 75 ODCM Rev. 21

NOTES:

(a) Distance and direction are relative to the Unit 2 containment, rounded to the nearest mile.

(b) These samples fulfill the requirements of the ODCM, Table 6-1.

(c) Refer to Figure 6-1 and Figure 6-2 for relative locations of sample sites.

(d) IR - inner ring OR - outer ring SP - school or population center WS - waterborne surface WG - waterborne ground WD - waterborne drinking SUP -designated supplemental sampling location 76 ODCM Rev. 21

I Graphic Scale In Wals I

\\5 4I

~N ~I I I

I 0

1 U e~

N A

~

-K~NNE V'

Mp w

OWNJ NW/

//E KEY TO MAP Paved Road

, Mlepost Unpaved Road ifPlVedNuea 4WD Road 11 Rou

,1nd~ary o

-4eo --

- GasPipellne

-Iol---Cio-OUPipeline r Thermolumlnescerd

-. Tt-- F-Power Lkie Dosimeters (TLO) i -i--I Ralfroad A Air Sample 4 Ah v Vegetaon Sample w Water Sample 0 SIme ED Sample Sun Palo Verde Nuclear Generating Station Figure 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLE SITES 0-10 MILES 77 ODCM Rev. 21

"1' "If H"

/

4....

E vnwnSL NE PoaAo MONAIN 3

k~I~yI H".r I~

AdI N. PW I I*.**

t -4. --

1-4.

N I

ir 4-suwae I

l*

31k

, C. Ava A"M Ho-

4.

Lke BMWGeoal 6z

ý-l PHo *Nlj(

!A P -

4-I 4-A ~4. A ~

~

A -

10

~dwn ho 00 Rm.CuRd~

ljl Am 6 "WE 35 Mileile 2

30 Miles fto*

PIo

+Oak ES KEY TO MAP Railroad Palo Verde Nuclear Airstrip/Airport Generating Station Boundary Schools Located T TmlumlueceaM (I)Sample Site.

Near Sample Sites Demem (iL) v vegetedie

Municipal Buildings A Air Sample M MAt Palo Verde Nuclear Generating Station Figure 6-2 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLE SITES 0-35 MILES

Figure 6-3 DELETED 79 ODCM Rev. 21

0 0 fisite Dose Calculation Manual Palo Verde Nuclear Generating Station f(site Dose Calculation Manual Palo Verde Nuclear Generating Station Buckeye-Salom.

Road Legend J~tY Centerline, of Containment t*

Property L-FPurchased

~

Exclusion Boundary

,Ste Boundary

~

doerty Purchased Elliot Road Figure 6-4 Stele (smilesi)

(Ward Road)

SITE EXCLUSION AREA BOUNDARY 10 ODCM Rev. 21 80

Offsite Dose Calculation Manual Palo Verde Nuclear GeneratIng Station Offsite Dose Calculation Manual Palo Verde Nuclear Generating Station 00 00 Diesel Gen Building tspmy I-non L

uontro Auxiflary Radwaste Building Building Building Elevation of Exhaust Point Above Grade Plant VentCondenser Vacuum'l 145' Fuel Building 1

Palo Verde Nuclear Generating Station Figure 6-5 GASEOUS EFFLUENTS RELEASE POINTS

Offsite Dose Calculation Manual Palo Verde Nuclear Generating Station Graphic Scale In Miles 0

1 2

3 4'

5' 9

2,Z ffair wfmM Wa t

NDMU"DOLAKMý, -1 9 WM Ig 4,

8C0, WI MP "

r" VIEf I:

1-N o

SE tf

-' -$4 AU LI m

lum

/

il o-=.

,-Ik-

=I.-b I

I 1:%

WJCI:

R STXWN LawESA KEY TO MAP Paved Road Unpaved Road 4WD Road

-- 1 1-oi-Gas Pipeline

-oI--- oI-Oil Pipeline

--. T---

T-Power Une

--I---I Railroad g,

Airstrip fl Palo Verde Nuclear 1i Generating Station i Boundary School

Siren Milepost Palo Verde Nuclear Generating Station Figure 6-6 LOW POPULATION ZONE 0-5 MILES 82 ODCM Rev. 21

7.0 RADIOLOGICAL REPORTS

7.1 Requirement

Annual Radioactive Effluent Release Report Routine Annual Radioactive Effluent Release Reports covering the operation of the units during the previous calendar year shall be submitted in accordance with Technical Specification 5.6.3.

The Annual Radioactive Effluent Release Reports 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 Annual 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 (Figure 6-4) 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 in accordance with the methodology and parameters in the ODCM.

The Annual 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 contributions are given Section 5.0 and Regulatory Guide 1. 109 Rev. 1, October 1977.

The Annual Radioactive Effluent Release Report shall also include information required by the Technical Requirements Manual, Section 5.0.600. 1.

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

- In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

83 83 ODCM Rev. 21

The Annual Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

a.

Container volume,

b.

Total curie quantity (specify whether determined by measurement or estimate),

c.

Principal radionuclides (specify whether determined by measurement or estimate),

d.

Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e.

Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f.

Solidification agent or absorbent (e.g., cement, urea formaldehyde).

The Annual 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.

Changes to the ODCM shall be submitted in the form of a complete, legible copy as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Changes made to the Process Control Program shall be submitted with the Annual Radioactive Effluent Release Report for the period of the report in which any change in the Process Control Program was made.

84 ODCM Rev. 21

7.2 Requirement

Annual Radiological Environmental Operating Report Routine Annual Radiological Environmental Operating Reports covering the operation of the units during the previous calendar year shall be submitted by May 15 of each year in accordance with Technical Specification 5.6.2.

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 Section 6.2.

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 Table 6-4 and Figure 6-1 and Figure 6-2 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 one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Section 6.3; discussion of all deviations from the sampling schedule of Table 6-1; and discussion of all analyses in which the LLD required by Table 6-3 was not achievable.

A single submittal may be made for a multiple unit station.

One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

85 ODCM Rev. 21

APPENDIX A DETERMINATION OF CONTROLLING LOCATION The controlling location is the location of the MEMBER OF THE PUBLIC who receives the highest doses.

The determination of a controlling location for implementation of I1OCFR50 for radioiodines and particulates is known to be a function of:

(1) Isotopic release rates (2) Meteorology (3) Exposure pathway (4) Receptor's age The incorporation of these parameters into Equation 5-2 results in the respective equations at the controlling location. The isotopic release rates are based upon the source terms calculated using the PVNGS Environmental Report, Operating License Stage, Table 3.5-12, without carbon.

All of the locations and exposure pathways, identified in the 1984 Land Use Census, have been evaluated. These include cow milk ingestion, goat milk ingestion, vegetable ingestion, inhalation, and ground plane exposure. An infant is assumed to be present at all milk pathway locations. A child is assumed to be present at all vegetable garden locations. The ground plane exposure pathway is only considered to be present where an infant is not present. Naturally, inhalation is present everywhere an individual is present.

For the determination of the controlling locations, the highest X/Q and DIQ values, based on the 9 year meteorological data base, for the vegetable garden, cow milk, and goat milk pathways, are selected for each unit. The receptor organ doses have been calculated at each of these locations. Based upon these calculations, it is determined that the controlling receptor pathway is a function of unit location. For Unit 1, the controlling receptor is a garden-child pathway; for releases from Unit 2 and Unit 3 the controlling receptor is a cow milk-infant pathway. These determinations are based upon Table 4-16,4-17, or 4-18, which, in turn, is based upon the 1984 Land Use Census. Locations of the nearest residences, gardens and milk animals, as determined in the 1984 Land Use Census, are given in Table 4-16, 4-17, and 4418.

86 86 ODCM Rev. 21

APPENDIX B BASES FOR REQUIREMENTS B-2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, 64 of Appendix A to 10 CFR PART 50.

There are two separate radioactive gaseous effluent monitoring systems: the low range effluent monitors for normal plant radioactive gaseous effluents and the high range effluent monitors for post-accident plant radioactive gaseous effluents. The low range monitors operate at all times until the concentration of radioactivity in the effluent becomes too high during post-accident conditions. The high range monitors only operate when the concentration of radioactivity in the effluent is above the setpoint in the low range monitors.

B-3.1 GASEOUS EFFLUENT - DOSE RATE This requirement provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either at or beyond the SITE BOUNDARY, in excess of the design objectives of Appendix I to 10 CFR part 50. This requirement is provided to ensure that gaseous effluents from all units on the site will be appropriately controlled. It provides operational flexibility for releasing gaseous effluents to satisfy the Section H.A and H.C design objectives of Appendix I to 10 CFR part 50. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually 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, shall be given in the ODCM.

The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the total 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. This requirement does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301(a).

This requirement applies to the release of radioactive materials in gaseous effluents from all reactor units at the site.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLD). Detailed discussion of the LLD and other detection limits can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

87 ODCM Rev. 21

B-3.2 SECONDARY SYSTEM LIQUID WASTE DISCHARGE TO ONSITE EVAPORATION PONDS - CONCENTRATION This requirement is provided to ensure that the annual total effective dose equivalent to individual members of the public from the licensed operation does not exceed the requirements of 10 CFR Part 20, due to the accumulated activity in the evaporation ponds from the secondary system discharges.

Restricting the concentrations of the secondary liquid wastes discharged to the onsite evaporation ponds will restrict the quantity of radioactive material that can accumulate in the ponds. This, in turn, provides assurance that in the event of an uncontrolled release of the pond's contents to an UNRESTRICTED AREA, the resulting total effective dose equivalent to individual members of the public at the nearest exclusion area boundary will not exceed the requirements of 10 CFR Part 20.

This requirement applies to the secondary system liquid waste discharges of radioactive materials from all reactor units to the onsite evaporation ponds.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLD). Detailed discussion of the LLD and other detection limits can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

B-4.1 GASEOUS EFFLUENT - DOSE, Noble Gases This requirement is provided to implement Sections II.B, Ill.A and IV.A of Appendix I, 10 CFR Part 50. This requirement implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The surveillance requirements implement the requirements in Section HI.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This requirement applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

88 ODCM Rev. 21

B-4.2 GASEOUS EFFLUENT - DOSE - Iodine-131, Iodine-133, Tritium, and All Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days This requirement is provided to implement the requirements of Sections II.C, III.A, IV.A of Appendix I, 10 CFR Part 50. This requirement is the guide set forth in Section H.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section II.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases for Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.

This requirement applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

B-4.3 GASEOUS RADWASTE TREATMENT The OPERABILITY of the GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. 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 reasonably achievable." This requirement implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 1).D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections U.B and I.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

This requirement applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

The minimum analysis frequency of 41M (i.e., at least 4 times per month at intervals no greater than 9 days and a minimum of 48 times a year) is used for certain radioactive gaseous waste sampling in Table 3-1. This will eliminate taking double samples when quarterly and weekly samples are required at the same time.

89 ODCM Rev. 21

B-4A SECONDARY SYSTEM LIQUID WASTE DISCHARGE TO ONSITE EVAPORATION PONDS - DOSE This requirement is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. This requirement implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix I be shown by'calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

This requirement applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.

B.5.1 TOTAL DOSE AND DOSE TO PUBLIC ONSITE This requirement is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR 20.1301(d). The requirement specifies the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 norems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 torems. Even if a site was to contain up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units (including outside storage tanks, etc.)

are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, submittal of the Special Report within 30 days with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to other requirements for dose limitation of 10 CFR Part 20, as addressed in Section 3.2 and 3.1 of the ODCM. 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. Demonstration of compliance with the limits of 40 CFR Part 190 or with the design objectives of Appendix I to 10 CFR Part 50 will be considered to demonstrate compliance with the 0.1 rem limit of 10 CFR 20.1301.

90 ODCM Rev. 21

B-6.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

The Radiological Environmental Monitoring Program required by this requirement provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLD). The LLDs required by Table 6-3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a prior (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.

Detailed discussion of the LLD and other detection limits can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

B-6.2 LAND USE CENSUS This requirement is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from 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 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2.

B-6.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

91 ODCM Rev. 21

APPENDIX C DEFINITIONS Note:

The following definitions were derived from the Palo Verde Nuclear Generating Station Technical Specifications. These selected definitions support those portions of the Technical Specifications which were transferred to the ODCM and have been incorporated into the Requirements sections of the ODCM.

Definitions:

The defined terms of this section appear in capitalized type and are applicable throughout the Requirements sections of this ODCM.

ACTION ACTION shall be that part of a requirement which prescribes remedial measures required under designated conditions.

CHANNEL CALIBRATION See the Technical Specification definition.

CHANNELCHECK See the Technical Specification definition.

CHANNEL F1INCTIONAL TEST See the Technical Specification definition.

DOSE EQUIVALENT 1-131 See the Technical Specification definition.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table C-I.

GASEOUS RADWASTE SYSTEM A GASEOUS RADWASTE SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, 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.

92 ODCM Rev. 21

APPENDIX C DEFINITIONS (Continued)

OPERABLE-OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

MODE See the Technical Specification definition.

PROCESS CONTROL PRGRAM The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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

RATED THERMAL POWER See the Technical Specification definition.

SITE BONDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

SO)URCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

93 ODCM Rev. 21

APPENDIX C DEFINITIONS (Continued)

UNETRICTED/AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for the 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 TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any 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 Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement 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.

94 ODCM Rev. 21

NOTATION S

D w

4/M M

Q SA ANNUALLY R

P S/U N.A.

TABLE C-1 FREQUENCY NOTATION FREQUENCY At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

At least once per 7 days.

At least 4 times per month at intervals no greater than 9 days and a minimum of 48 times per year.

At least once per 31 days.

At least once per 92 days.

At least once per 184 days.

At least once per 365 days At least once per 18 months.

Completed prior to each release.

Prior to reactor startup.

Not Applicable.

95 ODCM Rev. 21

APPENDIX D REFERENCES 1

Title 10, Code of Federal Regulations, Part 20, "Standards for Protection Against Radiation."

2 Title 10, Code of Federal Regulations, Part 50, "Domestic Licensing of Production and Utilization Facilities."

3 Title 40, Code of Federal Regulations, Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations."

4 Federal Register, Vol. 58, No. 245, Thursday, December 23, 1993, Notices, pages 68170-68179.

5 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.

6 Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977.

7 Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.

8 Regulatory Guide 4.1, "Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants," Revision 1, April 1975.

9 NUREG-0133, Preparation of Radiological Effluent Technical Specifications For Nuclear Power Plants, Oct. 1978.

10 NUREG 0841, "Final Environmental Statement Related to the Operation of Palo Verde Nuclear Generating Station, Units 1, 2, and 3", Section 5.9.1.4, February, 1982.

11 NUREG-1301, "Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactor", Arpil 1991.

12 Environmental Report Operating License Stage, Palo Verde Nuclear Generating Station, December 1981.

13 PVNGS Updated Final Safety Analysis Report 14 Calculation 13-NC-ZY-252, "Annual Average Dose from Normal Operation Liquid Discharge from the Evaporation Pond", Rev 0.

15 Calculation 13-NC-ZY-253, "Annual Average Dose from Normal Operation Airborne Direct and Sky Shine from the Evaporation Pond", Rev 0.

16 Calculation 13-NC-ZY-254, "Radiation Dose Due to an Evaporation Pond Dike Failure During a Seismic Event", Rev. 0.

96 ODCM Rev. 21

APPENDIX E Changes to the Process Control Program 201 PVNGS ARERR 2006

COMPANY CORRESPONDENCE 218-02207-DJH ID #:

DATE:

March 29, 2007 TO:

John Gaffney Sta. #

7297 Ext. #

8800 FROM:

David J. Heckman Sta.#

7815 Ext. #

5932

SUBJECT:

Documentation of 2006 Process Control Program (PCP) Change Authorization This letter meets the necessary documentation requirements for the October 2006 PCP change (CD-600) and provides information needed to complete the 2006 Annual Radioactive Effluent Release Report. The PCP is currently lacking a formal documentation process.

Procedures 75RP-ORP04, "Radiological Reports," and 76DP-ORP03, "Radwaste Process Control Program," make references to the following:

75DP-ORP04, 3.2.2 - The Radiological Service Department shall ensure the following PCP related information is included in the Annual Radioactive Effluent Release Report:

3.2.2.1 Changes and PRB reviews of changes to the PCP made during the reporting period as required by 76DP-ORP03...

3.2.2.2 Documentation that the changes to the PCP have been reviewed and accepted by the PRB and approved by the Director, Radiation Protection...

3.2.2.3... a documented determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State or other applicable regulations.

76DP-ORP03, 3.7.4 - For changes that are determined to be reportable, forward the PCP revision package to PRB for review and acceptance.

3.7.4.1 After receiving review and acceptance from the PRB, forward the package to the Director, Radiation Protection for approval of the revision.

The constituents of the UPCP revision package" are not defined in 76DP-ORP03 and appropriate formal documents are not provided for PRB or RP Director acceptance and approval. Additionally, the "change" is not provided consisely in a reportable format as elements of the change are contained in both the written PCP (procedure) and in the physical process as described in the PCP Revision Notice.

ACT 2984559 has been generated to develop a permanent resolution to this issue. The following information is provided to meet the procedural documentation requirements only for the 2006 CD-600 change:

218-02207-DJH March 29, 2007 Page 2 of 2 In addressing the documented determination of 75DP-ORP04, 3.2.2.3, 50.59 Screening S-06-0462 (provided as an attachment) makes this statement: =Waste is then transferred to a processing system (in this instance the CD-600) designed to create a final product that meets the PVNGS PCP." In meeting the final product requirements of the PVNGS PCP, the CD-600 product does umaintain the overall conformance of the solidified waste product to existing requirements of Federal, State or other applicable regulations."

9 In addressing the documentation requirement of 75DP-ORP04, 3.2.2.2, minutes of the October 8, 2006 PRB meeting are attached attesting to the acceptance of the CD-600 PCP change by the PRB.

In accordance with the aforementioned requirement and 76DP-ORP03, 3.7.4.1, the following electronic signature verifies that the Director, Radiation Protection did approve the change and participated as a member of the PRB during the presentation of the CD-600 PCP change:

Digitally signed by Gaffney, John Gaffney, Joh01 i,5 G n JDN: cn=Gaffney, John P(Z36459)

Reason: I am approving this P(Z36459) diocument Date: 2007.04.05 16:00:51 -07'00' Director, Radiation Protection

The specific PCP change referred to in 75DP-ORP04, 3.2.2.1, is contained in part in the PCP Revision Notice (included as an attachment) and in Revision 6 to 76DP-ORP03. Specifically, the reportable change to the PCP was the addition of the CD-600 system setup and operating procedures to the PCP. In facilitating LRS processing with the CD-600 (versus the CD-1000), minimum dewatering/drying and processing times were reduced constituting a change in processing parameters that could cause an alteration in the final waste product characteristics.

DJH/cdd Attachment

10 CFR 50.59 SCREENING ScreeningfEvaluation Number:

S-06-0462 Revision:

0 Page 1 of 7 ACTIVITY UNDER REVIEW (DOCUMENT NUMBERIREVISION NUMBER):

76RP-0RW78, R 0, "CD-600 System Setup" 76RP-0RW79, R 1, 'CD-600 System Operations" DESCRIPTION OF PROPOSED ACTIVITY:

Review use of the mobile RWE NUKEM Corporation (RNC) CD-600 Solid Radwaste Processing System at PVNGS along with setup and operating procedures for compliance with the PVNGS PCP and all regulations and quldance concerning the processing and product waste form produced from the processing of wet radioactive waste in the CD-600.

(continue on Response Justification Page) 10 CFR 50.59 SCREENING NO YES

1. Does the proposed activity adversely affect a design function described In the UFSAR?

2. Does the proposed activity adversely affect the method of performing or controlling a design function described In the UFSAR?

3. Does the proposed activity replace or adversely revise an evaluation or method of evaluation described In the UFSAR?

4. Does the proposed activity Involve a test or experiment not described In the UFSAR, where an SSC is used or controlled in a manner that is outside the reference bounds of the design for the SSC or is Inconsistent with analyses or descriptions as provided In the UFSAR?

5. Does the proposed activity require a change to the Technical Specifications?

X X

X Heci Davi' (ZOO I verify that the above screening Is accurate and that I am currently qualified to perform activities as a 10 CFR 50.59 Screener/Reviewer.

Digitally signed by Heckman, M

Digitally signed by M mrph Thomas W(Z01 906)

man DN: CN =Hman, Dan, David J DN: CN = Murphy, 1 d J R

DaMdThomas W

(Z01906)

Reason: I am the author of this Reason: I have reve document.

(01 6)document.

Date: 2006.11.16 07:51:05 -0700' Date: 2006.11.16 07:

urphy, oreas W wed this 57:38 -07,00' SCREENER (Digital Signature)

SCREENER (Digital Signature)

REVIEWER (Digital Signature)

PV-E0006 Ver. 13

$3DP-01.C07

10 CFR 50.59 EVALUATION Screening/Evaluation Number:

S-06-0462 Revision:

0 Page 2 of 7 IACTIVITY UNDER REVIEW (DOCUMENT NUMBERIREVISION NUMBER):

76RP-0RW78, R 0, -CD-600 System Setup" 76RP-0RW79, R 1, "CD-600 System Operations" 10 CFR 50.59 EVALUATION NO YES

1. Does the proposed activity result in more than a minimal Increase In the frequency of occurrence of an accident previously evaluated in the UFSAR?

2. Does the proposed activity result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR?

3. Does the proposed activity result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR?

4. Does the proposed activity result in more than a minimal Increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR?

5.

Does the proposed activity create a possibility for an accident of a different type than any previously evaluated in the UFSAR?

6. Does the proposed activity create a possibility for a malfunction of an SSC Important to safety with a different result than any previously evaluated in the UFSAR?

7. Does the proposed activity result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered?

8. Does the proposed activity result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses?

I verify that the above evaluation is accurate and that I am currently qualified to perform activities as a 10 CFR 50.59 Evaluator/Reviewer.

EVALUATOR (Digital Signature)

REVIEWER (Digital Signature)

PV-E000 Vw. 13 3OIP-OMC07

10 CFR 50.59 SCREENING and EVALUATION RESPONSE JUSTIFICATION FORM Page 3 of 7 SCREENINGIEVALUATION NUMBER:

REVISION:

S-06-04620 ACTIVITY UNDER REVIEW:

76RP-0RW78, R 0, "CD-600 System Setup" 76RP-0RW79, R 1, "CD-600 System Operations" RESPONSE JUSTIFICATION INTRODUCTION:

The RWE NUKEM CD-600 solid radwaste processing system (CD-600) is an upgrade to the portable CD-1000 solid radwaste processing model used at PVNGS. Because of the abandonment-in-place of the original installed Hittman system (per SARCN

3476 in 1994), PVNGS can currently only process radioactive concentrates using temporary, portable technologies. Succinctly, processing through both systems occurs after liquid waste is reduced in volume to the capabilities of the installed LRS system and is collected in the form of concentrates in the Concentrate Monitor Tanks (CMTs). These concentrates are held and recirculated until they can be transferred through the truck connection valve (SRN-VI 11) via the Wet Waste Processing Subsystem (described in section UFSAR 11.4.2) of the Solid Waste Management System. Waste is then transferred to a processing system (in this instance the CD-600) designed to create a final product that meets the PVNGS PCP. As per UFSAR section 11.4.2.3, complete waste processing is lAW 76DP-ORP03, "Radwaste Process Control Program."

Use of the CD-1000 was evaluated in February of 2002 lAW the 10CFR50.59 evaluation process as implemented by site procedure 93DP-OLCO7, "10 CFR 50.59 and 72.48 Screenings and Evaluations." At the time the evaluation was performed, it was determined that the connection of the CD-1000 to PVNGS systems "screened out" of the 10CFR50.59 process by reasons cited in an Applicability Determination (AD) documented in letter 1 15-02371-TSG/MHS (attachment). Specifics of this AD identified that connecting a mobile waste processing system does not constitute a temporary or permanent change to the power production facility as per section 1.2 and the flow chart in Appendix D of 93DP-OLCO7, Revision 5. This Screening coheres to the findings of the 2002 AD as it applies to connecting the CD-1000 to plant systems and extends the AD to include the connection of the CD-600, which is similar to the CD-1000 in design and identical in its interface with plant systems. However, questions have arisen as to whether or not the very presence and use of a portable solid waste processing system at PVNGS:

I. Changes a design function as per Regulatory Guide 1.143, "Design guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water-Cooled Nuclear Power Plants,"

2.

Constitutes an "unreviewed safety question" as per IE Circular 80-18, "10CFRS0.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems,"

3.

Requires a revision to an evaluation described in the PVNGS UFSAR,

4. Produces a final waste form that conforms to the PVNGS Process Control Program (PCP) lAW UFSAR 11.4.2.

Section 1.8 of the UFSAR commits PVNGS to Regulatory Guide 1.143 with exceptions. The CD-600 is constructed by RWE NUKEM to meet the guidance of ANS/ANS-40.37-1993, "Standard for Low-Level Radioactive Waste Processing Systems," and the intent of R.G. 1.143. As a mobile radioactive waste processing system, the CD-600 meets the quality criteria of R.G. 1.143 as part of NUKEM's contractual agreement with PVNGS.

IEC 80-18, "10CFR50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems" was written in August of 1980 and is still active and endorsed by HPPOS-086. IEC 80-18 instructs compliance with elements of I OCFRS0.59 that are not in the current (Oct 4, 1999), amended (Dec 14, 2001) revision of 10CFR50.59. For instance, the term "unreviewed safety question" is no longer used, and documented screenings and applicability determinations were not part of 50.59 in 1980. Functionally, it is apparent that the intent of 80-18 was to reinforce the requirements of 1 OCFR50.59 which were not generally being applied to radioactive waste treatment systems across the industry. It is the position of PVNGS Radiological Engineering that the proper way to implement IEC 80-18 is by strictly adhering to the 50.59 process in its current revision as implemented by procedure 93DP-OCLO7, with the following specifically considered in the evaluation:

The need for PRB review per UFSAR - PRB review is necessary for this action as discussed in answering question 2.

Evaluation against Regulatory Guide 1.143 - In this action, all PVNGS radwaste SSCs are used in accordance with their design functions (note previous discussion on LRS, SRS and Wet Waste Subsystem). As previously discussed, the CD-600 meets the intent of R.G. 1.143.

Evaluation against Regulatory Guide 1.21 and UFSAR, section 11.5 for effluent monitoring and sampling controls.

- The Effluent Monitoring Program is implemented at PVNGS per 74DP-9CY08, "Radiological Monitoring Program."

Operation of the CD-600 is within the Radwaste Building and is therefore bounded by in-plant RMS monitoring. No additional monitoring is necessary, nor is a change necessary to the ODCM. IEC 80-18 also requires that this action be bounded for uncontrolled releases to a small fraction of the 10CFROO guidelines. This bounding is discussed in depth in answering question 1.

PV-EOOOG Vor. 13 93ODP-OLOO7

10 CFR 50.59 SCREENING and EVALUATION RESPONSE JUSTIFICATION FORM Page 4 of 7 SCREENINGIEVALUATION NUMBER:

REVISION:

S-06-0462 0

ACTIVITY UNDER REVIEW:

76RP-0RW78, R 0, -CD-600 System Setup" 76RP-0RW79, R 1, "CD-600 System Operations" RESPONSE JUSTIFICATION The following UFSAR requirements were also considered during this screening:

TRM 5.0.500.4, "Radioactive Effluent Controls Program" promulgates the PVNGS ODCM and effluent release limits to members of the public. As previously discussed, effluents from the CD-600 are through the RW Building ventilation system, which is monitored. No changes to the ODCM are necessary as a result of this action. Catastrophic release is bounded by PVNGS UFSAR, 15.7.2, 15.7.3 and 2.4.13.

Table 9.5-1 of the UFSAR addresses the issue of fire protection in the Radwaste Truck Bay where the CD-600 is operated.

Notably, the area is not designated as a safety-related area as per UFSAR 9B.2.10.2. The only consequence of a fire might be the release of the contents of the evaporator body, which is bounded by PVNGS UFSAR, 15.7.2, 15.7.3 and 2.4.13.

TRM 5.0.500.17, "Process Control Program (PCP)," requires the maintenance of a PCP program. Changes will be initiated to 76DP-ORP03, "Radwaste Process Control Program," section 3.1.1 to add new CD-600 procedures, 76RP-0RW78, "CD-600 System Setup," and 76RP-0RW79, "CD-600 System Operations."

TLCO 3.10.200, "Liquid Holdup Tanks," imposes a 60 Curie limit to all outdoor radwaste tanks that are not surrounded by liners, dikes, or walls. The CD-600 will be operated inside the Radwaste Building and is therefore, not subject to this limit. However, even if the CD-600 were used outside, it would not reach the 60 Curie limit. Using the UFSAR maximum inventories for the CMTs from Table 12.2-5, a total curie content of approximately 866 Curies can be derived. Since the CD-600 can hold only 1/50 of the contents of the CMT, the maximum curie content for the CD-600 would be below 18 Curies.

QUESTIONS:

Question 1) Does the proposed activity adversely affect a design function described In the UFSAR? NO The design bases for the Liquid and Solid waste management systems that might be challenged by this action are described in Sections 11.2.1.C (10CFR50 Appendix I release limits) and 11.4.1.A (SRS process capabilities) of the UFSAR. As per the introduction, the CD-600 is constructed by RWE NUKEM under a Quality Assurance Program that meets ANSI/ANS-40.37-1993 and the intent of R.G. 1.143 and provides the capability to process and package wet waste. Therefore, this activity does not adversely affect the design function of the SRS described in 11.4.1.A.

Stated as part of the design basis in section 11.2.1.C, is the consideration of the effect of LRS design on dose to the public within 50 miles of the site (per 10CFR50 Appendix 1). While the CD-600 is technically not part of the LRS, it is processing liquid radwaste; therefore, the impact of its operation on dose to the public should be considered pursuant to NEI 96-07 and 10CFR50.59.

The CD-600 evaporator chamber processes concentrates in batches received from the CMTs. While the CMTs each hold 5,000 gallons of concentrates, the CD-600 can only process 100 gallons (200 for the CD-1000) at any given time. Each batch is dried and the product drummed in DOT-7A containers prior to receiving another subsequent batch. Since the CD-600 is located in the RW Building and all liquids would be captured by the designed berms and drain system, gaseous release from such an accident is of primary concern. During normal operation the CD-600 vents to the RW Building ventilation system, hence any release of radioactive material from the CD-600 is monitored and quantified by the plant RMS system and station procedures. In the event of a leak, the consequences of a complete breach of the CD-600 could easily be bounded by determining the consequences of a postulated rupture of a CMT (which could contain 50 times more activity than the CD-600). During licensing, PVNGS chose to use a Refueling Water Tank (RWT) rupture as its bounding Liquid Tank Failure in UFSAR sections 15.7.2 and 15.7.3. This was found acceptable to the NRC as noted in Q&A 15A.4 (NRC Question 460.19); however, the NRC did perform an analysis of a rupture in our CMTs per Standard Review Plan (NUREG-0800) SER 15.4.9. The conclusions reached in UFSAR SER 15.4.9 were that a catastrophic failure of the CMT would result in dose to the public <1% of applicable 10CFR20 limits. The analysis in UFSAR 15.7.3.4 demonstrates that a catastrophic failure of the RWT would result in dose to the public <1% of applicable 10CFR100 limits. In both scenarios, dose to the public within 50 miles is maintained within the limits of 10CFR50, Appendix I.

PV-EOD06 Ver. 13 93OP-OLCO?

10 CFR 50.59 SCREENING and EVALUATION RESPONSE JUSTIFICATION FORM Page 5 of 7 SCREENINGIEVALUATION NUMBER:

lREVISION:

S-06-0462L 0~s ACTIVITY UNDER REVIEW:

76RP-0RW78, R 0, "CD-600 System Setup" 76RP-0RW79, R 1, "CD-600 System Operations" RESPONSE JUSTIFICATION Hence, normal operation as well as the consequence of a catastrophic rupture of the CD-600 is bounded by existing analysis and will not adversely affect the design basis cited in UFSAR 11.2.1.

The description of the Low Level Storage Area (LLSA) in section 11.4.2 identifies that, by design, there is 350 square feet of usable floor area in the LLSA. Operation of the CD-600 constitutes an appropriate use of this floor space. Therefore. the presence of the CD-600 does not constitute a condition adverse to the design of the LLSA.

UFSAR ALARA design Table 12. 1-1 identifies the design radiation zones for areas within the plant. Drawing 13-N-RAR-002 shows that the Low Level Storage Area, in which processing and temporary storage occur, is a radiation zone 3 with design dose rates below 2.5 mR/hr. Drums of processed wet waste with dose rates of several hundred mRem/hr have been produced at PVNGS using the CD-1000. Drums of waste are temporarily posted, then shielded by placing them in large shielding containers.

A review of routine area radiation surveys indicates that the general area in the Low Level Storage Area has consistently been maintained below 2 mR/hr and excursions above 2.5 mR/hr have been temporary, with posting and access control JAW the PVNGS RP Program. Therefore. the 2rocessing of waste with the CD-600 does not constitute a condition adverse to the designed radiation zone 3 designation for the LLSA.

Question 2) Does the proposed activity adversely affect the method of performing or controlling a design function described in the UFSAR? NO The methods of performing and controlling the design functions identified in USFAR 11.2.1 and 11.4.1 for the LRS and SRS are contained in the site PCP (see answer to question 1). Because the CD-600 is marginally different than the CD-1000, new procedures have been written for the setup and operation of the unit. Setup of the CD-600 is lAW new procedure 76RP-0RW78, "CD-600 System Setup," and operation is lAW 76RP-0RW79, "CD-600 System Operations."

Procedures for Radioactive Waste Management are identified in UFSAR 13.5.2.2.E. Accordingly, changes to the PCP must be reviewed by the Plant Review Board (PRB) prior to becoming effective. lAW 76DP-ORP03, "Radwaste Process Control Program," step 3.7.1.1 states:

Any change in processing parameters that could cause an alteration In the final waste product characteristics (e.g., changing: minimum dewatering Idrying times or temperatures, processing time or temperature for concentrate evaporation, vendors, or methods for processing liquid waste, etc).

Because the CD-600 will reduce processing times andwill process smaller volume batches, the use of the CD-600 requires PRB approval. PRB approval for the use of the CD-600 is also conservatively necessary per 13.4.2.6 (h), as this action may be perceived to be a "major change" to the solid waste treatment system.

76DP-ORP03, "Radwaste Process Control Program" will require an administrative change to include the CD-600 procedures to section 3.1.1, "PCP Procedures." However, by design, the CD-600 will produce the same final waste product as that produced by the CD-1000. Therefore, this action does not adversely affect the method ofoerformine or controlling the design function of the LRS and the SRS as described in the UFSAR.

Question 3) Does the proposed activity replace or adversely revise an evaluation or method of evaluation described In the UFSAR? NO Evaluations relevant to this action are those concerning the wet waste product waste form (per PCP) and those involved in determining the consequences of an uncontrolled release in the form of a rupture of the CD-600 evaporator.

CD-600 product waste form is per 76DP-ORP03, "Radwaste Process Control Program," lAW NRC Technical Position on Waste Form, Rev 1, January 1991, 10CFR61, 10CFR71 and other pertinent publications. Administrative changes to the PCP resulting from this action are pending as discussed in the answer to question 2.

PV-EOOOBVer. 13 93DP-0LC07

10 CFR 50.59 SCREENING and EVALUATION RESPONSE JUSTIFICATION FORM Page 6 of 7 SCREENINGIEVALUATION NUMBER:

lREVISION:

S-06-0462J 0

ACTIVITY UNDER REVIEW:

76RP-0RW78, R 0, "CD-600 System Setup" 76RP-0RW79, R 1, "CD-600 System Operations" RESPONSE JUSTIFICATION PVNGS used a Refueling Water Tank (RWT) rupture as its bounding evaluation for a Liquid Tank Failure in UFSAR sections 15.7.2 and 15.7.3. The consequence to ground water of this postulated rupture is addressed in UFSAR section 2.4.13. The evaluation for this Liquid Tank Failure is contained in Calculation 13-NC-ZY-202. The resulting analysis in section 15.7.3.4 demonstrates that a gaseous release from catastrophic failure of the RWT would result in a dose to the public of<1% of 10CFRI00 limits. Consequently, this activity is bounded by existing UFSAR analysis and will not require any revision to supporting calculations. Hence, this activity does not replace or adversely revise any evaluation or method of evaluation described in the UFSAR.

Question 4) Does the proposed activity involve a test or experiment not described In the UFSAR, where an SSC Is used or controlled in a manner that Is outside the reference bounds of the design for the SSC or is inconsistent with analyses or descriptions as provided In the UFSAR? NO Initial Radwaste system testing criteria for the installed SSCs are located in UFSAR 14.B.48. The CD-600 is not subject to these test criteria, but must meet ANSI/ANS-40.37 and the relevant provisions of R.G. 1.143. All tests and experiments involved in the proper classification of wet wastes are identified in the PCP and its implementing procedures IAW TRM 5.0.500.17. Tests and experiments involved in the PCP are not specifically delineated in the UFSAR: however, they are consistent with analyses and descriptions within the reference bounds of the radioactive waste systems' design and will not be modified as a result of this action.

Question 5) Does the proposed activity require a change to the Technical Specifications? NO There are no Technical specifications directly associated with connecting or operating a mobile radioactive waste processing unit.

TS 5.7, "High Radiation Area," prescribes the administrative controls for High Radiation Areas at PVNGS. These controls are implemented by the PVNGS RP Program. As discussed in Question 1, the LLSA is subject to excursions in dose rates when processing wet waste with the CD-600. Transitory High Radiation Areas are dealt with by shielding or moving the source to an appropriate storage area and/or by posting and controlling access. No Change is required to TS 5.7.

PV-ECON V.t. 13 93*DP-OLC07

10 CFR 50.59 SCREENING and EVALUATION RESPONSE JUSTIFICATION FORM Page 7 of 7 SCREENINGIEVALUATION NUMBER:

/REVISION:

S-06-0462J 0

ACTIVITY UNDER REVIEW:

76RP-0RW78, R 0, "CD-600 System Setup" 76RP-ORW79, R 1, "CD-600 System Operations" RESPONSE JUSTIFICATION

REFERENCES:

1) UFSAR, Revision 13, List E, dated 08/2005.

2) 76DP-ORP03, "Radwaste Process Control Program," Rev. 5, effective 6/7/05.

3) 93DP-OLCO7, "10 CFR 50.59 and 72.48 Screenings and Evaluations," Revs. 4 (eff. 9/14/01) and 14 (7/i 1/06-active).

4)

PVNGS Letter 115-02371-TSG/MHS, 2/19/02.

5) USNRC IEC 80-18, "10CFR50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems," 8/22/1980.

6) USNRC Regulatory Guide 1.143, "Design guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water-Cooled Nuclear Power Plants," Rev. 0, 1978.

7) ANSI/ANS-40.37-1993, "Standard for Low-Level Radioactive Waste Processing Systems."

8) 74DP-9CY08, "Radiological Monitoring Program," Rev. 15, effective 7/28/06.

9) USNRC IOCFR50, "Domestic Licensing of Production and Utilization Facilities," 1/13/1998.

10) NEI 96-07, "Guidelines for IOCFR50.59 Implementation," Revision 1, Nov. 2000.

11) UFSAR SER 15.4.9, "Liquid Tank Failures," November 1981.

12) USNRC Standard Review Plan (NUREG-0800), Section 15.7.3, "Postulated Radioactive Releases Due to Liquid Containing Tank Failures," Rev. 2, July, 1981.

13) PVNGS Drawing 13-N-RAR-002, Rev. 3, 10/21/1997.

14) 76RP-0RW78, "CD-600 System Setup," effective 08/30/06.

15) 76RP-0RW79, "CD-600 System Operations," effective 08/30/06.

16) USNRC IOCFR61, "Licensing Requirements for Land Disposal of Radioactive Waste," as amended 11/2/01.

17) USNRC IOCFR71, "Packaging and Transportation of Radioactive Material," as amended 8/2/2006.

18) PVNGS Calculation 13-NC-ZY-202, "Storage Water Tank Failure, EAB and LPZ Dose," Rev. 11, April 2000.

19) ANSTIANS-40.37, "Mobile Radioactive Waste Processing Systems," 1993.

PV-E0006 Vet. 13 913DP-OLCO7

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 I.

VERIFICATION OF QUORUM Members:

10/18/06 J.H. Hesser (Chair)

X E.C. Sterling X

M.D. Shea X

L.C. Zell X

J.P. Gaffney X

M.E. Powell X

G.J. Bucci (Alternate)

X D.E. Coxon (Alternate X

The Board assembled at approximately 1:04PM on October 18, 2006. Mr. Hesser verified that a quorum existed (at least five members with no more than two alternates meeting the minimum requirement of five). Mr. Clyde conducted a peer check.

Verification completed for training and qualifications by Mr. Clyde, and the meeting convened at approximately 1:05PM.

10/18/06 Technical Assistant:

M.L. Clyde X

Guests:

E. Flodin X

D. Marks X

D. Vogt X

T. Weber X

R. Rogalski X

S. Koski X

T. Gray X

E. Heckman X

II.

PREVENT EVENTS BRIEFING Mr. Hesser conducted the Prevent Events briefing. Several members noted that they needed 50.59 overview training since it is approaching one year since their designation as member/alternate. Mr. Roehler is reviewing the conduct of past training and has indicated that he will provide the schedule for training.

Ill.

AGENDA Mr. Hesser reviewed the agenda for the August Monthly PRB meeting (Attachment A).

Page 1 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 IV.

MEETING MINUTES Meeting Minutes MMPRB06-017 were approved with abstentions by Messrs. Powell, and Coxon.

Meeting Minutes MMPRB06-020 were approved with abstentions by Messrs. Shea and Bucci.

Meeting Minutes SMPRB06-022 were approved with abstentions by Messrs. Coxon and Sterling.

V.

MONTHLY PRB REPORT Presentation of The PRB Monthly Report for September 2006 (Attachment B) occurred with Mr. Vogt, Mr. Flodin and Mr. Marks providing the details. No safety channels were in trip/bypass for greater than seven days. Unit I shutdown on September 19 for pressurizer heater replacement, Unit 2 shutdown on September 30 for commencement of U2R1 3 and Unit 3 operated at 100% for the month of September. No operator work-arounds required management attention.

Mr. Flodin discussed the reactivity management portion overall designated as white.

There were two level 4 human performance issues and six level 4 equipment issues.

The human performance issue was an overly conservative action required by procedures. The procedures are in revision to review the need for the action when changing from abnormal Blowdown to normal Blowdown. The equipment issues documented three COLSS failures, two of which were not associated with a CRDR. Mr.

Sterling questioned why a CRDR was not initiated. Messrs. Vogt and Flodin said they would ensure CRDRs were initiated.

Mr. Vogt discussed issues in the other safety function areas. There were no new issues for reactivity control.

For the safety function Maintenance of Vital Auxiliaries, Mr. Vogt reported the new issues in the following:

Issuance of an IOD occurred for Unit 2 due to minor oil leakage from the lube oil sight glass for the B EDG. The basis for the IOD was engineering's evaluation of the condition noting that the minor leakage was Insufficient to impact the ability of EDG to perform its specified function for the duration of its mission time.

" Initiation of a functional assessment was completed for the U3 EDG B starting air compressor A due to low oil viscosity with corresponding low flash point. Oil changes are scheduled every six weeks until the compressor is replace in December to provide assurance that the viscosity and flash point remain within specification.

Page 2 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006

" Six lODs were required due to fastener issues on EDGs for all three units. The identification of these issues resulted from concentrated walk downs of all of the diesels and has resulted in approximately sixteen items in each unit. Types of findings are planned for implementation in non-licensed operator training to allow better identification by Operations. Mr. Shea asked the question regarding the aggregate affect of all of the deficiencies. After the Board discussed the issues, an action was assigned to Mr. Andrews to have Steve Payne and other members of the DG system team brief the Board. The action is to address cumulative /

aggregate impact of the degraded conditions that resulted in the large number of lODs. The Board briefing should contain the recovery plan developed by the System Team.

The lOD and resulting POD for the K-1 relay was also discussed by the Board.

This subject needs to be covered in the briefing for the DG system.

For the safety function of Inventory Control, Mr. Vogt reported the new issues in the following:

" Initiation of a functional assessment was completed due to wear products causing a color change in the Unit 2 E charging pump gearbox. Engineering assessed the results of the lubrication laboratory tests and concluded that discoloration was due to wear products. Iron content had increased from 44 ppm in May 2006 to 238 ppm in September 2006. The oil will be changed out at the earliest opportunity but until then, monthly samples will be taken to assess any further degradation. The functional assessment concluded that the equipment remained functional.

" An IOD was initiated for the Unit 1 HPSI outboard pump bearings. Both Unit I pumps presently have light pre-load bearings Installed. This bearing type has no approved MEE. A follow-up POD was approved for the issue. The Board mentioned that this was a nonconforming condition.

" An POD was initiated and approved by the Unit 3 Shift Manager for the inconsistent flow rates for the charging pumps with respect to assumptions made in the safety analysis when instrument uncertainty is applied. The POD will be attached to CRDR 2925248.

For the safety function of Heat Removal, Mr. Vogt reported the new issues in the following:

Several PODs were revised or initiated with respect to the issues related to SP chemistry affecting all three units.

Page 3 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 For the safety function of Containment Temperature and Pressure Control, Mr. Vogt reported the following:

A POD was initiated in Unit 3 for scratches on the stem of SIA-HV684, the containment spray discharge to shutdown cooling heat exchanger discharge.

The IOD had concluded operability was maintained and the POD provided the additional basis support.

For the safety function of Containment Combustible Gas Control, one new issue was identified by the Unit I Control Room staff while performing the LOC 163 review of WM 2924075. The work mechanism identified a loose and intermittent trickle charge indicating light for the B HP recombiner. The System Engineer consulted the Vendor Technical Document and associated internal wiring diagrams to conclude that it has no impact on the recombiner's capability to perform its safety function. The Unit1 Shift Manager concurred.

The Board identified no issues affecting nuclear safety during the review of the Monthly report.

The Board reviewed six LER issued during September 2006. Unit I LERs were both supplements and Unit 2 had one supplement and three new LERs.

LER #2-2006-002-00 involved reporting that the watertight door between the A and B train Auxiliary Feedwater Pump rooms was not secured. There were also no compensatory measures in place. When door C-A-06 is open with the unit operating in Model, compensatory action must be taken in order to maintain both AF trains operable. Based on the Security Computer transactions, the Control Room staff determined that the door was open approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 20 minutes. With this condition, both A and B trains of AF were considered inoperable. Upon further investigation it was concluded that Fired Department personnel had failed to close C-A-06 after leaving the pump room. In the past three years there were two similar events reported.

LER #2-2006-003-00 documented the Unit 2 reactor trip that occurred on July 26, 2006.

The preliminary root cause was implementation of inadequate action plan for corrective maintenance, which did not accommodate for the build up of condensate in the isolated main turbine piping associated with Control Valve #2. This occurred due to a belief that the maintenance evolution was equivalent to control valve testing. A similar event caused by a main turbine upset occurred on June 7, 2004 in Unit 3, when a main turbine EHC system malfunction caused main turbine control and intercept valves to close, resulting in a reactor trip.

LER #1-2005-001-02 provided the direct cause of the Unit I NAN-S06 power failure.

The electrical fault initiated as a C phase to ground fault and transition to a three phase Page 4 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 to ground fault. The fault was terminated when the normal feeder breaker NAN-S06H opened on protective relaying. The actions to prevent recurrence were added in this supplement.

LER #1-2003-001-01 supplemented information regarding the lift pressure verification testing on four pressurizer safety valves (PSV) that were removed for testing during the tenth refueling outage in Unit 1. The testing revealed that the as-found lift pressure for on of the four PSVs was outside of the Technical Specification limits. The as-found PSV condition appears to be the result of a degraded spring. The equipment root cause failure analysis found that the valve spring did not meet the vendor's inspection criteria for squareness. The out of tolerance condition was evaluated and it was determined the results, based on the as-found conditions, were bounded by the peak RCS pressure results of the current Loss of Condenser Vacuum analysis of record. The supplement provided an update to the cause.

LER #2-2006-004-00 reported the failure of FWIV #174 on July 27, 2006 due to a failed four-way valve lodged in the center blocked position. Evaluation concluded that the condition would have prevented fast closure of the FWIV upon receipt of a main steam isolation signal and had existed since approximately 2100 on July 13, 2006. This exceeded the Technical Specification required action time. The four-way valve was replaced and the FWIV was declared operable on July 28, 2006. The preliminary cause was that the failed condition of the four-way valve was not known to operators until the accumulator failed to recharge on July 27, 2006. The cause of the failure of the four-way valve is under investigation.

LER #2-2003-003-01 provided a supplement to identify the direct cause associated with the inoperable Source Range Monitor in Unit 2 during Mode 6 operations. The direct cause was identified as a loose cable connection for channel 2 source range input cable at the Auxiliary Building/Containment Building penetration. The root cause was identified as an incorrect operability decision resulting in a violation of Technical Specification 3.9.2.

One non-reportable event occurred as follows:

CRDR 2928433 documents an improper use of Flashing Lights for posting a LHRA when it was possible to erect a physical barrier. This use of flashing lights versus physical barrier occurred during the Unit I forced outage to replace pressurizer heaters.

The Board noted no nuclear safety issues for any of the LERs.

Page 5 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 VI.

PROPOSED LICENSING CHANGES Mr. Rogalski presented LDCR 06-007, which makes three changes to the Technical Requirements Manual and deletes SR 3.11.101.10. The surveillance requirement is developed from two NFPA standards (24 and 25) that deal with installation of fire protection piping. Mr. Koski informed the Board that the past ten performances of the testing had not debris recovered. Mr. Zell questioned FME possibilities in the future with this testing being deleted. The response was the two standards and the test methodology was for piping installation and major flushing post-installation. Current practices ensure limiting FME. Mr. Hesser asked about retest activities post-maintenance without the surveillance test in place. Flushing could still be invoked without the surveillance requirement in place. Mr. Sterling mentioned that there was no content in the document discussing the chemical content of the water and its potential for corrosion, precipitation or fouling. The evaluation will be updated to address this concern. In response to a system level corrosion question by Mr. Bucci, annual water samples and corrosion coupons are taken and evaluated.

The Board withheld approval pending changes.

Mr. Heckman presented changes proposed to the Process Control Program (PCP) to the Board (Attachment C). PRB review for PCP changes is specified in UFSAR Section 13.4.2.6H. The proposed change consisted of using a different model of Solid Radwaste processing machine, the CD-600. The current PCP has the use of the CD-1000 as the Solid Radwaste processing machine. The 10CFR50.59 Screening addressed any potential impact to the licensing basis and found none. The final waste form is the same for both the CD-1 000 and the CD-600. Required procedures are in place for operation of the proposed unit. The primary reason for changing to the CD-600 is increased reliability of the machine. There were no comments from the Board.

The change to the PCP was approved unanimously.

Mr. Rogalski returned LDCR 06-007 as a revision only addressing TRM LCO 3.11.100.

The controversial section in the discussion above will be brought back for later review.

The only change proposed by this version of the LDCR is allowing the use of the Plant Computer and multipoint recorder with an audible alarm in lieu of requiring an eight-hour tour. The Board approved the revised LDCR unanimously.

VII.

PRB ACTION ITEMS Mr. Hesser announced that the future expectation for completion of action items is one month. Insufficient time remained for the coverage of the actions on the agenda plus the members assigned actions were not in attendance.

Page 6 of 64

Palo Verde Nuclear Generating Station Plant Review Board Monthly Meeting Meeting 06-023 October 18, 2006 VIII. OTHER ITEMS FOR REVIEW The Board decided that IOCFR50.59 training would be a continuing training requirement for members and alternates instead of a one-time performance.

IX.

PROPOSED AGENDA ITEMS for FUTURE PRB MEETINGS No additional items proposed.

X.

MEETING ADJOURNMENT The Chairman accepted a recommendation for adjournment of the meeting. He requested a response by all members on whether there were any nuclear safety issues that had been left unresolved or required further immediate action. With none noted by any of the members, he adjourned the meeting at approximately 3:36PM.

Submitted by: PRB Technical Assistant Digitally signed by Clyde, Miles L Clyde, Mile O:CN Clyd, as L(Z25215)

Reason: I am fth authr o f ths L(Z252Date: 2006.12.22 15:47:14 -0700 Approved by: PRB Chairman

Shea, Michael E (Z02452)

Digitally signed by Shea, Michael D(Z02452)

DN: CN = Shea, Michael

) D(Z02452)

Reason: I am approving this document.

Date: 2006.12.27 08:37:41 -07'00' Page 7 of 64

Radwaste Process Control Program I

76DP-ORP03 Revision 5 Appendix A, Page 1 of 1 (Sample)

PCP Revision Notice Date:

10/14/06 Page 1

of 1 Originator.

David Heckman 5932 I Description of Revision:

Approval of the NUKEM CD-600 radioactive waste processing system for use at PVNGS lAW UFSAR 13.5.2.2.E and 13.4.2.6 (h).

r Revision is NOT reportable - PRB review, R.P. Director approval, and reporting in the annual Radioactive Effluent Release Report are not required.

0l Revision is reportable - Requires PRB review, R.P. Director approval, reporting in the annual Radioactive Effluent Release Report, and a justification for the revision below.

Justification for Revision: (Ensure the following items are addressed)

( UFSAR 13.5.2.2.E)

1.

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and

2.

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

.,ij~tifinAtinn for chAnci, is lAW attanhed 5;0-F,9r9

.ning lAW 76DP-ORP03, 3.7.1.1, A reportable change to the PCP consists of:

...Any change in processing parameters that could cause an alteration in the final waste product characteristics (e.g., changing: minimum dewatering /

drying times or temperatures, processing time or temperature for concentrate evaporation, vendors, or methods for processing liquid waste, etc.).

The CD-600 has a smaller evaporator body and will reduce the process batch volume by 50% with the result being a shorter drying time.

Additionally, Pursuant to UFSAR, 13.4.2.6 (h), PRB approval is required for a major change to a liquid radwaste treatment system. The introduction of the CD-600 radwaste processing system is conservatively viewed as a major change to an liquid radwaste processing system.

, Digitally signed by Fladager. Mark AqZ3SS4S)

Fladager, Mark A D,: ON.,,==.,. Mr Az35,,5)

Reason: I arm approving this dowument Approved By:

(Z35545)

____o____.

Date:

Radiological Services Department Leader Use additional pages as required.

76DP-ORP03, Appendix-A NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 13 of 137]

I Radwaste Process Control Program I

76DP-ORP03 I Revision 6

1 The purpose of this procedure is to describe the Process Control Program (PCP) used at Palo Verde Nuclear Generating Station (PVNGS) to process various radioactive "wet wastes", including resin slurries, evaporator bottoms, and filter cartridges.

This procedure also describes the procedural controls governing revisions to the PCP, delineates criteria used to evaluate the reportability of changes made to the PCP, and describes the reporting requirements.

Procedure Level of Use is Information INUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 1 of 13

Radwaste Process Control Program I 76DP-ORP03 Revision 6 TABLE OF CONTENTS Section Page Number 1.0 PURPOSE and SCOPE 3

1.1 Purpose 3

1.2 Scope 3

2.0 RESPONSIBILITIES 4

2.1 The Vice President, Nuclear Production 4

2.2 The Director, Radiation Protection 4

2.3 Radiological Services Department Leader 4

3.0 PROCESS CONTROL PROGRAM 5

3.1 Description 5

3.2 Precautions and Prerequisites 6

3.3 Process Parameters 7

3.4 Vendors 7

3.5 Waste. Sampling 7

3.6 Stability Requirement 8

3.7 Process Control Program Revisions 8

3.8 Record Retention 9

4.0 DEFINITIONS and ABBREVIATIONS 10 4.1 Definitions 10 4.2 Abbreviations 11

5.0 REFERENCES

11 5.1 Implementing 11 5.2 Developmental 11 6.0 APPENDICES 12 6.1 Appendix A - PCP Revision Notice 13 INUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 2of1

Radwaste Process Control Program j

76DP-ORP03 Revision 6 1.0 PURPOSE and SCOPE 1.1 Purpose (RCTS 032630-01) 1.1.1 This procedure describes the Process Control Program (PCP) at Palo Verde Nuclear Generating Station (PVNGS) for processing radioactive wet waste.

This plant-specific PCP establishes a set of process parameters which provide boundary conditions within which reasonable assurance can be given that the processed waste will contain essentially zero free liquid and have appropriate waste form characteristics. (Branch Technical Position ETSB 11-3)

Technical Requirements Manual, section 5.0.500.17, states, "The purpose of the Process Control Program is to contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste."

1.1.2 Complete waste processing and absence of free liquid prior to shipment is assured by the implementation of a process control program consistent with the recommendations of Branch Technical Position ETSB 11-3. (UFSAR 11.4.2.3.1) 1.1.3 The program will comply with applicable federal and Arizona state regulations.

Implementation of the PCP will be in accordance with applicable portions of the PVNGS Quality Assurance program.

1.1.4 This PCP should be implemented to maintain any potential radiation exposure to plant personnel to "as low as is reasonably achievable" (ALARA) levels, in accordance with 75DP-ORP03, "ALARA Program Overview."

1.2 Scope 1.2.1 This program applies to processing of radioactive wet waste using plant-installed systems, plant portable processing systems, and vendor provided portable processing systems at PVNGS.

1.2.2 The process control program does not apply to radioactive waste that is shipped off site for additional processing prior to disposal.

NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 3 of 13

Radwaste Process Control Program 76DP-ORP03 Revision 6 2.0 RESPONSIBILITIES 2.1 The Vice President, Nuclear Production Ensure the performance of a review by a qualified individual/organization of changes to the Process Control Program (PCP).

2.2 The Director, Radiation Protection 2.2.1 Review and approve any reportable changes to the Process Control Program.

(UFSAR 13.52..2.E) 2.2.2 Ensure that reportable changes to the PCP are forwarded to the Plant Review Board (PRB) for review and acceptance prior to implementation.

(UFSAR 13.4.2.6.1) 2.3 Radiological Services Department Leader (RSDL) 2.3.1 Implement the Radwaste Process Control Program.

2.3.2 Provide an independent review of proposed changes to the Process Control Program.

2.3.3 Make changes to the PCP as necessary to maintain compliance with State and Federal Regulations, Licensing commitments, and burial site requirements.

2.3.4 Report changes to the Process Control Program to the NRC in the Annual Radioactive Effluent Release Report for the period in which they were made.

This submittal should contain: (UFSAR 13.52.21E)

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

" A determination that the change did not reduce the overall conformance of the processed waste product to existing criteria for solid wastes.

2.3.5 Monitoring the activities of vendor personnel to assure vendor compliance with the Process Control Program and the PVNGS Quality Assurance Program.

I NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 4 of 13 1

Radwaste Process Control Program I 76DP-0RP03 Revision 6 2.3.6 Review and approve vendor radioactive waste processing procedures through the PVNGS procedure approval process.

2.3.7 Ensuring personnel under his control are fully aware of, and operate equipment in compliance with, the Process Control Program.

2.3.8 Establish and maintain records documenting the PCP periodic reviews, revision technical reviews, records of the review and evaluation of changes to the PCP, and PCP implementing procedures.

3.0 PROCESS CONTROL PROGRAM 3.1 Description The process control program (PCP) consists of the procedures and processes by which processing and packaging of low-level radioactive wet waste is accomplished and which provide reasonable assurance of compliance with low-level waste requirements. While other procedures may be used in the course of processing and packaging the affected waste types, only those identified in this procedure are considered to be within the scope of the PCP. (CRDR 981853-06) 3.1.1 PCP Procedures:

76DP-ORP03, Radwaste Process Control Program

76RP-0RW05, Packaging and Classification of Radioactive Waste 76RP-ORWO8, High Integrity Container Setup and Closure

" 76RP-ORWO9, Transfer, Storage, and Processing of Radioactive Filters

76RP-ORW12, PVNGS Resin Drying System

" 76RP-0RW78, CD-600 System Setup

76RP-0RW79, CD-600 System Operations Any vendor procedure used in processing or packaging wet waite 3.1.2 PCP Processes

Dehydration (evaporation) or solidification, such as with evaporator concentrates

" Dewatering or drying, such as with bead resin

" Packaging any radioactive material in a High Integrity Container NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 5 of 13

Radwaste Process Control Program I 76DP-ORP03 Revision 6 3.1.3 Wet Waste Types The various wet waste types within the scope of the PCP at PVNGS are:

Evaporator concentrates from a forced recirculation evaporator.

Radioactive bead resin waste.

Radioactive spent filter cartridges.

Radioactive sludge.

Other miscellaneous wet wastes, as determined by R.P. supervision.

3.2 Precautions and Prerequisites 3.2.1 The radiological requirements necessary for implementing the Process Control Program are contained in 75DP-ORPO1, "RP Program Overview " and the other procedures of the RP program. (RCTS 032648-01) 3.2.2 All radioactive wet waste processing will be accomplished in accordance with approved procedures.

3.2.3 The final waste product of all wet waste processing evolutions must be verified by a PVNGS representative. (UEN 87-07) 3.2.4 Waste generators are allowed to stabilize Class B & C waste by placing waste in a High Integrity Container (HIC) provided there is an associated topical report that has been approved by the NRC or for which a Certificate of Compliance, or other State Approval document, has been issued. (EN 89-27) 3.2.5 Waste generators who use polyethylene containers for the disposal of Class B and Class C waste should either: (IEN 89-27)

Place and ship the polyethylene container in an approved HIC, or

obtain assurance and documentation from the disposal site operator that structural stability consistent with Part 61 will be provided at the site.

3.2.6 When packaging wet waste in a HIC, the effects of transportation on the amount of drainable liquid that might be present should be considered. (BT? on Waste Form, Rev.1, 1/91)

NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 6 of 13

Radwaste Process Control Program I 76DP-ORP03 Revision 6 3.3 Process Parameters (CRDR 981853-06) 3.3.1 Proper waste characteristics for Class A wet waste will be assured by adhering to the conditions prescribed in applicable procedure(s).

3.3.2 For Class B and Class C waste, a portable processing system may be used in accordance with approved operating procedures and a 10 CFR 61 Topical Report approved by the NRC, the burial facility, or its regulating agency.

3.3.3 Evaporator Concentrates - complete processing of the waste batch and absence of free liquid is assured by meeting the time and temperature requirements specified in operating procedures for the equipment used to process concentrates for burial.

3.3.4 Ion exchange resins may be dewatered in accordance with the Process Control Program and a 10 CFR 61 Topical Report approved by the NRC, the burial facility, or its regulating agency. The parameters specified in the topical report ensure the final waste product will have appropriate waste form characteristics.

3.3.5 Radioactive spent filters, and other appropriate radioactive material, may be placed in an approved High Integrity Container for disposal, in accordance with approved procedures and the container Certificate of Compliance (C of C).

3.4 Vendors 3.4.1 Vendor operating procedures will undergo the same review and approval process as PVNGS procedures.

3.4.2 Vendor activities will be monitored to assure vendor compliance with the Process Control Program and the PVNGS Quality Assurance Program 3.4.3 If vendor processing is utilized, a PVNGS representative will verify proper processing of the waste product in accordance with the PCP, the vendor's operating procedure, and a 10 CFR 61 Topical Report approved by the NRC, if applicable.

3.5 Waste Sampling 3.5.1 Sampling requirements for the various wet waste streams are contained in 76RP-ORWO3, Waste Stream Sampling and Database Maintenance.

NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 7 of 13

Radwaste Process Control Program I 76DP-ORP03 Revision 6 3.6 Stability Requirement The waste class of radioactive wet waste should be evaluated in accordance with 76RP-ORWO5, "Packaging and Classification of Radioactive Waste," prior to packaging to ensure meeting the stability specifications of 10CFR61.56, "Waste characteristics," and the Branch Technical Position ETSB 11-3 (Revision 2, July 1981).

3.7 Process Control Program Revisions 3.7.1 Reportable change(s) to the Process Control Program consist of the following:

(CRDR 981853-05) 3.7.1.1 Any change in processing parameters that could cause an alteration in the final waste product characteristics (e.g., changing: minimum dewatering/drying times or temperatures, processing time or temperature for concentrate evaporation, vendors, or methods for processing liquid waste, etc.);

3.7.1.2 Any change to the purpose, scope, or intent of the PCP; 3.7.1.3 Any change to the PCP that might cause inconsistencies with the NRC Waste Form Technical Position Paper (Rev. 1, January 1991),

or Branch Technical Position - Effluent Treatment Systems Branch 11-3, section II (Rev. 2, July 1981).

3.7.2 The Radiological Services Department Leader will review proposed changes to Process Control Program processes, including the determination of reportability.

3.7.2.1 If the proposed revision does not meet any of the criteria in step 3.7.1, then the change is not reportable.

That determination will be documented in one of two ways. Either by:

" placing a brief description on Appendix A, PCP Revision Notice, marking the appropriate box, and a signature by the RSDL, or

" annotating in the procedure change record that the change is not reportable. The RSDL signature for approval of the procedure revision will indicate his concurrence that the revision is not reportable.

NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 8 of 137]

Radwaste Process Control Program I 76DP-ORP03 Revision 6 3.7.2.2 If the proposed revision is reportable, then Appendix A will be completed, with the appropriate box marked, and it will be signed by the RSDL to indicate his approval.

3.7.3 Document the evaluation of reportable changes to any affected process on Appendix A, "PCP Revision Notice," and attach to the change document (e.g.,

procedure, 50.59, etc.). Documentation for reportable changes to the PCP shall include the following:

3.7.3.1 Sufficient details to totally support the rationale for the change; 3.7.3.2 A determination that the change did not reduce the conformance of the final waste product to existing criteria for waste disposal.

3.7.4 For changes that are determined to be reportable, forward the Process Control Program revision package to PRB for review and acceptance. (MFAR 13.42.6.h) 3.7.4.1 After receiving review and acceptance from the PRB, forward the package to the Director, Radiation Protection for approval of the revision. (UFSAR 13.5.2.2.E) 3.7.5 Copies of the following documents should be maintained on file by Radiological Engineering:

Appendix A, "PCP Revision Notice" Cross-Discipline Reviews, if applicable Any associated 10 CFR 50.59 reviews and evaluations 3.8 Record Retention 3.8.1 Records of reviews performed for changes made to the PCP shall be retained for the duration of the operating license, or the requirements of the insurer, whichever is greater. (UFSAR 17.2.6.4.1 (A)( 14))

3.8.2 Turnover applicable records to NIRM in accordance with the appropriate turnover instructions.

INUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 9 of 13 1

Radwaste Process Control Program I 76DP-ORP03 Revision 6 4.0 DEFINITIONS and ABBREVIATIONS 4.1 Definitions 4.1.1 Approved Hfigh Integrity Container - A container used to provide the long-term stability requirement of 10 CFR 61. Approval is verified by reviewing a copy of the "Certificate of Compliance" prior to the container's use and naamtaining the C of C on file during and subsequent to the container's use.

4.1.2 Batch - An isolated quantity of waste feed to be processed having essentially constant physical and chemical characteristics.

4.1.3 Certificate of Compliance (C of C) - for containers, an approval document, normally issued by the burial state licensing authority, which approves the listed container for use as a burial container.

It also prescribes handling requirements and conditions for use.

4.1.4 Low Level Radioactive Waste (LLW) 4.1.4.1 Those low-level radioactive wastes containing source, special nuclear, or by-product material that are acceptable for disposal in a near surface land disposal facility.

4.1.4.2 Radioactive waste that contains no hazardous materials as defined in RCRA.

4.1.4.3 Radioactive waste not classified as high-level radioactive waste, transuranic waste or spent nuclear fuel.

4.1.5 Process Control Program (PCP) - A program that provides assurance that the methods used for processing wet low-level radioactive waste will result in a waste form that is acceptable for disposal at a licensed land disposal fatcility in accordance with 10 CFR 61 requirements.

4.1.6 Reportable - For the purposes of this procedure, means that the change meets listed requirements; therefore, PRB review is required and reporting is mandated in the Annual Radioactive Effluent Release Report.

4.1.7 Stability - As used in this document, "stability" means structural stability. A structurally stable waste form will generally maintain its physical dimensions and its form under expected disposal conditions. Stability can be provided by the waste form itself, processing the waste into a stable waste form, or placing the waste into a disposal container or structure that provides stability.

4.1.8 Waste Form - usually refers to the stability of processed waste: stable waste form or unstable waste form. Also applied to physical state of waste (i.e.,

liquid, solid, gas).

INUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 10:of:1=3

Radwaste Process Control Program 76DP-ORP03 Revision 6 4.2 Abbreviations 4.2.1 ALARA - As Low As Reasonably Achievable 4.2.2 BTP - Branch Technical Position 4.2.3 MIC - High Integrity Container 4.2.4 PCP - Process Control Program 4.2.5 PRB -Plant Review Board 4.2.6 RCRA - Resource Conservation and Recovery Act 4.2.7 RSDL - Radiological Services Department Leader

5.0 REFERENCES

5.1 Implementing 5.1.1 75DP-ORPO1, RP Program Overview 5.1.2 75DP-0RP03, ALARA Program Overview (RCTS 032633-01) 5.1.3 76RP-ORWO5, Packaging and Classification of Radioactive Waste 5.1.4 76RP-ORWO3, Waste Stream Sampling and Database Maintenance 5.1.5 Palo Verde Nuclear Generating Station Technical Requirements Manual Section 5.0.500.17.

5.1.6 Palo Verde Nuclear Generating Station updated Final Safety Analysis Report, Sections 11.4, 12.1, 12.3, 13.4, 13.5, and 17.2. (RCTS 032632-01) 5.1.7 USNRC Branch Technical Position ETSB 11-3, Rev 2, July 1981 "Design Guidance for Solid Radioactive Waste Management Systems Installed in Light Water Cooled Nuclear Power Reactor Plants." (RCTS 032636-01) 5.1.8 NRC Technical Position on Waste Form, Rev 1, January 1991.

5.1.9 NRC Information Notice 89-27, "Limitations on the Use of Waste Forms and High Integrity Containers for the Disposal of Low-Level Radioactive Waste,"

March 8, 1989.

5.1.10 NRC Information Notice 87-07, 2/3/87, "Quality Control of Onsite Dewatering

/ Solidification Operations by Outside Contractors."

5.2 Developmental 5.2.1 93DP-OLCO7, 10 CFR 50.59 and 72.48 Screenings and Evaluations 5.2.2 10 CFR 20, "Standards for Protection Against Radiation" I NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 11 of 13

Radwaste Process Control Program I

76DP-ORP03 Revision 6 5.2.3 10 CFR 61, "Licensing Requirements for Land Disposal of Radioactive Waste" (RCTS 032637-01) 5.2.4 10 CFR 71, "Packaging and Transportation of Radioactive Material" 5.2.5 49 CFR Subchapter C, "Hazardous Materials Regulations" 5.2.6 "Quality Assurance During The Operations Phase," UFSAR 17.2 (RCTS 032634-01) 5.2.7 NUREG-0472, Rev 2, July 1979, "Radiological Effluent Technical Specification for PWRs" 5.2.8 NUREG-1301, Rev.0, 4/91, "Offsite Dose Calculation Manual Guidance:

Standard Radiological Effluent Control for Pressurized Water Reactors" 5.2.9 NUREG 0800, Rev 2, July 1981 U.S. NRC Standard Review Plan 11.4, "Solid Waste Management Systems," (RCTS 032635-01) 5.2.10 Reg Guide 1.2.1 - Measuring, Evaluating and Reporting Radioactivity in Solid Waste and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Nuclear Power Plants, Rev. 1.

5.2.11 IE Circular 80-18, 10 CFR 50.59, Radioactive Waste Treatment Systems 5.2.12 Commitment Action Tracking System Safety Evaluation for Changes to Partition RCTS RCTS RCTS RCTS RCTS RCTS RCTS CRDR CRDR Commitment Number 032632 032633 032634 032635 032636 032637 032648 981853 981853 Action Number 01 01 01 01 01 01 01 05 06 Procedure Section 5.1.6 5.1.2 5.2.6 5.2.9 5.1.7 5.2.3 3.2.1 3.7.1 3.1,3.3 6.0 APPENDICES Appendix A - PCP Revision Notice I NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL I

Page 12 of 13

Radwaste Process Control Program I

76DP-ORP03 I Revision 6 Appendix A, Page 1 of 1 (Sample)

PCP Revision Notice Date:

Page of Originator:

Ext.:

Description of Revision:

E]

Revision is NOT reportable - PRB review, R.P. Director approval, and reporting in the annual Radioactive Effluent Release Report are not required.

['j Revision is reportable - Requires PRB review, R.P. Director approval, reporting in the annual Radioactive Effluent Release Report, and a justification for the revision below.

Justification for Revision: (Ensure the following items are addressed)

( UFSAR 13.5.2.2.E)

1.

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and

2.

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

Approved By:

Date:

Radiological Services Department Leader Use additional pages as required.

76DP-ORP03, Appendix-A NUCLEAR ADMINISTRATIVE AND TECHNICAL MANUAL Page 13 of 13

ELECTRONIC PROCEDURE CHANGE RECORD o

PROCEDURE NO:

76DP-ORP03 OTITLE Radwaste Process Control Program

@ REVISION NO:

6 CATEGORY 1[3 20 30(1 I

@PROCEDURE ACTION: REVISION NEWi]

SUPERSEDE J CANCELf E

EXPEDITED? YES J NO[

OMRL UPDATE? YESj NOO (1) FULL BASIS CHECK? YES NOJ 0i LEVEL OF USE IINFORMATION (1 DESCRIPTION OF CHANGE I CD-600 Setup and Operating procedures added to the list of "PCP Procedures" In step 3.1.1. These procedures would have been considered part of the PCP as vendor procedures as they are "...used In processing or packaging wet waste.'

TEXT DOES NOT AUTOMATICALLY ROLL TO CONTINUATION SHEET R'

DESCRIPTION-CONTINUATION YES

50.59172.48 DOC NUMBER:

S-06-0462 63)

A 1 OCFR50.59 Screening was performed lAW 93DP-OLCO7, rev.14. It was determined that the use of the CD-600 does not require a full Evaluation per step 4.10.3.

RMC review by Mark Fladager.

Applicability Determination performed by David J. Heckman TEXT DOES NOT AUTOMATICALLY ROLL TO CONTINUATIO N SHEET 1

APPLICABILITY - CONTINUATION YES EFFECTIVE DATE REQUIRED PROCEDURE CHANGE RECORD PACKAGE CONTENTS FOR PROCESSING

! (3 NADREQUIRED?

YES[3 NOE)

NAD PAGE COUNT:

EFFECTIVE DATE EPCR, OTHER DOCUMENTS, etc.................. PAGE COUNT:

1 10/19/2006 PROCEDURE PAGE COUNT:

13 EFFECTIVE TIME [OPTIONAL]

Qý TOTAL PAGE COUNT:

14

()

APPROVALS Heckman, David J(Z00977)

Digitally signed by Heckman, David J (Z00977)

DN: CN = Heckman, David J(Z00977)

Reason: I am the author of this document.

Date: 2006.10.17 1621:56 -0700

(@ PREPARER - SIGNATURE DENOTES THAT DOCUMENT IS READY FOR REVIEW AND APPROVAL Digitally signed by Bungard, James P Bungard, James (Z, 8012 DN: CN -Bungard, James P(Z18012)

Reason: I have reviewed this P(Z1 8012) document.

Date: 2006.10.17 16:48:39-0700' REVIEWER - SIGNATURE DENOTES REVIEW COMPLETION AND 25 QUALIFIED IN SWMS AS PROCEDURE TECHNICAL REVIEWER Digitally signed by Gray, Thomas S Gray, Thomas Raso DN: CN = Gray, Thomas S(Z99610)

S (9960)*.o~a*Reason:,I am approvng this S(Z9961 0) document.

Date: 2008.1 0.17 117:110:111

-0700'

(@3 NAID REVIEWER [IF REQUIRED]

OWNERIDESIGNEE - DIGITAL SIGNATURE SECURES DOCUMENT FOR TRANSMITTAL AND USE

ML071220307

Contents

Text

1.0 INTRODUCTION

2.1 Requirements

3.1 Requirements

3.2 Requirements

4.1 Requirements

4.2 Requirement

4.3 Requirements

4.4 Requirements

5.1 Requirement

6.1 Requirement

6.2 Requirement

6.3 Requirement

7.1 Requirement

7.2 Requirement

1.0 INTRODUCTION

2.1 Requirements

3.1 Requirements

Reference:

Reference:

3.2 Requirements

4.1 Requirements

4.2 Requirement

4.3 Requirements

References:

References:

4.4 Requirements

5.1 Requirement

6.1 Requirement

6.2 Requirement

6.3 Requirement

7.1 Requirement

7.2 Requirement

SUBJECT:

REFERENCES:

5.0 REFERENCES

5.0 REFERENCES

Navigation menu

ML071220307

Text

1.0 INTRODUCTION

2.1 Requirements

3.1 Requirements

3.2 Requirements

4.1 Requirements

4.2 Requirement

4.3 Requirements

4.4 Requirements

5.1 Requirement

6.1 Requirement

6.2 Requirement

6.3 Requirement

7.1 Requirement

7.2 Requirement

1.0 INTRODUCTION

2.1 Requirements

3.1 Requirements

Reference:

Reference:

3.2 Requirements

4.1 Requirements

4.2 Requirement

4.3 Requirements

References:

References:

4.4 Requirements

5.1 Requirement

6.1 Requirement

6.2 Requirement

6.3 Requirement

7.1 Requirement

7.2 Requirement

SUBJECT:

REFERENCES:

5.0 REFERENCES

5.0 REFERENCES

Navigation menu

Search