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Proposed Tech Specs Re ODCM
	ML18151A257
Person / Time
Site:	Surry
Issue date:	06/29/1990
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18151A258	List: ML18151A257; ML18153C281;
References
	NUDOCS 9007060136
	Download: ML18151A257 (205)
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{{#Wiki_filter:Attachment 1 Proposed Technical Specification Changes

Surry Units 1 and 2 Virginia Electric and Power Company

TSi TEQHNIQAL SPEQIFIQATIQNS TABLE QF QQNTENTS SEQTION TITLE PAGE 1.0 DEFINITIONS rs 1.0-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TS2.1-1 2.1 SAFETY LIMIT, REACTOR CORE TS2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE TS 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE TS 2.3-1 INSTRUMENTATION 3.0 LIMITING QONDITIONS FOR OPERATION TS 3.0-1 3.1 REACTOR COOLANT SYSTEM TS 3.1-1 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM TS 3.2-1 3.3 SAFETY INJECTION SYSTEM TS 3.3-1 3.4 SPRAY SYSTEMS TS 3.4-1 3.5 RESIDUAL HEAT REMOVAL SYSTEM TS 3.5-1 3.6 TURBINE CYCLE TS 3.6-1 3.7 INSTRUMENTATION SYSTEM TS 3.7-1 3.8 CONTAINMENT TS 3.8-1 3.9 STATION SERVICE SYSTEMS TS 3.9-1 . 3.10 REFUELING TS 3.10-1 3.11 RADIOACTIVE GAS STORAGE TS 3.11-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS TS 3.12-1 3.13 COMPONENT COOLING SYSTEM TS 3.13-1 3.14 CIRCULATING AND SERVICE WATER SYSTEMS TS 3.14-1

TS ii TECHNICAL SPECIFICATION TABLE OF CONTENTS SECTION TITLE PAGE 3.15 CONTAINMENT VACUUM SYSTEM TS 3.15-1 3.16 EMERGENCY POWER SYSTEM TS 3.16-1 3.17 LOOP STOP VALVE OPERATION TS 3.17-1 3.18 MOVABLE INCORE INSTRUMENTATION TS 3.18-1 3.19 MAIN CONTROL ROOM BOITLED AIR SYSTEM TS 3.19-1 3.20 SHOCK SUPPRESSORS (SNUBBERS) TS 3.20-1 3.21 FIRE PROTECTION FEATURES TS 3.21-1 3.22 AUXILIARY VENTILATION EXHAUST FILTER TRAINS TS 3.22-1 3.23 CONTROL AND RELAY ROOM VENTILATION SUPPLY FILTER TRAINS TS 3.23-1 4.0 SURVEILLANCE REQUIREMENTS TS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW TS 4.1-1 4.2 AUGMENTED INSPECTIONS TS 4.2-1 4.3 ASME CODE CLASS 1, 2, AND 3 SYSTEM PRESSURE TESTS TS 4.3-1 4.4 CONTAINMENT TESTS TS 4.4-1 4.5 SPRAY SYSTEMS TESTS TS 4.5-1 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING TS 4.6-1 4.7 MAIN STEAM LINE TRIP VALVE TS 4.7-1 4.8 AUXILIARY FEEDWATER SYSTEM TS 4.8-1 4.9 RADIOACTIVE GAS STORAGE MONITORING SYSTEM TS 4.9-1 4.10 REACTIVITY ANOMALIES TS 4.10-1 4.11 SAFETY INJECTION SYSTEM TESTS TS4.11-1 4.12 VENTILATION FILTER TESTS TS 4.12-1 4.13 DELETED 4.14 DELETED

TSiii

- SECTION TECHNICAL SPECIFICATION TABLE OF CONTENTS TITLE PAGE 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY TS4.15-1 LINES OUTSIDE OF CONTAINMENT 4.16 LEAKAGE TESTING OF MISCELLANEOUS RADIOACTIVE MATERIALS TS 4.16-1 SOURCES 4.17 SHOCK SUPPRESSORS (SNUBBERS) TS 4.17-1 4.18 FIRE DETECTION AND PROTECTION SYSTEM SURVEILLANCE TS 4.18-1 4.19 STEAM GENERATOR INSERVICE INSPECTION TS 4.19-1 4.20 CONTROL ROOM AIR FILTRATION SYSTEM TS 4.20-1 5.0 DESIGN FEATURES TS5.1-1 5.1 SITE TS 5.1-1 5.2 CONTAINMENT TS 5.2-1 5.3 REACTOR TS 5.3-1 5.4 FUEL STORAGE TS 5.4-1 6.0 ADMINISTRATIVE CONTROLS TS 6.1-1 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW TS 6.1-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS TS 6.2-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED TS 6.3-1 6.4 UNIT OPERATING PROCEDURES TS 6.4-1 6.5 STATION OPERATING RECORDS TS 6.5-1 6.6 STATION REPORTING REQUIREMENTS TS 6.6-1

6. 7 ENVIRONMENTAL QUALIFICATIONS TS 6.7-1 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE TS 6.8-1 CALCULATION MANUAL

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TS 1.0-7 K. Low Power Physics Tests Low power physics tests conducted below 5% of rated power which measure fundamental characteristics of the core and related instrumentation. L. Fire Suppression Water System A Fire Suppression Water Systems shall consist of: a water source(s); gravity tank(s) or pump(s); and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser. M. Offsite Dose Calculation Manual The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.6.B.2 and 6.6.B.3. N. Dose Equivalent 1-131 The dose equivalent 1-131 shall be that concentration of 1-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table Ill of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" or in NRC Regulatory Guide 1.109, Revision 1, October 1977.

TS 1.0-8

0. Gaseous Radwaste Treatment System A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

P. Process Control Program (PCP) The process control program shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the 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 and other requirements governing the disposal of the waste . 1

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

TS 1.0-9 R. Ventilation Exhaust Treatment System A ventilation exhaust treatment system is 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. S. Venting Venting is 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. T. Site Boundary The site boundary shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee .

TS 1.0-10 U. Unrestricted Area An unrestricted area shall be any area at or beyond the site boundary where access is not controlled by the licensee for purpose 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. V. Member (s) of the Public Member(s) of the public shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the license who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals/)from exposure to radiation and radioactive materials.

TS 3.7-2 C. In the event of subsystem instrumentation channel failure permitted by Specification 3.7.82, Tables 3.7-2 and 3.7-3 need not be observed during the short period of time an operable subsystem channel is tested where the failed channel must be blocked to prevent unnecessary reactor trip. D. The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4. E. The explosive gas monitoring instrumentation channels shown in Table 3.7-5(a) shall be operable with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.A.1 are not exceeded.

1. With an explosive gas monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, declare the channel inoperable and take the action shown in Table 3.7.5(a).

2. With less than the minimum number of explosive gas monitoring instrumentation channels operable, take the action shown in Table

3. 7-5(a). Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission (Region II) to explain why this inoperability was not corrected in a timely manner .

TS 3.7-8

4. The steam line high differential pressure limit is set well below the differential pressure expected in the event of a large steam line break accident as shown in the safety analysis. (3)

5. The high steam line flow differential pressure setpoint is constant at 40% full flow between no load and 20% load and increasing linearly to 110% of full flow at full load in order to protect against large steam line break accidents. The coincident low T avg setting limit for SIS and steam line isolation initiation is set below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide protection in the event of a large steam line break. (3)

Accident Monitoring Instrumentation The operability of the accident monitoring instrumentation in Table 3.7-6 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. On the pressurizer PORV's, the pertinent channels consist of limit switch indication and acoustic

TS 3.7-9

monitor indication. The pressurizer safety valves utilize an acoustic monitor channel and a downstream high temperature indication channel. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975, and NUREG-0578, "TMl-2 Lessons Learned Task Force Status Report and Short Term Recommendations." Potential accident effluent release paths are equipped with radiation monitors to detect and measure concentrations of noble gas fission products in plant gaseous effluents during and following an accident.

The effluent release paths monitored are the Process Vent Stack, Ventilation Vent Stack, Main Steam Safety Valve and Atmospheric Dump Valve discharge and the Auxiliary Feedwater Pump Turbine Exhaust. These monitors meet the requirements of NUREG 0737.

TS 3.7-9a Instrumentation is provided for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 or Appendix A to 10 CFR Part 50. Containment Hydrogen Analyzers Continuous indication of hydrogen concentration in the containment atmosphere is provided in the control room over the range of O to 10 percent hydrogen concentration. These redundant, qualified hydrogen analyzers are shared by Units 1 and 2 with the capability of measuring containment hydrogen concentration for the range of O to 1O percent and the installation of instrumentation to indicate and record this measurement. A transfer switch with control circuitry is provided for the capability of Unit 1 to utilize both analyzers or for Unit 2 to utilize both analyzers . Each unit's hydrogen analyzer will receive a transferable power supply from Unit 1 and Unit 2. This will ensure redundancy for each unit. Indication of Unit 1 and Unit 2 hydrogen concentration is provided on Unit 1 PAMC panel and Unit 2 PAMC panel. Hydrogen concentration is also recorded on qualified recorders. In addition, each hydrogen analyzer is provided with an alarm for trouble/high hydrogen content. These alarms are located in the

TS 3.7-9c References (1) FSAR - Section 7.5 (2) FSAR - Section 14.5 (3) FSAR - Section 14.3.2

TABLE 3.7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION MONITORING ALARM SETPOINT MONITOR CHANNEL AT ALARM CONDITIONS REQUIREMENTS µCl/cc 1 . Component cooling water radiation Shuts surge tank vent valve See Specification Twice Background monitors HCV-CC-100 3.13

2. Containment particulate and gas Trips affected unit's purge supply See Specification Particulate :,; 9 x 1o-9 monitors (RM-RMS-159 & fans, closes affected unit's purge 3.10 Gas:,; 1 x 10-5 RM-RMS-160, RM-RMS-259 & air butterfly valves (MOV-VS-1 OOA, RM-RMS-260) B, C & D or MOV-VS-200A, B, C & D)

3. Manipulator crane area monitors Trips affected unit's purge supply See Specification :,; 50 mrem/hr (RM-RMS-162 & RM-RMS-262) fans, closes affected unit's purge 3.10 air butterfly valves (MOV-VS-100A, B, C & D or MOV-VS-200A, B, C & D)

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TABLE 3.7-S(a) EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1 . Waste Gas Holdup System Explosive Gas Monitoring System (a) Hydrogen Monitor 1 (b) Oxygen Monitor 1 ACTION 1 - With the number of channels operable less than required by the minimum channels operable requirement, operation of this waste gas hold up system may continue provided grab samples are collected at least once per 24 hours and analyzed within the following 4 hours. (/} w

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TS 3.11-1 3.11 RADIOACTIVE GAS STORAGE Applicability Applies to the storage of radioactive gases. Objective To establish conditions by which gaseous waste containing radioactive materials may be stored. Specification A. Exglosive Gas Mixture

1. The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

a. With the concentration of oxygen in the waste gas holdup

b.

system greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours. With the concentration of oxygen in the waste gas holdup system greater than 4% by volume, immediately suspend all additions of waste gases to the affected tank and reduce the concentration of oxygen to less than or equal to 4% by volume, then take the above action.

2. The requirements of Specification 3.0.1 are not applicable.

B. Gas Storage Tanks

1. The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 24,600 curies of noble gases (considered as Xe-133).

2. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all addition of radioactive material to the tank and within 48 hours reduce the tank contents to within the limits.

3. The requirements of Specification 3.0.1 are not applicable.

TS 3.11-2 Explosive Gas Mixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. Gas Storage Tanks The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity which provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event of 2 hours. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.

TABLE 4.1-1 A EXPLOSIVE MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL CHANNEL DESCRIPTION CHECK CALIBRATION FUNCTIONAL TEST

1. Waste Gas Holdup System Explosive Gas Monitoring System (a) Hydrogen Monitor D Q ( 1) M (b) Oxygen Monitor D Q ( 2) M (1) The channel calibration shall include the use of standard gas samples containing a nominal:

1. one volume percent hydrogen, balance nitrogen, and

2. four volume percent hydrogen, balance nitrogen.

(2) The channel calibration shall include the use of standard gas samples containing a nominal:

1. one volume percent oxygen, balance nitrogen, and

2. four volume percent oxygen, balance nitrogen.

D - Daily M - Monthly Q - Quarterly

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TS 4.9-1 4.9 RADIOACTIVE GAS STORAGE MONITORING SYSTEM Applicability Applies to the periodic monitoring of radioactive gas storage. Objective To ascertain that waste gas is stored in accordance with Specification 3.11. Specification A. The concentration of hydrogen or oxygen in the waste gas holdup system shall be determined to be within the limits of Specification 3.11.A by continuously monitoring the waste gases in the waste gas holdup system with the hydrogen or oxygen monitors required operable by Table 3.7-5(a) of Specification 3.7.E. B. The quantity of radioactive material contained in each gas storage tank shall be determined to be within the limits of Specification 3.11.B at least once per month when the specific activity of the primary reactor coolant is

2200 µCi/gm dose equivalent Xe-133. Under the conditions which result in a specific activity >2200 µCi/gm dose equivalent Xe-133, the Waste Gas Decay Tanks shall be sampled once per day.

TS 6.4-8 N. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,

TS 6.4-9

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE

        , BOUNDARY conforming to the doses associated with 1o CFR Part 20, Appendix B, Table II, Column 1,

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

L_

TS 6.4-10

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

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and .that modifications to the monitoring program are made if required by the results of this* census, and

3) Participation in a lnterlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

TS 6.5-3

9. Records of the service lives of all hydraulic and mechanical snubbers on safety-related systems, including the data at which the service life commences and associated installation and maintenance records.

1o. Records of the annu~I audit of the Station Emergency Plan and implementing procedures.

11. Records of the annual audit of the Station Security Plan and implementing procedures.

12. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM .

                                                                               --1 TS 6.6-10 B. Unigue Reporting Reguirements

1. lnservice lnsgection Evaluation Special summary technical report shall be submitted to the Director of Reactor Licensing, Office of Nuclear Reactor Regulation, NRC, Washington, D.C. 20555, after 5 years of operation. This report shall include an evaluation of the results of the inservice inspection program and will be reviewed in light of the technology available at that time.

2. Annual Radiological Environmental Operating Report 1.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

3. Semiannual Radioactive Effluent Release Report3 The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50 .

TS 6.6-11

4. Containment Leak Rate Test Each containment integrated leak rate test shall be the subject of a summary technical report. Upon completion of the initial containment leak rate test specified by proposed Appendix J to 1 O CFR 50, a special report shall, if that Appendix is adopted as an effective rule, be submitted to the Director, Division of. Reactor Licensing, USNRC, Washington, D. C. 20555, and other containment leak rate tests specified by Appendix J that fail to meet the acceptance criteria of the appendix, shall be the subject of special summary technical reports pursuant to Section V.B of Appendix J:

a. "Report of Test Results - The initial Type A tests shall be subject of a summary technical report submitted to the Commission approximately 3 months after the conduct of the test. This report shall include a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supplemental test method, and the test program selected as applicable to the initial test, and all subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data to the extent necessary to demonstrate the acceptability of the containment's leakage rate in meeting the acceptance criteria."

            "For periodic tests, leakage rate results of Type A, B, and C tests that meet the acceptance criteria of Sections 111.A.7, 111.B.3, respectively, shall be reported in the licensee's periodic operating report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria of Sections 111.A.7, 111.B.3, and 111.C.3, respectively, shall be reported in a separate summary report that includes an

TS6.6-12 analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrument error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included." C. Special Reports In the event that the Reactor Vessel Overpressure Mitigating System is used to mitigate a RCS pressure transient, submit a Special Report to the Commission within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or the administrative controls on the transient and any corrective action necessary to prevent recurrence. FOOTNOTES

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.

2. This tabulation supplements the requirements of Section 20.407 of 10 CFR Part 20.

3. A single submittal may be made for a multi-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.

TS 6.8-1 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE CALCULATION MANUAL A. Process Control Program (PCP) Changes to the PCP:

1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.12. This documentation shall contain:

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

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

2.

regulations. Shall require review and acceptance by the SNSOC and the approval of the Station Manager prior to implementation. B. Offsite Dose Calculation Manual (ODCM) Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.12. This documentation shall contain:

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

TS 6.8-2

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall require review and acceptance by the SNSOC and the approval of the Station Manager prior to implementation.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented .

Attachment 2 Discussion of Proposed Changes

Surry Units 1 and 2 Virginia Electric and Power Company

Discussion of Proposed Technical Specification Change Introduction The proposed changes to the Surry Units 1 and 2 Technical Specifications removes the Radiological Effluent Technical Specifications. These specifications are being removed to the Offsite Dose Calculation Manual (ODCM) or the Process Control Program (PCP).- Technical Specifications relating to these documents are also being amended due to their expanded role.

Background

This proposed change is based on the NRC's Generic Letter 89-01 dated January 31 , 1989. The letter stated that the NRC will approve a Technical Specification amendment to delete the Radiological Effluent Technical Specifications if the requirements are relocated to the ODCM or PCP. The letter was very specific about what changes are acceptable and warned that proposed amendments that deviate from its guidance will require a longer, more detailed review. It stated that conforming amendment requests will be expeditiously reviewed. This proposed change follows the guidance in the letter. Some changes were needed due to differences between Surry and Standard Technical Specifications. One requirement was corrected when it was moved to the ODCM. The correction was necessary due to an error in Specification 3.11.A.3.a, which requires that the liquid radwaste treatment system be used to reduce monthly projected doses due to liquid effluents to 0.06 mrem whole body and 0.2 mrem to the critical organ. The phrase "from each unit" was inadvertently omitted from the Specification. The basis of the Specification is that in order to keep effluents as low as reasonably achievable, the limits were set at a "suitable fraction" (1 /4) of the limits in Section II.A of Appendix I, 10 CFR 50. The Appendix I limits are on a per reactor basis. The missing phrase is present in the equivalent North Anna Specification and in Specification 3.11.1.3 of the draft Revision 5 of the Standard Technical Specifications for Westinghouse PWRs. Note that the ODCM is intended to apply to both North Anna and Surry and it could be confusing to have an exception for Surry where it is not logically expected. The ODCM Section 6.2.4.a has therefore been corrected to be consistent with North Anna and Standard Technical Specifications and Appendix I.

Description of the Proposed Change

1. In the index, item 3.11 "Effluent Release" is changed to "Radioactive Gas Storage."

2. In the index, item 4.9, the phrase "Effluent Sampling and Radiation" is changed to "Radioactive Gas Storage."

3. In the index, item 6.9 is deleted.

4. Specification 1.0.M, the ODCM definition, is replaced with item number 1.17 from Enclosure 3 of the Generic Letter, except references to Specifications 6.8.4, 6.9.1.3 and 6.9.1.4 are changed to 6.4, 6.6.B.2 and 6.6.B.3 respectively.

The revision reflects the expanded role of the ODCM.

5. Specification 1.0.P, the PCP definition, is replaced with item number 1.22 from Enclosure 3 of the Generic Letter. This adds references to 10CFR61 and burial ground requirements which were previously included in "other requirements."

6. Section 1.0.R is deleted. The requirements are added to the PCP. Definitions S through Ware re-lettered R through V.

Although not reflected in all of the titles, Specifications 3.7, 3.11, 4.1 and 4.9 cover waste gas storage and radioactive effluents. The following changes delete effluent monitoring requirements, which have been added to the ODCM, but retain the gas storage monitoring requirements.

7. The phrase "radioactive liquid and gaseous effluent" in Specification 3.7.E is replaced with "explosive gas."

8. The phrase "and Table 3.7-5(b)" is deleted.

9. The phrase "Specifications 3.11.A.1 and 3.11.B.1" is changed to "Specification 3.11.A.1."

10. The last sentence of 3.7.E, before 3.7.E.1 is deleted.

11. The phrase "a radioactive liquid or gaseous effluent" in 3. 7.E.1 is changed to "an explosive gas." *

12. The phrase "without delay suspend the release of radioactive liquid or gaseous effluents monitored by the affected channel and" in 3. 7.E.1 is deleted.

13. The phrase "or change the setpoint so it is acceptably conservative" in 3.7.E.1 is replaced with "and take the action shown in Table 3.7-5(a)."

14. The phrase "radioactive liquid or gaseous effluent" in Specification 3. 7. E.2 is changed to "explosive gas."

15. The phrase "or Table 3.7-5(b)" in 3.7.E.2 is deleted.

16. The phrase "explain in the next Semiannual Radioactive Effluent Release Report" in 3.7.E.2 is replaced with "submit a Special Report to the Commission (Region 11) to explain."

17. The paragraph titled "Automatic Function Operated from Radiation Monitors" in the basis section, page 3.7-8 is deleted.

18. On page 3.7-9, the paragraph titled "Radioactive Liquid Effluent Monitoring Instrumentation" is deleted.

19. The first two sentences of the next paragraph are deleted.

20. In the next sentence the phrase "This instrumentation also includes provisions" is changed to "Instrumentation is provided.II

21. Reference number four on page 3.7-9c is deleted.

22. In Table 3.7-5, items 1, 3, 4, and 7 are deleted. The remaining items are renumbered. References to Specification 4.9 are deleted. The words "and exhaust" are deleted to reflect the removal of the purge exhaust fans by a previous design change.

23. Table 3.7-5(a) is deleted.

24. Table 3.7-5(b) is changed to 3.7-5(a). In the title, "Radioactive Gaseous Effluent" is changed to "Explosive Gas." Items 1, 3 and 4 and Action items 1, 2 and 3 are deleted. "Action 4" is renumbered "Action 1." The page number is changed to 3.7-20a.

25. The title of section 3.11 is changed to "Radioactive Gas Storage."

26. The "Applicability" section of 3.11 is changed to: "Applies to the storage of radioactive gases."

27 Under "Objective," "and liquid" is deleted, "released" is changed to "stored" and everything after "released" is deleted.

28. All of 3.11.A and sections 3.11.B.1 through 3.11.B.4 are deleted.

29. Our letter, serial number 90-297, dated May 25, 1990 proposed changes to section 3.11.5. The Specification in Attachment 1 includes these changes, which are indicated by a double bar. In addition to the previously proposed changes, the "5" in 3.11.B.5 is changed to "A" and the subsection labels "a" and "b" are changed to "1" and "2." A new subsection 3 is added: "The requirements of Specification 3.0.1 are not applicable." The new subsection is needed

because section 3.11. F is to be deleted. The Specifications are moved to page 3.11-1.

30. The "6" in 3.11.B.6 is changed to "B" and the subsection labels "a" and "b" are changed to "1" and "2." Because section 3.11.F is to be deleted, a new subsection 3 is added: "The requirements of Specification 3.0.1 are not applicable." The Specifications are moved to page 3.11-1 .

31. Sections 3.11.C through 3.11.F are deleted.

32. In the 3.11 Bases section, everything except the "Explosive Gas Mixture" and "Gas Storage Tanks" subsections is deleted. The remaining sections are moved to page 3.11-2.

33. Table 4.1-1 A is deleted.

34. Table 4.1-1 B is changed to Table 4.1-1 A. In the title "Radioactive Gaseous Effluent" is changed to "Explosive Gas." Items 1, 3 and 4 and the "Source Check" column and all frequency footnotes except "D," "M" and"Q" are deleted.

Item 2 is renumbered and the page number is changed to 4.1-8c

35. In the title of section 4.9, "Effluent Sampling and Radiation" is changed to "Radioactive Gas Storage."

36. Under "Applicability," "and recording" is deleted and "effluents" is changed to

37.

38 "gas storage."

The "Objective" section of 4.9 is changed to "To ascertain that waste gas is stored in accordance with Specification 3.11."

Sections 4.9.A through 4.9.E and 4.9.H through 4.9.K are deleted. The requirements have been added to the ODCM.

39. The labels for subsections F and G are changed to "A" and "B" and they are moved to page 4.9-1. References to Specifications 3.11.B.5 and 3.11.B.6 and Table 3.7-5(b) are changed to 3.11.A, 3.11.B and 3.7-5(a) respectively.

40. Tables 4.9-1 through 4.9-5 are deleted. The requirements are added to the ODCM.

41. Two new subsections, N and O are added to section 6.4. These are the same as sections 6.8.4.g and 6.8.4.h of Enclosure 3 of Generic Letter 89-01 except in 6.8.4.g, paragraph 10, which does not apply to PWRs, is deleted and paragraph 11 is renumbered 10. The additions are programmatic requirements deleted elsewhere.

42. A new item number 12 is added to section 6.5.B: "Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL

and the PROCESS CONTROL PROGRAM."

43. Sections 6.6.B.2 and 6.6.B.3 are replaced with the text of sections 6.9.1.3 and 6.9.1.4 from Enclosure 3 of the Generic Letter. The change simplifies the requirements for the Annual Radiological Environmental Operating Report and the Semi-Annual Radioactive Effluent Release Report. Details have been added to the ODCM. The pages in the remainder of section 6.6 are renumbered.

44. 6.8.A and 6.8.B are replaced with the text of section 6.13 and 6.14 of the Generic Letter's Enclosure 3 except all references to Specification "6.10.3.o" are changed to "6.5.B.12" and "URG" is changed to "SNSOC." Also, subsection labels "a," "b" and "c" are changed to "1," "2" and "3" and labels "1" and "2" are changed to "a" and "b."

45. Section 6.9 is deleted. The requirements are added to the PCP .

Safety Analysis Although the proposed changes simplify the. Technical Specifications, there is no

reduction in requirements because of additions to the ODCM and PCP. The following table outlines the disposition of each requirement removed from the Technical

 . Specifications.

Specification Addition 1.0.R PCP

3. 7. E,

4.1 and 4.9.A (liquid effluents) ODCM 6.2.2

                                                . TS 6.4.N.l 3.7.E, 4.,1 and 4.9.A (gaseous effluents)      ODCM 6.3.2 TS 6.4.N.1 3.11.A.1                                       ODCM 6.2.1 TS 6.4.N.2-3 3.11.A.2                                       ODCM 6.2.3 TS 6.4.N.4-5 3.11.A.3

ODCM 6.2.4 TS 6.4.N.6 3.11.B.1
ODCM 6.3.1 TS 6.4.N.3 TS 6.4.N.7 3.11.B.2 ODCM 6.3,3 TS 6.4.N.5 TS 6.4.N.8 3 11.B.3. ODCM 6.3.4 TS 6.4.N.5 TS 6.4.N.9 3.11.B.4 ODCM 6.3.5 TS 6.4.N.6 3.11.C ODCM 6.4 TS 6.4.N.10 3.11.D.1 ODCM 6.5.1 TS 6.4.0.1 3.11.0~2 ODCM 6.5.2 TS6.4.0.2 3.11.D.3 ODCM 6.5.3 TS 6.4.0.3 3.11.E PCP 4.9.B
ODCM 6.2.5 4.9.C . ODCM 6.2.3 ODCM 6.3.3 4.9.D
ODCM 6.2.4 ODCM 6.3.5

Specification Addition 4.9.E ODCM 6.3.1 ODCM 6.3.3 ODCM 6.3.4 4.9.H ODCM 6.5.1 4.9.I ODCM 6.5.2 4.9.J ODCM 6.5.3 4.9.K PCP 6.6.B.2 ODCM 6.6.1 6.6.B.3 ODCM 6.6.2 6.9 PCP

Attachment 3 10 CFR 50.92 Evaluation

Surry Units 1 and 2 Virginia Electric and Power Company

Basis for No Significant Hazards Determination The proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92 because operation of Surry Units 1 and 2 in accordance with this change would not: (1) involve a significant increase in the probability or consequence of an accident previously evaluated. This change does not alter the conditions or assumptions of any accident analysis. (2) create the possibility of a new or different kind of accident from any accident previously identified. This change does not alter the conditions or assumptions of any accident analysis. This is not an actual hardware change. (3) involve a significant reduction in a margin of safety. This change does not alter the conditions or assumptions of any accident analysis. It is not an actual hardware change. Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that this change does not involve a significant hazards consideration.

SUADM-LR-12 ATTACHMENT 1 PAGE 1 OF 13 SAFETY EVALUATION NO. JAr~ 1 .J ; 3s.: STATION/UNIT(S): __S__v_~_r...v _____l_-+-_L... SAFETY EVALUAnON F'ORH PART A - RESOLUTION

SUMMARY

REPORT (1) List the governing document(s) for which the safety evaluation is being performed: T,:t.hr:ii~a( !;t'e~if,'t;,4l,tz(TS (2) Briefly describe the change, test,. or experiment being evaluated:

1

              ~2~{=fr!; l~~c;it,:;/~

(3) Briefly describe the purpose for this change, test, or experiment:

T~ ~irr':J:Y ,~?2:~~al .r~4£f~~qf:11,

                        ?                         .     .

a< Based on the information contained herein, the following is required and is attached (check as appropriate):

  ~      10 CFR 50.59 safety evaluation (PART D, QUESTIONS 1-4) a 10 CFR 72.48 safety evaluation (SPS/ISFSI only - PART D, QUESTIONS 1-6)

Briefly state the major issues considered, the reason for the change, test, or experiment should be allowed, and why an unreviewed safety question does or does not exist (a simple statement of conclusion f';,/' ;,;if ~dt ~;"~"l ,=~£ie~~"' :* ~bece+Mbc alone is insufficient; attach additional sheets if needed): J,~ r~~ ~'< Ir &'"1. >>e;,_r.'$ ~:;:~ i~v,,;.,~ ltle'< ts

SUADM-LR-12 ATTACHMENT 1 PAGE 2 OF 13 JA:l 1 3 ;S9Q SAFETY EVALUATION NO. STAT ION/UN IT ( S):

                         >vt'l'Y              )+   "2 7......-""""'-...;....---

PART A - RESOLUTION

SUMMARY

REPORT (Continued) Recommended approval - Cognizant Supervisor: _ _Approved ___Disapproved Approved Requires further

                                                  --as modified ---evaluation SNSOC Chairman
                 --------------- Date----------

Comments:

SUADM-LR-12 ATTACHMENT 1 PAGE 3 OF 13 PART B - APPLICABLE REFERENCES J I (1) Identify applicable UFSAR section(s): (2) Identify applicable Technical Specification section(s):

         /. OJ 3. 7) 3. I)) Lf:. f ) 't* 1 J p. if- ,1 /, - 5 > -6. _6_5 ___,.£-.~f!?-'-

c::,

(3) Identify any other references used in this review:

           ~:S7;+~       Drz5~       c(4!£v/~h~~    ~~a~ - - -

PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION NOTE: Items denoted by a double asterisk(**) which are answered with a "YES," require Engineering approval.

1. Will the operation of any safety related system

 ---Yes **
                  -  No or component as described in the SAR and/or the Technical Specifications be altered? This includes abandonment of equipment or extended

periods of equipment out of service
Explain: * * *

                              'i£G;;;!0e,;rc":Ji' b~~~

Yes v No 2. Will the activity alter the performance character-

 - - - **                  istics of any safety related system or compo-nent? (Note: Action Statements, jumpers, and temporary modifications should be reviewed.)

Explain:

                                       ;;;3i t;;;;-; t ;t:rh d1 .

Ll e <.b..~~ e: w/ 1;1,

                                                   ~L 4      ~    t:_I,,:    !"2:ET-,

6

SUADM-LR-12 ATTACHMENT 1 PAGE 4 OF 13 PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued) v* No 3. Will the ability of operators to control or

 ---Yes                monitor the plant be reduced in any way?

3:0; Explain:

                                   ~....,.f:i._:Z_f_r;,-;;-~-~:-t;-.L,-/<.--£-15-W1-fJ,,--.

Yes V No 4. If a jumper is involved, are testing requirements as stated on the jumper adequate to ensure operability after installation as well as after removal? Explain: Na _i_v_M-~-e-c--,-j--,-4-Ve_l_v-~-id~.--------

                                  ,J         )

Yes /No 5. Could the proposed activity affect reactivity?

 ---**                 If "Yes," explain (the Reactor Engineer/designee must approve the explanation by initialing):

(Rx. Eng.

                                     ------)

Yes 6. Will the activi'ty significantly increase the potential for personnel injury or equipment damage? Explain: 1h !!'. chcan:9c <<n !~1 r-r kc.a f,:;-, R !?1-5

SUADM-LR-12 ATTACHMENT 1 PAGE 5 OF 13 JAU 1 PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (Cc~tinued) J - *

Yes 1 ~ No 7. Will the activity create or increase the levels of radiation or airborne radioactivity? If so, will the change result in a significant unre-viewed environmental impact, a significant increase in occupational exposure, or signifi-cant change to dose to operators performing tasks outside the filtered air boundary during a DBA (GDC-19). If "Yes," explain (the Superin-tendent of Health Physics/designee must approve the explanation by initialing):

T. .5 . .

8. Will the activity change or decrease the

 ---Yes                 effectiveness of the emergency plan? If "Yes,"

explain (the Emergency Preparedness Coordinator/ designee must approve the explanation by Tt0 <?~?=:k2 initialing): f~;fi :mat:in,ey 't~,/v,-~

                                                                         ;~e fe (tJ1 at      C b@"l& ~ {

(E.P. Coordinator ) Yes ../' No 9. Will the consequences of failure for this activity affect the ability of systems or components to perform safety functions? Briefly describe the modes and consequences of failure considered duiing this evaluation: l h~ c.h-;;ri~ ~ anJ;: cel,ec:,af<<,

SUADM-LR-12 ATTACHMENT l PAGE 6 OF 13 PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued) .,, .... " J,-.,1 l

 ---Yes**   ./ No    10. Will the activity cause equipment to be exposed (or potentially exposed) to adverse conditions including      those     created      be    temperature, pressure, humidity, or radiation? If adverse conditions are possible, could these conditions lead     to equipment failure, or a dangerous atmosphere?

Explain: Na c haoge£i

r"1"1!¢1:e a~ J",, vd.Ji.r:d*

py t?f2¢:r'>aVRl'J4/ Yes v' No 11. Could the failure of the activity feedback into protective circuitry? Explain: f.2£471:kla fu4 ~;;l ;'; f' /7v G-ha

gi ~ ; t"'.h:f-an&c: r";';;'tes

            ~
 ---Yes**       No

12. Could the activity cause a loss of separation of instrument channels/trains or electrical power supplies?

                          ~P~~i~: "C,~}-b :J'1 'Y[;; b J!,;~/t.~ r ~t';,;:;'2

Yes 13. Will the activity_. involve the addition deletion of any electrical loads on the vital bus?

Explain: or [<£12 ___Yes~No 14. Will the activity adversely affect the ability of a system or component to maintain its integrity or code requirements? Will the activity add or adversely affect components in the ASME XI/ISI program? Explain=--~----------------~---------------- Thc::: c)1all\@< enJ>c cela~ate,:j RET.s

SUADM-LR-12 ATTACHMENT 1 PAGE 7 OF 13

PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued)

 ---Yes v',No     15. Will the activity reconfigure, eliminate, or add components     and/or piping to the single or two-phase erosion/corrosion piping inspection program?

JANl Explain: rbe wt?-/-,'.,; f-v i?/1 bv re laccafc!i RETz witnevt :2i<b:zfgr1f,ve: ch qrz@e-

 ---Yes ---  ,._,/'No 16. Will additional surveillance requirements, as defined in the Technical Specifications, be necessitated by the activity?

Explain: ~e~~ 5 J/Yv'~i//a~e-r; r.t:.e1tu~eft[l?HB i3(f- reld?c£!led) log+ !::1..£:n ~ ar"t!!! added Yes v' No 17. Will the applicable Technical Specification basis description be altered by the activity? Explain: J,, J,~ h

de? c ,* H
--- Yes ~No

18. Will the activity result in a violation of any Limiting Conditions for Operation (LCO's), as defined in the Technical Specifications?

Explain: -5am~ Le??.., will be r~lec-a&,d t~f- ,;15;//;;~d w,fh t/1<:nt 1111 II _, v""No 19. Were any other concerns or items identified

 ---Yes                   during this review?          If "Yes," explain:

SUADM-LR-12 ATTACHMENT 1 PAGE 8 OF 13

PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued) * ***"'!

JAN 2. NOTE: THESE ITEMS ARE INCLUDED FOR CONSIDERATION OF POTENTIAL IMPACT. IF THE ANSWER TO ANY OF THE FOLLOWING QUESTIONS IS "YES", A DETAILED REVIEW MUST BE PERFORMED, AND THE RESULTS OF THIS REVIEW MUST BE DOCUMENTED ON A SEPARATE SHEET WHICH REFERENCES THE SAFETY EVALUATION NUMBER AND THE RESPECTIVE PART C ITEM NUMBER. ATTACHMENT 2 PROVIDES GUIDELINES FOR THE DETAILED ENGINEERING REVIEW OF SOME OF THESE ITEMS.

20. STATION SECURITY Yes ...,/No Will the activity deactivate a security-related system or breach a security barrier?

21. FIRE PROTECTION/APPENDIX R:

 ---   Yes   ../" No      a. Will the activity add or eliminate combustible material from plant areas?

any Yes /No b. Will the activity change or affect and plant

 --- **                      structure that acts as a fire barrier?

Yes /No c. Will the activity impact the performance of an existing fire protection or detection system?

22. EQUIPMENT QUALIFICATION/CLASSIFICATION

a. Will the activity adversely affect any

 ---Yes ** ___               Class IE el~ctrical equipment located in a potentially harsh environment (as designated by the Environmental Zone Descriptions/EZDs)?

Yes /No b. Will the activity have the potential to alter

        **                   any     of    the   environmental   parameters identified in the EZDs?

Yes ~No c. Will the activity have the potential to

        **                   affect any of the electrical distribution systems (i.e., 4KV, 480V, 120VAC, etc.)?

Yes /No d. Will the activity change or affect equipment on the EQML or Q-List. Yes v"No e. Will the activity add, eliminate, or have the potential to affect ASME Section XI equipment?

 ---Yes        v'No       f. Will the activity change a setpoint in the PLS Document?

SUADM-LR-12 ATTACHMENT 1 PAGE 9 OF 13 PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued) .* .....

                   /     23. SEISMIC Yes      ./ No         Could the activity be adversely affected by a seismic event or could the activity affect surrounding equipment during a seismic event?

24. HUMAN FACTORS Yes / No a. Will the activity change instrumentation or controls in the control room or on the auxiliary shutdown panel?

 ---Yes                      b. Will the activity alter the control room the auxiliary shutdown panels?

or

25. SAFETY PARAMETER DISPLAY SYSTEM/ERF Yes v"'" No a. Will the activity change any of the equipment

       **                       associated with the        SPDS/ERF,    including SPDS/ERF computer inputs?

26. STATION COMPUTERS Yes .,..-- No a. Will the activity have a significant

 --- **                         potential   to modify or add software           to station computers?

--- Yes ~ No

27. ENVIRONMENTAL IMPACT/FLOODING

a. Will the */ ac.tivity impact more one-fourth of an acre of land, work in navigable waters, wells, dams, or wetlands, and/or involve any wastes or discharges?

than

 ---Yes     ~No              b. Will the activity involve changes to site terrain, features, or structures?
                 /
              .,. No
 ---Yes** __       _         c. Will the activity have a significant potential to expose safety related equipment to     flooding       via      fluid      system equipment/piping malfunction or failure?
 ---Yes** --- /No       28. REG. GUIDE 1.97 Will the activity have a significant potential to    modify equipment and/or instrumentation associated with Reg. Guide 1.97 variables?

SUADM-LR-12 ATTACHMENT 1 PAGE lei OF 13

PART C - ITEMS CONSIDERED BY THIS SAFETY EVALUATION (continued)

Yes

29. HEATING-VENTILATION-AIR-CONDITIONING

a. Will the activity have a significant potential to increase the heating or cooling loads in plant areas and/or to plant

                                                                                   **. i equipment?

Yes / No b. Will the activity change the existing

 ---**                     ventilation system in any way?

Yes / No c. Will the activity change any building

 ---**                     structures,     including   walls,     ceilings, windows, doors, or floors, such that existing HVAC systems may be affected?

30. HEAVY LOADS

            /  No       Will     the    activity   involve    heavy   loads
 ---  Yes (including the transfer of heavy loads in areas housing safety related equipment)?

31. Will there be an introduction of detrimental Yes v" No materials into the containment or other plant areas? (For example, Zinc, and Aluminum alloys are not allowed in the containment because of

the potential generation of H2 gas from chemical reactions with these materials.

                        "Yes,"

explain: If

SUADM-LR-12 ATTACHMENT 1 PAGE 11 OF 13

PART D - 10 CFR 50.59 SAFETY EVALUATION l!'.'.I'" .. ~ . ~

Note: This section is based on the results of the items considered in PART C, and therefore must be completed subsequent to PART C. UNREVIEWED SAFETY QUESTION DETERMINATION:

1. Which accidents previously evaluated in the SAR Yes / No a. Could the activity increase the probability of occurrence for* the accidents identified above?

basis for your conclusion:

                                     .*Fvi'f                   ?

f T, fiJJ,JJ~ Yes /No b. Could the activity increase the consequences of the accidents identified above? State the basis

Yes I/"' No c. Could the activity create the possibility for an accident of a different type than was previously evaluated in the SAR? State the basis for your conclusion: --- ~ J1 /1/111.;P/Vt!'

                             ~f.     -f

SUADM-LR-12 ATTACHMENT 1 PAGE 12 OF 13 PART D - 10 CFR 50.59 SAFETY EVALUATION (continued)

2. What malfunctions of equipment related to safety previously evaluated in the SAR were considered?

yV6DJ:

 ---Yes      .,/"No   a. Could the activity increase the probability of occurrence for the malfunctions identified above?

State the basis for your conclusion: I~vc>n~~\f, H:z -~ad {j,? lf;;it-1 will b <"' r 0ed I l~b }

                                                                                 ~.c 2,
 ---Yes    ,.,/  No   b. Could the activity increase the consequences of the malfunctions identified above? State the basis
                                                   )
 ---Yes     .,/ No    c. Could the activity create the possibility for a malfunction of equipment of a different type than was previously evaluated in the SAR? State the basis for your conclusion:

T}v:! . *¥t,' viar--t, y haw: t:io h ,!;)t:-

w//J_ t1-a-t'";"".-c-:-fu-dt-t14--e-~..,.h-!?-

Jt [~ .r£,!:!cfc4ft!!'d.

             ~ No
 ---Yes               3. Has the margin of safety of any part of the Technical Specifications as described in the BASES section been reduced?

Explain:

                            ~ r'
                                     -4/   e iJ i <;t+/-i.1?Yl-:7 ~ h-;-J 1:::l:J.£.V~
                                            .c ff ,.,~_,;17~~i '"~f:Lz;:iifii;~
   ~ Yes ___No        4. Does the proposed change, test, or experiment require a change to the Technical Specifications?

h Explain=-----------....---..------------ tJo <(2 c. e111e:v=:".- '1"f1uwa ta J&Y-£c.e4!/v1r~1-t~ $ 2,r2 e<./ £(?4 '7~/J-:,

SUADM-LR-12 ATTACHMENT 1 PAGE 13 OF 13 PART D - 10 CFR 50.59 SAFETY EVALUATION (Continued)

  - -Yes     ./' No     5. Does the proposed change, test, or experiment involve a significant unreviewed environmental impact? (10CFR72.48 ONLY)

Explain: Tl? e. g_ ;::.Tj £:1..Y-l'E' b rt'ng k'.)4 a ,1~;J witbk+ .(.eX;taatr*v,::

 - -Yes      v,,..-./No 6. Does the proposed change, test, or experiment involve a significant increase in occupational exposure? (10CFR72.48 ONLY) State the basis for your conclusions:
                           -rq e [?ET~ 11(~ .beiny J1YJav~d NOTE: IF THE RESPONSE TO QUESTIONS 1-4 (ABOVE) IS "NO," THE PROPOSED ACTIVITY MAY BE IMPLEMENTED, FOLLOWING SNSOC APPROVAL, PROVIDING THAT COMPLETE DOCUMENTATION IS MAINTAINED. IF THE RESPONSE TO ANY PART OF QUESTIONS 1-~ IS "YES," AN APPLICATION FOR AMENDMENT TO THE OPERATING LICENSE MUST BE SUBMITTED AND APPROVED BY THE NRC PRIOR TO IMPLEMENTATION OF THE CHANGE, TEST, OR EXPERIMENT.

IN ADDITION. FOR THE SURRY ISFSI, IF THE RESPONSE TO QUESTION 5 OR 6 IS "YES", AN APPLICATION FOR AMENDMENT TO THE ISFSI LICENSE

MUST ALSO BE SUBMITTED AND APPROVED PRIOR TO IMPLEMENTING THE CHANGE, TEST, OR EXPERIMENT. ' ' ...:.

BASED ON THE PRECEDING, THE PROPOSED ACTIVITY (V) WILL-OR-

 -{) WILL NOT RESUI:.1' IN AN UNREVIEWED SAFETY QUESTION Alffi/OR REQUIRE A LICENSING AMENDMENT.

Prepared by: Pahe1-r J1 Ne,/ Title____.)_t_..a.....l....t___1::._=-_11:9._,...,-_v1_~_-~

                                                                                                      ....v___

Signature:~~ /11 /2&£ Date :_ _ 3_/; __l_~.....,,-/_.._CJ,P Reviewed by:

               ------------Date:------------

Date: Design Authority Reviewed by:

               ------------Title------------

Signature:

               ------------Date:-----------=

(Documenting concurrence of** items in Part C answered "YES") (May be N/A)

Attachment 4 Offsite Dose Calculation Manual Virginia Electric and Power Company

Station Administrative VIRGINIA POWER

         ~ : Offsite Dose Calculation Manual Procedure Lead Department: Radiological Protection Procedure Number:               Revision Number:          Effective Date:

VPAP-2103 0 05/31/90 Surry Power Station North Anna Power Station Approved by: Approved by: 3,.t~-'ro ~~ J/,-1-1@ Date S N S O ~an Date Approved ~y:

                                      ~hit&

Date rations -'f'oo SJq oJx)B 111 pf.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE20F 116

Section TABLE OF CONTENTS Page 1.0 PURPOSE 5 2.0 SCOPE 5

3. 0 REFERENCE/COMMITMENT DOCUMENTS 5

4. 0 DEFINITIONS 7

5. 0 RESPONSIBILITIES 10 6.0 INSTRUCTIONS 11 6 .1 Sampling and Monitoring Criteria 11

6. 2 Liquid Radioactive Waste Effluents 11 6.2.1 Liquid Effluents Concentration Limitations 11 6.2.2 Liquid Monitoring Instrumentation 12 6.2. 3 Liquid Effluent Dose Limit 15 6.2.4 Liquid Radwaste Treatment 18 6.2.5 Liquid Sampling 19

6. 3 Gaseous Radioactive Waste Effluents 19 6.3.1 Gaseous Effluent Dose Rate Limitation 19 6.3.2 Gaseous Monitoring Instrumentation 21 6.3.3 Noble Gas Effluent Air Dose Limit 24 6.3.4 I-131, H-3, and Radionuclides In Particulate Form Effluent Dose Limit 26 6.3.5 Gaseous Radwaste Treatment 29

6. 4 Total Dose Limit to Public From Uranium Fuel Cycle Sources 31

6. 5 Radiological Environmental Monitoring 32 6.5.1 Monitoring Program 32 6.5.2 Land Use Census 34 6.5.3 Interlaboratory Comparison Program 35

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6 . 6 Reporting Requirements 6.6.1 Annual Radiological Environmental Operating Report 6.6.2 Semiannual Radioactive Effluent Release Report 36 36 37 6.6.3 Annual Meteorological Data 38 6.6.4 Changes to the ODCM 38 7 .0 Records 39 ATTACHMENTS 1 Surry Radioactive Liquid Effluent Monitoring Instrumentation 40 2 North Anna Radioactive Liquid Effluent Monitoring Instrumentation 41 3 Surry Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 43 4 North Anna Radioactive Liquid Effluent Monitoring
5 6

Instrumentation Surveillance Requirements Liquid Ingestion Pathway Dose Factors for Surry Station North Anna Liquid Ingestion Pathway Dose Factor Calculation 44 46 47 7 NAPS Liquid Ingestion Pathway Dose Commitment Factors for Adults 51 8 Surry Radioactive Liquid Waste Sampling and Analysis Program 52 9 North Anna Radioactive Liquid Waste Sampling and Analysis Program 55 10 Surry Radioactive Gaseous Waste Sampling and Analysis Program 58 11 North Anna Radioactive Gaseous Waste Sampling and Analysis Program 62 12 Gaseous Effluent Dose Factors for Surry Power Station 65 13 Gaseous Effluent Dose Factors for North Anna Power Station 68

14 Surry Radioactive Gaseous Effluent Monitoring Instrumentation 71

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15 North Anna Radioactive Gaseous Effluent Monitoring Instrumentation 16 Surry Radioactive Gaseous Effluent Monitoring 73 Instrumentation Surveillance Requirements 75 17 North Anna Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 76 18 Critical Organ and Inhalation Dose Factors for Surry 78 19 Critical Organ and Inhalation Dose Factors for North Anna 80 20 Surry's Radiological Environmental Monitoring Program 81 21 North Anna's Radiological Environmental Monitoring Program 83 22 Surry's Environmental Sampling Locations 87 23 North Anna's Environmental Sampling Locations 91 24 Detection Capabilities for Surry Station Environmental Sample
25 Analysis Detection Capabilities for North Anna Station Environmental Sample Analysis 2 6 Reporting Levels for Radioactivity Concentration in 95 97 Environmental Samples at Surry Station 99 2 7 Reporting Levels for Radioactivity Concentration in Environmental Samples at North Anna Station 100 28 Surry Meteorological, Liquid and Gaseous Pathway Analysis 101 2 9 North Anna Meteorological, Liquid and Gaseous Pathway Analysis 109

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1.0 PURPOSE The Offsite Dose Calculation Manual (ODCM) establishes the requirements of the Radioactive Effluent and Radiological Environmental Monitoring Programs. Methodology and parameters are provided for calculation of offsite doses resulting from radioactive gaseous and liquid effluents, for gaseous and liquid effluent monitoring alarm/trip setpoints, and for conduct of the Environmental Monitoring Program. Requirements are given for the completion of the Annual Radiological Environmental Operating Report and the Semi-Annual Radioactive Effluent Release Report required by Station Technical Specifications. Calculation of offsite doses due to radioactive liquid and gaseous effluents are performed to assure that:
Concentration of radioactive liquid effluents to the UNRESTRICfED AREA will be limited to the concentration levels of 10 CFR 20, Appendix B, Table II, column 2 for radionuclides other than dissolved or entrained noble gases;
Exposure to the maximum exposed MEMBER OF THE PUBLIC in the UNRESTRICTED AREA from radioactive liquid effluents will not result in doses greater than the liquid dose limits of 10 CFR 50, Appendix I;
Dose rate at and beyond the SITE BOUNDARY from radioactive gaseous effluents will be limited to the annual dose rate limits of 10 CFR 20;
Exposure to the maximum exposed MEMBER OF THE PUBLIC in the UNRESTRICTED AREA from radioactive gaseous effluents will not result in doses greater than the gaseous dose limits of 10 CFR 50, Appendix I; and
Exposure to the maximum exposed MEMBER OF THE PUBLIC will not exceed 40 CFR 190 dose limits 2.0 SCOPE This procedure is applicable to the Radioactive Effluent and Environmental Monitoring Programs performed at Surry and North Anna Stations.

3. 0 REFERENCES/COMMITMENT DOCUMENTS 3 .1 References 3 .1.1 10 CFR 20, Standards for Protection Against Radiation 3 .1.2 10 CFR 50, Domestic Licensing of Production and Utilization Facilities 3 .1. 3 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power

3 .1.4 Operations TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites L

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3.1.5 3 .1. 6 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, Rev. 1, U.S. NRC, June 1974 Regulatory Guide 1.109, Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CPR 50, Appendix I, Rev. 1, U.S. NRC, October 1977 3.1. 7 Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light - Water - Cooled Reactors, Rev. 1, U.S. NRC, July 1977 3.1.8 Surry and North Anna Technical Specifications (Units 1 and 2) 3.1.9 NUREG-0324, XOQDOQ, Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, U.S. NRC, September 1977 3.1.10 NUREG/CR-1276, Users Manual for the LADTAP II Program, U.S. NRC, May, 1980 3.1.11 NUREG-0597, User's Guide to GASPAR Code, U.S. NRC, June, 1980 3.1.12 Radiological Assessment Branch Technical Position on Environmental Monitoring, November, 1979, Rev. 1
3.1.13 3.1.14 NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Stations", October, 1978 NUREG-0543, February 1980, Methods for Demonstrating LWR Compliance With the EPA Uranium Fuel Cycle Standard (40 CPR Part 190) 3.1.15 NUREG-0472, Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors, Rev. 3, March 1982 3.1.16 Environmental Measurements Laboratory, DOE HASL 300 Manual 3.1.17 NRC Generic Letter 89-01, Implementation of Programmatic Controls for Radiological Effluent Technical Specifications (RETS) 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 3.1.18 UFSAR (Surry and North Anna) 3.1.19 Nuclear Reactor Environmental Radiaiton Monitoring Quality Control Manual, IWL-0032-361

3. 2 Commitment Documents None

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4. 0 DEFINITIONS NOTE: Terms which are defined in Surry and North Anna Technical Specifications appear as all capitalized letters in the text of this procedure for identification.

4. 1 Channel Calibration CHANNEL CALIBRATION is defined as the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter

       . which the channel monitors. *The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

4. 2 Channel Check CHANNEL CHECK is defined as the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.

4. 3 Channel Functional Test A CHANNEL FUNCTIONAL TEST is defined as:

4.3.1 Analog Channels The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions. 4.3.2 Bistable Channels The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

4. 4 Dose Equivalent 1-131 DOSE EQUNALENT I-131 is defined as that concentration of I-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, 1-134 and I-135 actually present The thyroid dose conversion factors used for

this calculation shall be those listed in Table m ofTID-14844, Calculation of Distance Factors for Power and Test Reactor Sites. Surry's definition ofOOSE EQUNALENT I-131 allows use of thyroid dose conversion factors from NRC Regulatory Guide 1.109, Revision 1.

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4. 5 Frequency Notations NOTE: Frequencies are allowed a maximum extension of 25%.

Frequency notations are defined as follows: NOTATION FREQUENCY D-Daily At least once per 24 hours W-Weekly At least once per 7 days M-Monthly At least once per 31 days release Q - Quarterly At least once per 92 days SA - Semi-annually At least once per 184 days R- Refueling At least once per 18 months SIU - Startup Prior to each reactor startup P - Prior to release Completed prior to each release N.A. - Not applicable Not applicable

4. 6 Gaseous Radwaste Treatment System A GASEOUS RADWASTE TREATMENT SYSTEM is the 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. North Anna's Technical Specifications define system composition as the waste gas decay tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tanks and waste gas diaphragm compressor.

4. 7 General Nomenclature X = Chi: concentration at a point at a given instant (curies per cubic meter)

D = Deposition: quantity of deposited radioactive material per unit area (curies per square meter) Q = Source strength (instantaneous; grams, curies, etc.)

            =   Emission rate (continuous; grams per second, curies per second, etc.)
            =   Emission rate (continuous line source; grams per second per meter, etc.)

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4. 8 Member of the Public MEMBER OF TIIE PUBLIC shall include individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

4. 9 Operable - Operability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functions, and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system,

subsystem, train, component, or device to perform its functions are also capable or performing their related support functions.
4 .10 Purge - Purging PURGE or PURGING is defined as 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.

4 .11 Rated Thermal Power RA1ED THERMAL POWER shall be a total reactor core heat transfer rate to reactor coolant of:

Surry: 2441 Megawatt Thermal (MWt)
North Anna: 2893 MWt 4 .12 Site Boundary The SIIB BOUNDARY is defined as that line beyond which the land is not owned, leased, or otherwise controlled by Virginia Power.

4 .13 Source Check A SOURCE CHECK is defined as the qualitative assessment of channel response when the channel sensor is exposed to radiation. This applies to installed radiation monitoring systems. L

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4.14 Special Report A report submitted to the NRC in accordance with Technical Specification requirements:

(Surry Technical Specification 6.2) (North Anna Technical Specification 6.9.2) 4 .15 Thermal Power 1HERMAL POWER is defined as the total reactor core heat transfer rate to the reactor coolant. 4 .16 Unrestricted Area UNRESTRICTED AREA is defined as any area at or beyond the SITE BOUNDARY where access is not controlled by Virginia Power for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional or recreational purposes. 4 .17 Ventilation Exhaust Treatment System VENTILATION EXHAUST TREATMENT SYSTEM is defined as the 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 absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the

5.0 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.

RESPONSIBILITIES 5.1 Health Physics Health Physics is responsible for: 5.1.1 Establishing and maintaining necessary procedures for sampling and monitoring radioactive effluents and the environment 5 .1.2 Performing and documenting surveys, sampling, and analyses of plant effluents and environmental monitoring. 5.1.3 Performing trend analysis on plant effluents and recommending actions to correct adverse trends. 5 .1.4 Preparing Effluent and Environmental Monitoring Program records. 5.2 Operations Department

The Operations Department is responsible for requesting samples, analysis, and authorization to release effluents.

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6.0 INSTRUCTIONS NOTE: Meteorological, liquid and gaseous pathway analyses are presented in Attachments 28 and 29, Meteorological, Liquid and Gaseous Pathway Analysis (Surry and North Anna, respectively).

6. 1 Sampling and Monitoring Criteria 6.1.1 Surveys, sampling, and analyses shall be performed with instruments calibrated for the type and range of radiation monitored and the nature of the discharge monitored.

6.1.2 Installed monitoring systems shall be calibrated for the type and range of radiation or parameter monitored 6.1.3 A sufficient number of survey points or samples shall be taken to adequately assess the status of the discharge monitored. 6.1.4 Samples shall be representative of the volume and nature of the monitored discharge. 6.1.5 Surveys, sampling, analyses, and monitoring records shall be accurately and legibly documented and sufficiently detailed so that the meaning and intent is clear. 6.1.6 Surveys, analyses, and monitoring records shall be reviewed for trends, completeness, and accuracy.

6. 2 Liquid Radioactive Waste Effluents 6.2.1 Liquid Effluent Concentration Limitations

a. Liquid waste concentrations from the site will not exceed the following applicable limits:

1. For radionuclides (other than dissolved or entrained noble gases) the concentration released in liquid effluents to UNRESTRICTED AREAS shall be limited to those specified in 10 CFR 20, Appendix B, Table II, Column 2.

2 For dissolved or entrained noble gases, the concentration shall be limited to 2E-4 µCi/ml.

b. If the concentration of liquid effluents released from the site exceed the above

limits, promptly restore concentrations to within limits .

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c. Daily concentrations of radioactive materials in liquid waste to UNRESTRICfED AREAS shall meet the following limitation:

Volume of Waste Discharged+ Volume of Dilution Water > 1

                                                                     ~       Ci/ml*          -

Volume of Waste Discharged x ,£...i µMPC/ where:

               µCi/m4     = the concentration of nuclide i in the liquid effluent discharge; MPCi       = the maximum permissible concentration in UNRESTRICTED AREAS of nuclide, i, expressed as µCi/ml from 10CFR Part 20, Appendix B, Table II, for radionuclides other than noble gases and 2E-04 µCi/ml for dissolved or entrained noble gases.

6.2.2 Liquid Monitoring Instrumentation

a. Radioactive liquid effluent monitoring instrumentation channels shown on Attachments 1 and 2, Radioactive Liquid Effluent Monitoring Instrumentation (Surry and North Anna, respectively), shall be OPERABLE with their alarm/trip

setpoints set to ensure that limits of step 6.2.1.a are not exceeded.

1 . Alarm/trip setpoints of these channels shall be determined and adjusted in accordance with step 6.2.2.d, Setpoint Calculation.

2. If a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint is less conservative than required by step 6.2,2.a, perform one of the following:

Promptly suspend release of radioactive liquid effluents monitored by affected channel
Declare the channel inoperable
Change the setpoint to an acceptable conservative value

b. Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Attachments 3 and 4, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements (Surry and North

Anna, respectively).

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1. With the number of channels OPERABLE less than the minimum channels required by tables shown in Attachment 1 and 2, perform the ACTION shown in these tables.

2. Attempt to return the instruments to OPERABLE status within 30 days. If unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. Applicable Monitors Liquid effluent monitors for which alann/trip setpoints are determined are:

1. Surry Release Point Instrument Number Liquid Radwaste Effluent Line LW-108 Service Water System Effluent Line SW-107 Circulating Water Discharge Line SW-120, SW-220

2. North Anna Release Point Instrument Number Liquid Radwaste Effluent Line LW-111 Service Water System Effluent Line SW-108 Condenser Circulating Water SW-130, SW-230

d. Setpoint Calculation NOTE: This methodology does not preclude the determination of more conservative setpoints.

1. Maximum setpoint values shall be calculated using the following equation:

CF C =f where: c = the setpoint, in µCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution; C = the effluent concentration limit for this monitor used in implementing 10 CFR 20 for the Station, in µCi/ml; f = the flow setpoint as measured at the radiation monitor location, GPM;

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F = the dilution water flow calculated as:

(Surry) F = f + (200,000 GPM x Number of Circ. Pumps in Service) (N. Anna) F = f + (218,000 GPM x Number of Circ. Pumps in Service)

2. Eachofthecondensercirculatingwaterchannels (Surry: SW-120, SW-220)

(North Anna: SW-130, SW-230) monitors the effluent (service water including component cooling service water, circulating water, and liquid radwaste) in the circulating water discharge tunnel beyond the last point of possible radioactive material addition. No dilution is assumed for this pathway. Therefore, the equation in step 1 above becomes: c=C The setpoint for Station monitors used in implementing 10 CFR 20 for the site becomes the effluent concentration limit.

3. In addition, for added conservatism, setpoints are calculated for the liquid radwaste effluent line (Surry: LW-108, North Anna: LW-111) and the component cooling service water system effluent line (Surry: SW-107, North Anna: SW-108).

For the liquid radwaste effluent line, the equation in step 1 becomes: CFKLw C = f where; KLw = The fraction of the effluent concentration limit used in implementing 10CFR20 for the site attributable to liquid radwaste effluent line pathway. For the service water system effluent line, the equation in step 1 becomes: CFKsw C = f where; Ksw = The fraction of the effluent concentration limit used in implementing 10 CFR 20 for the Station attributable to the service water effluent line pathway. The sum KLw + Ksw ~ 1.0.

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6.2.3 Liquid Effluent Dose Limit

a. Requirement At least once per 31 days, perform the dose calculation in subsections 6.2.3.c and 6.2.3.d to ensure that the dose or dose commitment to the maximum exposed MEMBER OF 1HE PUBLIC from radioactive materials in liquid releases (from each reactor unit) to UNRESTRICTED AREAS shall be limited to the following:

1. During any calendar quarter to:

Less than or equal to 1.5 mrem to the total body
Less than or equal to 5 mrem to the critical organ
2. During any calendar year to:
Less than or equal to 3 mrem to the total body
Less than or equal to 10 mrem to the critical organ

b. Action

If the calculated dose from release of radioactive materials in liquid effluents exceeds any of the above limits, prepare and submit to the Commission within 30 days, a Special Report that identifies causes for exceeding limits and defines corrective actions taken to reduce releases of radioactive materials in liquid effluents to ensure that subsequent releases will be in compliance with the above limits.

c. Surry Dose Contribution Calculations NOTE Thyroid and GI-LLI organ doses must be calculated to determine which is the critical organ for the period being considered.

Dose contributions shall be calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS based on the following expression:

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where:

D = the cumulative dose commitment to the total body or critical organ, from the liquid effluents for the time period t, in mrem; t = the length of the time period over which q and F are averaged for all liquid releases, hours; M = the mixing ratio (reciprocal of the dilution factor) at the point of exposure, dimensionless, 0.2 from Appendix 1 lA, Surry UFSAR; F = the near field average dilution factor for q during any liquid effluent release. Defined as the ratio of the average undiluted liquid waste flow during release to the average flow from the site discharge structure to UNRESTRICTED AREAS; q = the average concentration of radionuclide, i, in undiluted liquid effluent during time period, t, from any liquid releases, in µCi/ml;

          ~  = the site related ingestion dose commitment factor to the total body or critical organ of an adult for each identified principal gamma and beta emitter in mrem-ml per hr-µCi. Values for Ai are given in Attachment 5, Liquid Ingestion Pathway Dose Factors For Surry Power Station.

Ai= 1.14 E+05 (21BFi + 5Bli) DFi where: 1.14 E+o5 = 1 E+o6 pCi/µCi x 1 E+o3 m]/kg + 8760 hr/yr, units conversion factor; 21 = adult fish consumption, kg/yr, from NUREG-0133; 5 = adult invertebrate consumption, Kg/yr, from NUREG-0133; Bli = the bioaccumulation factor for nuclide, i, in invertebrates, pCi/kg per pCi/1, from Table A-1 of Regulatory Guide 1.109, Rev. 1; BFi = the bioaccumulation factor for nuclide, i, in fish, pCi/kg per pCi/1, from Table A-1 of Regulatory Guide 1.109, Rev. 1. DFi = the critical organ dose conversion factor for nuclide, i, for adults, in mrem/pCi, from Table E-11 of Regulatory Guide 1.109, Rev. 1.

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d. North Anna Dose Contribution Calculations NOTE: North Anna's dose contribution calculation for liquid effluents released to UNRESTRICTED AREAS has been modified. The derivation is given in Attachment 6, North Anna Liquid Ingestion Pathway Dose Factor Calculation.

Dose contribution shall be calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS based on the following expressions: D = LQi.XBi i Where: D = the cumulative dose commitment to the total body or critical organ, from the liquid effluents for the time period t, in mrem; Bi . . . Dose Commitment Factors (mrem/Ci) for adults. Values for Bi are given in Attachment 7, North Anna Liquid Ingestion Pathway Dose Commitment Factors for Adults. Q = Total released activity for the considered time period and the ith nuclide. Q = t x Ci x Waste Flow Where: t = the length of the time period over which q and F are averaged for all liquid releases, hours; Ci = the average concentration of radionuclide, i, in undiluted liquid effluent during time period, t, from any liquid releases, in µCi/ml;

e. Quarterly Composite Analyses For radionuclides not determined in each batch or weekly composite, dose contribution to current monthly or calendar quarter cumulative summation may be approximated by assuming an average monthly concentration based on previous monthly or quarterly composite analyses. However, for reporting purposes, calculated dose contribution shall be based on the actual composite analyses.

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6.2.4 Liquid Radwaste Treatment

a. Requirement 1 . The Liquid Radwaste Treatment System shall be used to reduce the radioactive materials in liquid waste prior to discharge when projected dose due to liquid effluent, from each reactor unit, to UNRESTRICI'ED AREAS would exceed 0.06 mrem to total body or 0.2 mrem to the critical organ in a 31 day period.

2. Doses due to liquid releases shall be projected at least once per 31 days.

b. Action If radioactive liquid waste is discharged without treatment and in excess of the above limits, within 30 days, prepare and submit to the Commission, a Special Report that includes the following information:

1 . Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-system, and the reason for the inoperability .

2. Actions taken to restore inoperable equipment to OPERABLE status.

3. Summary description of actions taken to prevent a recurrence.

c. Projected Total Body Dose Calculation

1. Determine Drn = total body dose from liquid effluents in the previous 31 day period, calculated according to subsection 6.2.3.c or d (Surry and North Anna, respectively).

2. Estimate R 1 =ratio of the estimated volume of liquid effluent releases in the present 31 day period to the volume released in the previous 31 day period.

3. Estimate F 1 =ratio of the estimated liquid effluent radioactivity in the present 31 day period to liquid effluent activity in the previous 31 day period (µCi/ml).

4. Determine PDrn = projected total body dose in a 31 day period.

PDrn = °'rB (R1F1)

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d. Projected Critical Organ Dose Calculation NOTE: Historical data pertaining to the volumes and radioactivity of liquid effluents released in connection with specific Station functions, such as maintenance or refueling outages, shall be used in projections as appropriate.

1. Determine D 0 = critical organ dose from liquid effluents in the previous 31 day period, calculated according to subs~tion 6.2.3.c or d (Surry and North Anna, respectively).

2. Estimate R 1 as in step 6.2.4.c.2.

3 .. Estimate F 1 as in step 6.2.4.c.3.

4. Determine PD0 =projected critical organ dose in a 31 day period.

PD = D (R1F1) 0 0 6.2.5 Liquid Sampling Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis requirements shown in Attachments 8 and 9, Radioactive Liquid Waste Sampling and Analysis Program (Surry and North Anna, respectively). 6.3 Gaseous Radioactive Waste Effluents

6. 3 .1 Gaseous Effluent Dose Rate Limitation

a. Requirement Dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY are limited to the following:

1. The dose rate limit for noble gases shall be .:5 500 mrem/year to the total body and .:5 3000 mrern/year to the skin.

2. The dose rate limit for 1-131, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days shall be .:5 1500 mrem/year to the critical organ.

b. Action

1. If the dose rates exceed the above limits, promptly decrease the release rate to

within the above limits.

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2. Dose rates due to noble gases in gaseous effluents shall be determined continuously to be Within the limits specified in subsection 6.3.1.a.

3. Dose rates due to 1-131, 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 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified on Attachments 10 and 11, Radioactive Gaseous Waste Sampling and Analysis Program (Surry and North Anna, respectively).

c. Calculations of Gaseous Effluent Dose Rates

1. The dose rate limit for noble gases shall be determined to be within the limit by limiting the release rate to the lessor of:

                *   ~ [Kivv Qivv + Kipv Oipv] ~ 500 mrem/yr to the total body; l

or,

L [(Livv + l.lMivv) Oivv + CLipv + l.lMipv) Oipv]

i

                                                                      ~ 3000 mrem/yr to the skin.

where: Subscripts= vv, refers to vent releases from the building ventilation vent; pv, refers to the vent releases from the process vent; i, refers to individual radionuclide; Kivv, Kipv = The total body dose factor for ventilation vent or process vent release due to gamma emissions for each identified noble gas radionuclide, i, in mrem/yr per Curie/sec. Factors are listed in Attachments 12 and 13, Gaseous Effluent Dose Factors (Surry and North Anna, respectively). Livv* Lipv = The skin dose factor for ventilation vent or process vent release due to beta emissions for each identified noble gas radionuclide i, in mrem/yr per Curie/sec. Factors are listed in Attachments 12 and 13. Mivv, Mipv = The air dose factor for ventilation vent or process vent release due to gamma emissions for each identified noble gas radionuclide, i, in mrad/yr per Curie/sec. Factors are listed in Attachments 12 and 13.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 21 OF 116 Qvv* Qipv = The release rate for ventilation vent or process vent of noble gas radionuclide, i, in gaseous effluents in Curie/sec (per site); 1.1 = The unit conversion factor that converts air dose to skin dose, in mrem/mrad.

2. The dose rate limit for 1-131, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days shall be determined to be within the limit by restricting the release rate to:

i [Pivv Qivv + Pipv Qpv] ~ 1500 mrem/yr to the critical organ. 1 where: Pivv* Pipv = The critical organ dose factor for ventilation vent or process vent for 1-131, H-3, and all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway, in mrem/yr per Curie/sec. Factors are listed in Attachments 12 and 13.

Qvv* Qipv = The release rate for ventilation vent or process vent of 1-131, H-3, and all radionuclides, i, in particulate form with half-lives greater than 8 days in gaseous effluents in Curie/sec (per site).

3. All gaseous releases, not through the process vent, are considered ground level and shall be included in the determination of Qivv*

6.3.2 Gaseous Monitoring Instrumentation

a. Requirement

1. The radioactive gaseous effluent monitoring instrumentation channels shown in Attachments 14 and 15, Radioactive Gaseous Effluent Monitoring Instrumentation (Surry and North Anna, respectively), shall be OPERABLE with alarm/trip setpoints set to ensure that limits specified for noble gases in subsection 6.3.1.a are not exceeded. Alann/trip setpoints of these channels shall be determined and adjusted in accordance with subsection 6.3.2.d.

VIRGINIA VPAP-2103 POWER REVISIONO

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2. 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 Attachments 16 and 17, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements (Surry and North Anna, respectively).

b. Action

1. With a radioactive gaseous effluent monitoring instrumentation channel alann/trip setpoint less conservative than required by the above requirement, promptly:

Suspend the release of radioactive gaseous effluents monitored by the affected channel; mld
Declare the channel inoperable;m:
Change the setpoint so it is acceptably conservative
2. With the number of channels OPERABLE less than the minimum channels required by tables shown in Attachment 14 and 15, take the ACTION shown in these tables.

3. Return the instruments to OPERABLE status within 30 days. If unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. Applicable Monitors Radioactive gaseous effluent monitors for which alann/trip setpoints are determined are:

1. Surry Release Point Instrument Number Process Vent GW-102, GW-130-1 Condenser Air Ejector SV-111, SV-211 Ventilation Vent VG-110, VG-131-1

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 23 OF 116

2. North Anna Release Point Instrument Number Process Vent GW-102, GW-180-1 Condenser Air Ejector SV-121, SV-221 Ventilation Vent A VG-104, VG-178-1 Ventilation VentB VG-113, VG-179-1

d. Setpoint Calculations

1. The setpoint calculations for each monitor listed above shall be determined such that the following relationship is maintained:

D ~ Dpv +Dcae +Dvv where: D = Subsection 6.3.1.a dose limits implementing 10 CFR 20 for the Station, mrem/yr; Dpv = The noble gas Station boundary dose rate from process vent gaseous

D~

Dvv effluent releases, rnrem/yr;

                        = The noble gas Station boundary dose rate from condenser air ejector gaseous effluent releases, mrem/yr; *
                        = The noble gas Station boundary dose rate from:
                           ~:             Ventilation vent gaseous effluent releases, mrem/yr North Anna: Summation of ventilation vent A plus B gaseous effluent releases, mrem/yr

2. Setpoint values shall be detennined using the following equation:

Rm x 2.12 E-03 Cm Fm where: m = The release pathway, process vent (pv), ventilation vent (vv) or condenser air ejector (cae); Cm = The effluent concentration limit implementing subsection 6.3.1.a for the Station, µCi/ml; Rm = The release rate limit for pathway m determined from methodology in

subsection 6.3.1.c, using Xe-133 as nuclide to be released, µCi/sec; 2.12E-03 = CFM per ml/sec;

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 24 OF 116 Fm =The maximum flow rate for pathway m, CFM.

3. According to NUREG-0133, the radioactive effluent radiation monitor alann/trip setpoints should be based on the radioactive noble gases. It is not considered to be practicable to apply instantaneous alann/ trip setpoints to integrating monitors sensitive to radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases.

6.3.3 Noble Gas Effluent Air Dose Limit

a. Requirement

1. The air dose in UNRESTRICI'ED AREAS due to noble gases released in gaseous effluents from each reactor unit from the site at and beyond the SITE BOUNDARY shall be limited to the following:

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.
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.

2. Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with subsection 6.3.3.c, Dose Calculations, at least once per 31 days.

b. Action If the calculated air dose from radioactive noble gases in gaseous effluents exceeds any of the above limits, prepare and submit to the Commission within 30 days, a Special Report that identifies the causes for exceeding the limits and defines corrective actions that have been taken to reduce releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the limits stated in subsection 6.3.3.a.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 25 OF 116

c. Noble Gas Effluent Air Dose Calculation NOTE: Gaseous releases, not through the process vent, are considered ground level and shall be included in the detennination of Q.vv.

1. The air dose to areas at or beyond the SITE BOUNDARY due to noble gases shall be detennined by the following:

For gamma radiation: Dg = 3.17E-08 i CMivv Qivv + Mipv Qipvl l For beta radiation: Db = 3.17E-08 i [Nivv Qivv + Nipv Qipvl l Where: Subscripts = vv, refers to vent releases from the building ventilation vent.

Dg Db

pv, refers to the vent releases from the process vent i, refers to individual radionuclide the air dose for gamma radiation, in mrad

                                 = the air dose for beta radiation, in rnrad; Mivv, Mipv = the air dose factors for ventilation vent or process vent release due to gamma emissions for each identified noble gas radionuclide, i, in mrad/yr per Curie/sec. Factors are given in Attachments 12 and 13.

Nivv, Nipv = the air dose factor for ventilation vent or process vent release due to beta emissions for each identified noble gas radionuclide, i, in rnrad/yr per Curie/sec. Factors are listed in Attachments 12 and 13.

                   <2i.vv, Qipv  = the release for ventilation vent or process vent of noble gas radionuclide, i, in gaseous effluents for 31 days, quarter, or year as appropriate in Curie (per site);

VIRGINIA VPAP-2103 POWER REVISIONO

6.3.4 1-131, H-3, and Radionuclides In Particulate Form Effluent Dose Limit

a. Requirement PAGE 26 OF 116

1. Methods shall be implemented to ensure that the dose to any organ of a MEMBER OF THE PUBLIC from 1-131, tritium,-and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to UNRESTRICIED AREAS from each reactor unit shall be limited to the following:

During any calendar quarter, to 5 7.5 mrem to the critical organ
During any calendar year, to 5 15 mrem to the critical organ.

2. Cumulative dose contributions to a MEMBER OF TIIB PUBLIC from 1-131, tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to UNRESTRICTED AREAS for the current calendar quarter and current calendar year shall be determined in accordance with subsection 6.3.4.c, Surry Dose Calculations, or subsection 6.3.4.d, North Anna Dose Calculations, at least once per 31 days.

b. Action If the calculated dose from the release of 1-131, 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 containing the following:

I . Causes for exceeding limits.

2. Corrective actions taken to reduce releases.

3. Proposed corrective actions to be taken to assure that subsequent releases will be in compliance with limits stated in subsection 6.3.4.a .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 27 OF 116

c. Surry Dose Calculations NOTE: Gaseous releases, notthrough process vent, are considered ground level and shall be included in the determination of Qvv*

1. The dose to the maximum exposed MEMBER OF THE PUBLIC from 1-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be determined as follows:

Dr= 3.17E-08 l: [CRMivv <'.1vv + RMipv Qpv) + (Rlivv Qivv + Rlipv Qipv)] 1 Where: Subscripts = vv, refers to vent releases from the building ventilation vent; pv, refers to the vent releases from the process vent; Dr = the dose to the critical organ of the maximum exposed

MEMBER OF THE PUBLIC in mrem.

RMivv, RMipv = the milk pathway dose factor for ventilation vent or process vent release due to 1-131, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days, in mrem/yr per Curie/sec. Factors are listed in Attachment 18, Critical Organ and Inhalation Dose Factors For Surry. Rlivv, Rlipv = the inhalation pathway dose factor for ventilation vent or process vent release due to 1-131, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days, in mrem/yr per Curie/sec. Factors are listed in Attachment 18. Oivv, Oipv = the release for ventilation vent or process vent of 1-131, tritium, and from all radionuclides in particulate from with half-lives greater than 8 days in Curies (per site). 3.17 E-08 = the inverse of the number of seconds in a year.

 . VIRGINIA                                                                                  VPAP-2103 POWER                                                                                REVISIONO PAGE 28 OF 116

d. North Anna Dose Calculations NOTE: Gaseous releases, not through process vent, are considered ground level and shall be included in the determination of Oivv*

1. The dose to the maximum exposed MEMBER OF THE PUBLIC from 1-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be determined as follows:

Dr= 3.17E-08 l: [Rivv <2ivv + Ripv C2ipv] 1 Where: Subscripts = vv, refers to vent releases from the building ventilation vent; pv, refers to the vent releases from the process vent;

Dr = the dose to the critical organ of the maximum exposed MEMBER OF THE PUBLIC in mrem.

                                       = the dose factor for ventilation vent or process vent release due to 1-131, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days, in mrem/yr per Curie/sec. Factors are listed in Attachment 19, Critical Organ and Inhalation Dose Factors for North Anna.

Oi.vv, Oipv = the release for ventilation vent or process vent of 1-131, tritium, and from all radionuclides in particulate from with half-lives greater than 8 days in Curies (per site). 3.17 E-08 = the inverse of the number of seconds in a year.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 29 OF 116 6.3.5 Gaseous Radwaste Treatment NOTE: Historical data pertaining to the volumes and radioactive concentrations of gaseous effluents released in connection to specific Station functions, such as containment purges, shall be used in the above estimates as appropriate.

a. Requirement

1. The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive material in gaseous waste prior to their discharge when projected gaseous effluent air doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation averaged over 31 days.

2. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to

reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY would exceed 0.3 mrem to the critical organ averaged over 31 days.

3. Doses due to gaseous releases from the site shall be projected at least once per 31 days based on calculations performed in subsections 6.3.5.c, d, and e.

b. Action With 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 that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reason for the inoperability.

2. Actions taken to restore the inoperable equipment to OPERABLE status.

3. Summary description of actions taken to prevent a recurrence.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 30 OF 116

c. Projected Gamma Dose

1. Determine Dg =the 31 day gamma air dose in the previous 31 day pericxl calculated according to subsection 6.3.3.c.

2. Estimate Rg =ratio of the estimated volume of gaseous effluent in the present 31 day pericxl to the volume released during the previous 31 day pericxl.

3. Estimate Fg =ratio of the estimated noble gas effluent activity in the present 31 day pericxl to the noble gas effluent activity during the previous 31 day period

(µCi/ml).

4. Determine PDg = projected 31 day gamma air dose:

PDg =Dg (Rg x Fg)

d. Projected Beta Dose

1. Determine Db = the 31 day beta air dose in the previous 31 day pericxl, calculated according to subsection 6.3.3.c .

2. Estimate Rg and Fg as in steps 6.3.5.c.2 and 3 above.

3. Determine PDg = projected 31 day pericxl beta air dose:

PDb = Db (Rg x Fg)

e. Projected Maximum Exposed Member of the Public Dose

1. Determine Dmax =the 31 day maximum exposed MEMBER OF THE PUBLIC dose in the previous 31 day period, calculated according to subsection 6.3.4.c.

2. Estimate Fi =ratio of the estimated activity from I-131, radioactive materials in particulate form with half-lives greater than 8 days, and tritium in the present 31 day period to the activity ofl-131, radioactive materials in particulate form with half-lives greater than 8 days, and tritium in the previous 31 day period

(µC/ml).

3. Determine PDmax =projected 31 day maximum exposed MEMBER OF TIIE PUBLIC dose:

PDmax =Dmax (Rg x Fi)

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 31 OF 116

6. 4 Total Dose Limit to Public From Uranium Fuel Cycle Sources 6.4.1 Requirement The annual (calender year) dose or dose commitment to the maximum exposed MEMBER OF THE PUBLIC due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or the critical organ (except the thyroid, which shall be limited to less than or equal to 75 mrems).

6.4.2 Action

a. If the calculated doses from release of radioactive materials in liquid or gaseous effluents exceed twice the limits of Subsections 6.2.3.a, 6.3.3.a, or 6.3.4.a, calculations shall be made, including direct radiation contribution from the reactor units and from outside storage tanks, to determine whether limits of 6.4.1 have been exceeded.

b. If the limits of 6.4.1 have been exceeded, 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 limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include the following:

1. An analysis that estimates the radiation exposure (dose).to the maximum exposed MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the releases covered by this report.

2. A description of the levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.

3. If the estimated doses exceeds the limits of 6.4.1, and if the release condition resulting in violation of 40 CPR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CPR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 32 OF 116 6; 5 Radiological Environmental Monitoring 6.5.1 Monitoring Program

a. Requirement

1. The Radiological Environmental Monitoring Program shall be conducted as specified in Attachments 20 and 21, Radiological Environmental Monitoring Program (Surry and North Anna, respectively).

2. Samples shall be collected from specific locations given in Attachments 22 and 23, Environmental Sample Locations (Surry and North Anna, respectively).

3. Samples shall be analyzed in accordance with:

Requirements of Attachments 20 and 21
Detection capabilities required by Attachments 24 and 25, Detection Capabilities for Environmental Sample Analysis (Surry and North Anna, respectively)
Guidance of the Radiological Assessment Branch Technical Position on
b. Action Environmental Monitoring dated November, 1979, Revision No. 1.

1. With the radiological environmental monitoring program not being conducted as required in 6.5.1.a, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification (Surry T.S. 6.6.B.2) (North Anna T.S. 6.9.1.8), a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 33 OF 116

2. H, when averaged over any calendar quarter, the level of radioactivity exceeds the reporting levels of Attachments 26 and 27, Reporting Levels for Radioactivity Concentrations in Environmental Samples (Surry and North Anna, respectively), prepare and submit to the Commission within 30 days, a Special Report that:

Identifies the causes for exceeding the limits; 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 subsection 6.2.3, 6.3.3, and 6.3.4.

When more than one of the radionuclides in Attachments 26 and 27 are detected in the sampling medium, this report shall be submitted if: concentration (1) concentration (2) reporting level (1) + reporting level (2) + *** ~ l.O

3. When radionuclides other than those listed in Attachment 26 and 27 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 subsections 6.2.3, 6.3.3, and 6.3.4. 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.

4. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Attachment 20 and 21, 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.

Identify the cause of the unavailability of samples and identify the new locations for obtaining replacement samples in the next Semi-annual Radioactive Effluent Release Report. Include in the report a revised figure and table for the ODCM reflecting the new locations .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 34 OF 116 6.5.2

Land Use Census

a. Requirement 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 following:

Nearest milk animal
Nearest residence
Nearest garden of greater than 50 m2 (500 ft2) producing broad leaf vegetation

1. The land use census shall be conducted during the growing season at least once per 12 months using that information which will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. Results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

2. 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 given in Attachments 20 and 21 shall be followed, including analysis of control samples.

b. Action

1. With a land use census identifying locations that yield a calculated dose or dose commitment greater than the values currently being calculated in step 6.3.4.a.2, identify the new locations in the next Semiannual Radioactive Effluent Release Report.

2. With a land use census identifying locations that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained, add the new locations to the Radiological Environmental Monitoring Program within 30 days. The sampling locations, excluding the control station location, having the lowest calculated dose or dose commitments (via the same exposure pathway) may be deleted from the monitoring program after October 31 of the year in which this land use census was conducted. Identify the new locations in the next Semiannual Radioactive Effluent Release Report and also include in the report revised figures and tables reflecting the new locations.

VIRGINIA VPAP-2103 POWER REVISIONO

6.5.3 Interlaboratory Comparison Program

a. Requirement PAGE 35 OF 116 Analyses shall be performed on radioactive materials (which contain nuclides produced at nuclear power stations) supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission.

b. Action

1. Analyses shall be performed as part of the Environmental Protection Agency's Environmental Radioactivity Laboratory Intercomparison Studies (Cross Check) Program and include:

Program Cross-Check Of: Mille 1-131, Gamma, K, Sr-89 and 90 Water Gross Beta, Gamma, 1-131, H-3 (Tritium), Sr-89/90, Blind - any combinations of above radionuclides. Air Filter Gross Beta, Gamma, Sr-90

2. 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.

c. Methodology and Results

1. Methodology and results of the cross-check program shall be maintained in the contractor supplied Nuclear Reactor Environmental Radiation Monitoring Quality Control Manual, IWL-0032-361.

2. Results will be reported in the Annual Radiological Environmental Monitoring Report.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 36 OF 116 6.6 REPORTING REQUIREMENTS 6.6.1 Annual Radiological Environmental Operating Report

Routine Radiological Environmental Operating Reports covering the operation of the units during the previous calendar year shall be submitted prior to May 1 of each year.

A single submittal may be made for the Station. Radiological Environmental Operating Reports shall include:

a. Summaries, interpretations, and analysis of trends of results of radiological environmental surveillance activities for the report pericx:l, including:

A comparison (as appropriate) with preoperational studies, operational controls, and previous environmental surveillance reports
An assessment of the observed impacts of the plant operation on the environment
Results ofland use census per subsection 6.5.2, Land Use Census

b. Results of analysis of radiological environmental samples and of environmental

radiation measurements taken per subsection 6.5.1, Monitoring Program. Results shall be summarized and tabulated in the format of the table in the Radiological Assessment Branch Technical Position (Reference 3.1.11).

1. If some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining reasons for missing results.

2. Missing data shall be submitted as soon as possible in a supplementary report.

c. A summary description of the radiological environmental monitoring program.

d. At least two legible maps covering sampling locations keyed to a table giving distances and directions from the centerline of one reactor. One map shall cover stations near the SITE BOUNDARY; a second shall include more distant stations.

e. Results of Station's participation in the Interlaboratory Comparison Program; per Subsection 6.5.3, Interlaboratory Comparison Program.

f. Discussion of deviations from the Station's environmental sampling schedule per Attachment 20 or 21 (as appropriate) .

g. Discussion of analyses in which the lower limit of detection (LLD) required by Attachment 24 or 25 (as appropriate) was not achievable.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 37 OF 116 6.6.2 Semiannual Radioactive Effluent Release Report

a. Requirement Radioactive Effluent Release Reports covering operation of the units during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. A single submittal may be made for the Station and should combine those sections that are common to both units. Radioactive Effluent Release Reports shall include:

1. A summary of quantities of radioactive liquid and gaseous effluents and solid waste released. Data shall be summarized on a quarterly basis following the format of Regulatory Guide 1.21, Appendix B (Reference 3.1.5).

2. An assessment of the radiation doses to the maximum exposed :MEMBERS OF THE PUBLIC due to the radioactive liquid and gaseous effluents released from the Station during the previous calendar year. This assessment shall be performed in accordance with subsection 6.6.2.b, Dose Assessment, and shall

only be included in Radioactive Effluent Release Reports submitted within 60 days after January 1 of each year.

3. A list of unplanned releases from the site to UNRESTRICTED AREAS occurring during the reporting period that exceed the limits set forth in subsections 6.2.1, Liquid Effluent Concentration Limitations, and 6.3.1, Gaseous Effluent Dose Rate Limitation.

4. Major changes made during the reporting period to radioactive liquid, gaseous, and solid waste treatment systems.

5. Changes made to VPAP-2103, Offsite Dose Calculation Manual (see subsection 6.6.4, Changes to the ODCM).

6. A listing of new locations for dose calculations or environmental monitoring identified by the Land Use Census (Subsection 6.5.2).

b. Dose Assessment

1. Radiation doses to individuals due to radioactive liquid and gaseous effluents from the Station during the previous calendar year shall either be calculated in

accordance with this procedure or in accordance with Regulatory Guide 1.109.

Population doses shall not be included in dose assessments.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 38 OF 116

2. The dose to the maximum exposed MEMBER OF TIIE PUBLIC due to the radioactive liquid and gaseous effluents from the Station shall be incorporated with the dose assessment performed above. If the dose to the maximum exposed MEMBER OF TIIE PUBLIC exceeds twice the limits of Subsections 6.2.3.a.l, 6.2.3.a.2, 6.3.3.a.1, or 6.3.4.a.l, the dose assessment shall include the contribution from direct radiation.

NOTE: NUREG-0543 (Reference 3.1.13), states "There is reasonable assurance that sites with up to four operating reactors that have releases within Appendix I design objective values are also in conformance with the EPA Uranium Fuel Cycle Standard, 40 CFR Part 190".

3. The meteorological conditions during the previous calendar year or historical annual average atmospheric dispersion conditions shall be used for determining the gaseous pathway doses.

6.6.3 Annual Meteorological Data

a. Meteorological data collected over the previous year shall be in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.

b. Meteorological data shall be retained in a file on site and shall be made available to the NRC upon request.

6.6.4 Changes to the ODCM Changes to the ODCM shall be:

a. Reviewed and approved by Station Nuclear Safety and Operating Committee (SNSOC) prior to implementation.

b. Documented and records of reviews performed shall be retained as Station records.

Documentation shall include:

1. Sufficient information to support the change together with appropriate analyses or evaluations justifying changes .

VIRGINIA VPAP-2103 POWER REVISIONO

                                                                                          *PAGE 39 OF 116

2. A determination that the change will not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations and will maintain the level of radioactive effluent control required by:
10 CPR 20.106
40 CPR Part 190
10 CPR 50.36a
10 CPR Part 50, Appendix I

c. Submitted to the NRC in the fonn of a complete legible copy of the entire ODCM as a part of, or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.,

month/year) the change was implemented. 7.0 RECORDS The following individuaVpackaged documents and related correspondence completed as a result of the performance or implementation of this procedure are records. Records shall be transmitted to Records Management in accordance with VPAP-1701, Records Management.

These records shall include, but are not be limited to, the following:
Records of changes to the ODCM in accordance with subsection 6.6.4
Records of meteorological data in accordance with subsection 6.6.3
Records of sampling and analyses
Records of radioactive materials and other effluents released to the environment
Records of maintenance, surveillances, and calibrations

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 40 OF 116 ATTACHMENT 1 (Page 1 of 1) SURRY RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT CHANNELS ACTION OPERABLE

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE (a) Liquid Radwaste Effluent Line 1 1

2. GROSS BETA OR GAMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (a) Circulating Water Discharge Line 1 2 (b) Component Cooling Service Water Effluent Line 1 2

3. FLOW RATE MEASUREMENT DEVICES (a) Liquid Radwaste EffluentLine 1 3 ACTION 1: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases shall be suspended.

ACTION 2: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours, grab samples are collected and analyzed for principal gamma emitters, as defined in Attachment 8, Surry Radioactive Liquid Waste Sampling and Analysis Program. ACTION 3: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway shall be suspended.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 41 OF 116 ATTACHMENT 2 (Page 1 of 2) NORTH ANNA RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT CHANNELS ACTION OPERABLE

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE (a) Liquid Radwaste Effluent Line 1 1

2. GROSS BETA OR GAMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (a) Service Water System Effluent Line 1 1 (b) Circulating Water System Effluent Line 1 4

3. FLOW RATE MEASUREMENT DEVICES (a) Liquid Rad waste Effluent Line I 2

4. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR (a) Clarifier Effluent Line I 1

5. TANK LEVEL INDICATING DEVICES (Note I)

(a) Refueling Water Storage Tanks 1 3 (b) Casing Cooling Storage Tanks 1 3 (c) PC Water Storage Tanks (Note 2) 1 3 (d) Boron Recovery Test Tanks (Note 2) 1 3

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 42 OF 116 ATTACHMENT 2 (Page 2of2) NORTH *ANNA RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION ACTION 1: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least lxl0-7 µCi/g or an isotopic radioactivity at a lower limit of detection of at least 5x 10-7 µCi/g. ACTION 2: 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 hours during actual releases. Design capacity performance curves generated in situ may be used to estimate flow. ACTION 3: With the number of channels OPERABLE less than required oy the minimum channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank. ACTION 4: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, make repairs as soon as possible. Grab samples cannot be obtained via this pathway. NOTE 1: Tanks included in this requirement are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. NOTE 2: This is a shared system with Unit 2 .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 43 OF 116

ATTACHMENT 3 (Page 1 of 1)

SURRY RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTIONAL TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC lERMINATION OF RELEASE (a) Liquid Rad waste Effluent Line D PR R Q

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC lERMINATION OF RELEASE (a) Circulating Water Discharge Line D M R Q (b) Component Cooling Service Water D M R Q System Effluent Line

3. FLOW RAlE MEASUREMENT DEVICES (a) Liquid Radwaste Effluent Line D N.A. R N.A.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 44 OF 116

ATTACHMENT 4 (Page 1 of 2)

NORTH ANNA RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTIONAL TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC 1ERMINATION OF RELEASE (a) Liquid Radwaste Effluent Line D D R Q (Note 1)

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC 1ERMINATION OF RELEASE (a) Service Water System Effluent Line D M R Q (Note*2)

(b) Circulating Water System Effluent Line D M R Q (Note 2)

3. FLOW RA1E MEASUREMENT DEVICES (a) Liquid Radwaste Effluent Line D (Note 3) N.A. R Q

4. CONTINUOUS COMPOSI1E SAMPLERS AND SAMPLER FLOW MONITOR (a) Clarifier Effluent Line N.A. N.A. R N.A.

5. TANK LEVEL INDICATING DEVICES (Note6)

(a) Refueling Water Storage Tanlc D (Note4) N.A. R Q (b) Casing Cooling Storage Tanlc D (Note4) N.A. R Q (c) PC Water Storage Tanks (Note*S) D(Note4 N.A. R Q (d) Boron Recovery Test Tanlcs (Note 5) D (Note4) N.A. R Q

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 45 OF 116 ATTACHMENT 4 (Page 2of2) NORTH ANNA RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTE 1: The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the alann/trip setpoint.

b. Instrument controls not set in operate mode.

NOTE2: The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alann/trip setpoint.

b. Instrument controls not set in operate mode .

NOTE3:

NOTE4: CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be make at least once per 24 hours on days on which continuous, periodic, or batch releases are made. During liquid additions to the tank. NOTES: This is a shared system with Unit 2. NOTE6: Tanks included in this requirement are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 46 OF 116 ATTACHMENTS (Page 1 of 1) LIOUID INGESTION PATHWAY DOSE FACTORS FOR SURRY STATION UNITS 1 AND 2 Total Body A1 Thyroid A1 Gl*LLI A1 Radionuclide mrem/hr mrem{br mrem/hr

                         µCl/ml                     µCl/ml     µCl/ml H-3              2.82E-01                   2.82E-01   2.82E-01 Na-24            4.57E-01                   4.57E-01   4.57E-01 Cr-51             5.58E+OO                   3.34E-01   1.40E+03 Mn-54             1.35E+03                       -     2.16E+04 Fe-55            8.23E+03                       -     2.03E+04 Fe-59            7.27E+04                       -     6.32E+05 Co-58*            1.35E+03                       -      1.22E+04 Co-60            3.82E+03                       -     3.25E+04 Zn-65             2.32E+05                       -     3.23E+05 Rb-86            2.91E+02                       -      1.23E+02 Sr-89             1.43E+02                       -     8.00E+02 Sr-90             3.01E+04                       -     3.55E+03 Y-91              2.37E+OO                       -     4.89E+04 Zr-95            3.46E+OO                        -     1.62E+04 Zr-97             8.13E-02                       -     5.51E+04 Nb-95             1.34E+02                       -     1.51E+06 Mo-99             2.43E+01                       -     2.96E+02 Ru-103           4.60E+01                        -     1.25E+04 Ru-106           2.01E+02                        -     1.03E+05 Ag-110m          8.60E+02                        -     5.97E+05 Sb-124            1.09E+02                   6.70E-01  7.84E+03 Sb-125           4.20E+01                    1.79E-01  1.94E+03 Te-125m          2.91E+01                    6.52E+01  8.66E+02 Te-127m          6.68E+01

1.40E+02 1.84E+03 Te-129111 1.47E+02 3.20E+02 4.69E+03 Te-131m 5.71E+01 1.08E+02 6.80E+03 Te-132 1.24E+02 1.46E+02 6.24E+03 1-131 1.79E+02 1.02E+05 8.23E+01 1-132 9.96E+OO 9.96E+02 5.35E+OO 1-133 3.95E+01 1.90E+04 1.16E+02 1-134 5.40E+OO 2.62E+02 1.32E-02 1-135 2.24E+01 4.01E+03 6.87E+01 Cs-134 1.33E+04 - 2.85E+02 Cs-136 2.04E+03 - 3.21E+02 Cs-137 7.85E+03 - 2.32E+02 Cs-138 5.94E+OO - 5.12E-05 Ba-140 1.08E+02 - 3.38E+03 L.a-140 2.10E-01 - 5.83E+04 Ce-141 2.63E-01 - 8.86E+03 Ce-143 4.94E-02 - 1.67E+04 Ce-144 9.59E+OO - 6.04E+04 Np-239 1.91 E-03 - 7.11E+02

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 47 OF 116 ATTACHMENT 6 (Page 1 of 4) NORTH ANNA LIOUID INGESTION PATHWAY DOSE FACTOR CALCULATION UNITS 1 AND 2 1.0 EXPRESSION "1" D = t F LJi Ci Ai i where: D = the cumulative dose commitment to the total body or critical organ, from the liquid effluents for the time period t, in mrem; t = the length of time period over which Ci and F are averaged for all liquid releases, hours; F = the near field average dilution factor for q during any liquid effluent release. Defined as the ratio of the average undiluted liquid waste flow during release to the average flow

fi =

from the Station discharge structure to UNRESTRICIED AREAS; the individual dilution multiplication factor to account for increases in concentration of long-lived nuclides due to recirculation, listed on page 4 of 4 of this attachment. "fi" is the ratio of the total dilution flow over the effective dilution flow. Ci = the average concentration of radionuclide, i, in undiluted liquid effluent during time period, t, from any liquid releases, in µCi/ml; Ai = the site related ingestion dose commitment factor to the total body or critical organ of an adult for each identified principal gamma and beta emitter listed on page 4 of 4 of this attachment, in mrem-mi per hr-µCi; Ai = 1.14 E+o5 (730/Dw + 21BFi/Da} DFi where: 1.14 E+o5 = 1 E+o6 pCi/µCi x 1 E+o3 ml/kg+ 8760 hr/yr, units conversion factor; 730 = adult water consumption, kg/yr, from NUREG-0133;

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 48 OF 116 ATTACHMENT 6 (Page 2 of 4) NORTH ANNA LIOUID INGESTION PATHWAY DOSE FACTOR CALCULATION UNITS 1 AND 2 Dw = dilution factor from the near field area within one-quarter mile of the release point to the potable water intake for the adult water consumption. Dw includes the dilution contributions from the North Anna Dam to Doswell (0.73), the Waste Heat Treatment Facility (Cc/CL), and Lake Anna (Cr)CR_). The potable water mixing ratio is calculated as: l /(Cc/CL) (CL/ CR x 0.73 =CR/ (Cc x 0.73) where Cc / CL and CR are the respective concentrations for the considered nuclide in the Discharge Channel, Waste Heat Treatment Facility (Lagoon) and the Reservoir. Calculation is per expressions 11.2-5, 11.2-6, and 11.2-8 of North Anna's UFSAR.

BFi Da

              = the bioaccumulation factor for nuclide, i, in fish, pCi/kg per pCi/1, from Table
              =

A-1 of Regulatory Guide 1.109, Rev. 1. dilution factor for the fish pathway, calculated as CL /Cc where CL and Cc are the concentrations for the considered nuclide in the Discharge Channel and the Waste Heat Treatment Facility (Lagoon). Calculation is per Expressions 11.2-5, and 11.2-6 of North Anna's UFSAR. DFi = the critical organ dose conversion factor for nuclide, i, for adults, in mrem/pCi, from Table E-11 of Regulatory Guide 1.109, Rev. 1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 49 OF 116

ATTACHMENT 6 (Page 3 of 4)

NORTH ANNA LIOUID INGESTION PATHWAY DOSE FACTOR CALCULATION UNITS 1 AND 2 2.0 EXPRESSION "2" Expression "l" is simplified for actual dose calculations by introducing: WASTE FLOW WASTE FLOW F = CIRC.(WA TER) FLOW. + WASTE FLOW = CIRC. FLOW and CIRC. FLOW fi = EFFECTIVE OIL. FLOWi Effective dilution flow rates for individual nuclides "i" are listed on Attachment 7, North Anna Liquid Pathway Dose Commitment Factors for Adults. Then the total released activity (Qi) for the considered time period and the ith nuclide is written as:

and Expression "1" reduces to:

D Qi= txCixWASTEFLOW

                                           ~
                                         = £.J A

Qi EFF. DIL~ FLOW i For the long lived, dose controlling nuclides the effective dilution flow is essentially the over (dam) flow rate out of the North Anna Lake system (i.e., the liquid pathway dose is practically independent from the circulating water flow rate. However, to accurately assess long range average effects of reduced circulating water flow rates during outages or periods of low lake water temperatures, calculations are based on an average of 7 out of 8 circulating water pumps running at 218,000 gpm = 485.6 cft/sec per pump. By defining Bi = Ai/ EFF. OIL. FLOWi, the dose calculation is reduced to a two factor formula: D =L Qi X Bi i Values for Bi (mrem/Ci) and EFF. OIL. FLOWi are listed in Attachment 7 .

I I

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 50 OF 116 ATTACHMENT 6 (Page4of 4) NORTH ANNA LIQUID INGESTION PATHWAY DOSE FACTOR CALCULATION UNITS 1 AND 2 Total Body At Critical Organ At Individual Dilution Radionuclide Multlpllcatlon Factor (fi) mcem{bc mcem{bc

                                                     µCl/ml           µCl/ml H-3                     14.9                  6.18E+OO        6.18E+OO Na-24                    1.0                  3.71E+Ol        3.71E+Ol Cr-51                    1.7                  l.IOE+OO              -

Mn-54 7.0 8.62E+02 4.52E+03 Fe-55 11.3 1.30E+02 5.56E+02 Fe-59 2.2 9.47E+02 2.47E+03 Co-58 2.8 2.49E+02 1.l 1E+02 Co-60 13.3 8.27E+02 3.75E+02 Zn-65 6.1 3.28E+04 7.25E+04 Rb-86 1.5 3.53E+04 7.59E+04 Sr-89 2.3 8.70E+02 - Sr-90 15.8 2.39E+05 - Y-91 2.5 3.42E-01 - Zr-95 2.7 2.98E-01 -

Zr-97 Nbs95 Mo-99 Ru-103 Ru-106 Ag-llOm 1.0 1.0 1.0 2.0 7.6 6.2 1.50E-04 4.87E+Ol 7.48E+OO 4.lOE+OO 2.65E+Ol 4.94E+OO 3.27E-04 9.07E+Ol 3.93E+Ol 8.32E+OO Sb-124 2.6 4.37E+Ol 2.08E+OO Sb-125 11.4 2.46E+Ol 1.16E+OO Te-125m 2.5 3.23E+02 8.73E+02 Te-127m 3.7 7.82E+02 2.29E+03 Te-129m 1.9 1.52E+03 3.58E+03 Te-13Im 1.0 1.12E+02 1.35E+02 Te-132 1.0 5.04E+02 5.37E+02 I-131 1.2 9.66E+Ol 1.69E+02 I-132 1.0 1.03E-01 2.95E-01 I-133 1.0 3.47E+OO 1.14E+Ol I-134 1.0 2.15E-02 6.00E-02 1-135 1.0 6.58E-01 1.78E+OO Cs-134 10.3 5.80E+05 7.09E+05 Cs-136 1.3 6.01E+04 8.35E+04 Cs-137 15.8 3.45E+05 5.26E+05 Cs-138 1.0 9.18E-Ol l.85E+OO Ba-140 1.3 2.65E+Ol 5.08E-Ol La-140 1.0 4.47E-03 l.69E-02 Ce-141 1.8 2.14E-02 l.89E-01 Ce-143 1.0 1.35E-04 l.22E+OO Ce-144 6.6 1.41E+OO 1.IOE+Ol Np-239 1.0 5.13E-04 9.31E-04

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 51 OF 116 ATTACHMENT 7 (Page 1 of 1) NAPS LIQUID PATHWAY DOSE COMMITMENT FACTORS FOR ADULTS (Bi = Ai Fi/CIRC FLOW = Ai/Effluent Dilution Flowi) Effective Dilution Flow Total Body Bt Critical Organ Bt Radionuclide (cft/sec) (mrem/CI) (mremlCI) H-3 2.28E+02 2.66E-04 2.66E-04 Na-24 3.39E+03 1.07E-04 1.07E-04 Cr-51 l.99E+03 5.44E-06 NIA Mn-54 4,88E+02 l.73E-02 9.0SE-02 Fe-55 3.01E+02 4.23E-03 l.SlE-02 Fe-59 l.57E+03 5.93E-03 l.55E-02 Co-58 l.20E+03 2.04E-03 9.lOE-04 Co-60 2.55E+02 3.lSE-02 l.44E-02 Zn-65 5.60E+02 5.74E-Ol l.27E+OO Rb-86 2.34E+03 1.48E-Ol 3.ISE-01 Sr-89 l.46E+03 5.84E-03 NIA Sr-90 2.16E+02 l.09E+Ol NIA Y-91 l.34E+03 2.SOE-06 NIA Zr-95 l.27E+03 2.30E-06 1.3 lE-06 Zr-97 3.39E+03 4.33E-10 9.46E-10 Nb-95 3.25E+03 1.47E-04 2.74E-04 Mo-99 _ 3.30E+03 2.22E-05 l.17E-04 Ru-103 l.68E+03 2.40E-05 NIA Ru-106 4.48E+02 5.SOE-04 NIA Ag-llOm 5.52E+02 8.78E-05 1.48E-04 Sb-124 l.32E+03 3.25E-04 1.55E-05 Sb-125 2.98E+02 8.lOE-04 3.SOE-05 Te-125rn l.35E+03 2.35E-03 6.35E-03 Te-127rn 9.16E+02 8.37E-03 2.46E-02 Te-I29rn l.82E+03 8.19E-03 l.93E-02 Te-13lrn 3.38E+03 3.27E-04 3.92E-04 Te-132 3.27E+03 1.SlE-03 l.61E-03 1-131 2.94E+03 3.22E-04 5.62E-04 1-132 3.40E+03 2.98E:01 8.SlE-07 1-133 3.39E+03 1.00E-05 3.29E-05 1-134 3.40E+03 6.19E-08 l.73E-07 1-135 3.40E+03 l.90E-06 5.lSE-06 Cs~134 3.29E+02 l.73E+Ol 2.llE+Ol Cs-136 2.62E+03 2.25E-01 3.12E-Ol Cs-137 2.15E+02 l.57E+Ol 2.40E+Ol Cs-138 3.40E+03 2.65E-06 5.34E-06 Ba-140 2.65E+03 9.83E-05 l.SSE-06 La-140 3.36E+03 1.31E-08 4.94E-08 Ce-141 l.85E+03 l.14E-07 1.00E-06 Ce-143 3.37E+03 3.93E-10 3.55E-06 Ce-144 5.14E+02 2.70E-05 2.lOE-04 Np-239 3.32E+03 l.SlE-09 2.75E-09

VIRGINIA VPAP'-2103 POWER REVISIONO PAGE 52 OF 116 ATTACHMENT 8 (Page 1 of 3) SURRY RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Type of Activity Lower Limit of Liquid Release Detection (LLD) Analysis Type Frequency Frequency Analysis (µCi/ml), (Note 1) Principal Gamma Emitters (Note 3) 5x10-7 p p (Each Batch) (Each Batch) 1-131 lxlQ-6 A. Batch Releases p Dissolved and M Entrained Gases lxlQ-5 (One Batch/M) (Note 2) (Gamma Emitters) p MComposite H-3 lxlQ-5 (Each Batch) Gross Alpha lxl0-7 (Note 4) p QComposite Sr-89, Sr-90 5x1Q-8 (Each Batch) (Note 4) Fe-55 lxlQ-6 Principal Gamma 5x1Q-7 Continuous WComposite Emitters (Note 6) B. Continuous (Note 6) (Note 6) 1-131 lxlQ-6 Releases Dissolved and M M Entrained Gases lxl0-5 (Note 5) Grab Sample (Gamma Emitters) Continuous MComposite H-3 ixI0-5 (Note 6) (Note 6) Gross Alpha lxl0-7 Continuous QComposite Sr-89, Sr-90 5x1Q-8 (Note 6) (Note 6) Fe-55 lxl0-6

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 53 OF 116 ATTACHMENT 8 (Page2 of 3) SURRY RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Note 1: The LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD_ 4.66 Sb

- E

V
2.22 x 106
Y
e(-Allt)

Where: IlD = the "a priori" (before the fact) Lower Limit of Detection as defined above (as microcuries per unit mass or volume) . Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm). E = the counting efficiency (as counts per disintegration). V = the sample size (in units of mass or volume). 2.22x1Q6 = the number of disintegrations per minute (dpm) per microcurie. Y = the fractional radiochemical yield (when applicable). A = the radioactive decay constant for the particular radionuclide.

             .1.t  =   the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y and .1.t should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 54 OF 116 ATTACHMENT 8 (Page 3 of 3) SURRY RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Note 2: A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and appropriate methods will be used to obtain representative sample for analysis. Note 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable an4 identifiable, at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported. Note 4: A composite sample is one in whicti the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. Note 5: 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 .

Note 6: To be representative of the quantities and concentrations of radioactive materials in liquid effluents, composite sampling shall employ appropriate methods which specimen representative of the effluent release . will result in a

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 55 OF 116 ATTACHMENT 9 (Page 1 of 3) NORTH ANNA RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Liquid Release Sampling Type of Activity Analysis Detection (LLD) Type Frequency Analysis Frequency (µCi/ml), (Note 1) Principal Gamma p Emitters (Note 3) 5xl0-7 p (Each Batch) (Each Batch) 1-131 lxlQ-6 Batch Releases p Dissolved and M Entrained Gases lxI0-5 (One Batch/M) (Notes 2 and 7) (Gamma Emitters) p MComposite H-3 lxl0-5

(Each Batch) p (Each Batch)

(Note 4) QComposite (Note 4) Gross Alpha Sr-89, Sr-90 Fe-55 lxI0-7 5xl0-8 lxIQ-6 Principal Gamma Emitters (Note 6) 5xI0-7 Continuous Continuous WComposite 1-131 lxIQ-6 Releases (Note 6) (Note 6) Dissolved and Entrained Gases lxlQ-5 (Note 5) (Gamma Emitters) Continuous MComposite H-3 lxI0-5

                    * (Note 6)      (Note 6)        Gross Alpha          lxI0-7 Continuous    QComposite        Sr-89, Sr-90         5x1Q-8 (Note 6)      (Note 6)           Fe-55             lxl0-6

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 56 OF 116

                                           .ATTACHMENT 9 (Page 2 of 3)

NORTH ANNA RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Note 1: The LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that

         . will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation): LLD= 4.66 Sb E

V
2.22 x 106
Y
e(-A~t)

Where: IlD = the "a priori" (before the fact) Lower Limit of Detection as defined above (as

Sb E

microcriries per unit mass or volume). the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm).

                    = the counting efficiency (as counts per disintegration).

V = the sample size (in units of mass or volume).

            , 2.22x1Q6 = the number of disintegrations per minute (dpm) per microcurie.

Y = the fractional radiochemical yield (when applicable). A. = the radioactive decay constant for the particular radionuclide. L\t = the elapsed time between the midpoint of sample collection and time of counting. Typical values of E, V, Y and .1t should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 57 OF 116 ATTACHMENT 9 (Page 3 of 3) NORTH ANNA RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Note 2: 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 as the situation permits, to assure representative sampling. Note 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does nqt mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported. Note 4: A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

Note 5: 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.

Note 6: To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent releases. Note 7: Whenever the secondary coolant activity exceeds lQ-5 µCi/ml, the turbine building sump pumps shall be placed in manual operation and samples shall be taken and analyzed prior to release. Secondary coolant activity samples shall be collected and analyzed on a weekly basis. These samples are analyzed for gross activity or gamma isotopic activity within 24 hours .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 58 OF 116

ATTACHMENT 10 (Page 1 of 4)

SURRY RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Gaseous Release Sampling Type of Activity Detection (LLD) Type Frequency Analysis Analysis Frequency (µCi/ml), (Note 1) Prior to release. A. Waste Gas (Each Taruc) Prior to release. Principal Gamma (Each Taruc) Emitters (Note 2) lxlQ-4 Storage (Grab Sample) Tank Principal Gamma lxl0-4 B. Containment Prior to release. Prior to release. Emitters (Note 2) PURGE (Each PURGE) (Each PURGE) (Grab Sample) H-3 lxl0-6 C. Process and Weekly Weekly Principal Gamma lxl0-4 Ventilation (Grab Sample) Emitters (Note 2) Vent (Note 3) (Note 3) H-3 lxI0-6 Continuous Weekly (Note 5) lxlQ-12 (Note 4) (Charcoal Sample) 1-131 D. All Release Continuous Weekly (Note 5) Principal Gamma (Note 4) Particulate Sample Emitters (Note 2) lxI0-11 Types as Weekly Continuous Gross Alpha listed in A, (Note 4) Composite lxI0-11 Particulate Sample B, and C. Quarterly Continuous Composite Sr-89, Sr-90 (Note 4) lxI0-11 Particulate Sample Continuous Noble Gas Noble Gases Gross Beta lxlQ-6 (Note 4) Monitor and Gamma E. Condenser Weekly Principle Gamma Emitters (Note 2) lxl0-4 Air Grab Sample Weekly Ejector (Note 3) (Note 3) H-3 lxl0-6

VIRGINIA VPAP-2103 POWER REVISION 0 PAGE 59 OF 116

ATTACHMENT 10 (Page 2 of 4)

SURRY RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Principle Gamma Prior to release. Prior to release. lxl0-4 Emitters (Grab Sample) (Each release H-3 lxlQ-6 Continuous Charcoal Sample lxlQ-11 F. Containment 1-131 (Note 4) (Note 6) Continuous Particulate Sample Principle Gamma Hog lxl0-10 (Note 4) (Note 6) Emitters (Note 2) Depressuri- Composite Continuous lxlQ-10 Particulate Sample Gross Alpha zation (Note 4) (Note 6) Composite Continuous lxlQ-10

                                   *Particulate Sample   Sr-89, Sr-90 (Note 4)

(Note 6) !e

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 60 OF 116 ATTACHMENT 10 (Page 3 of 4) SURRY RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Note 1: The LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD_ 4.66 Sb

                                    -   E

V
2.22 x 1 Q6
Y
e(-11.A-r)

Where: IlD = the "a priori" (before the fact) Lower Limit of Detection as defined above (as

E

microcuries per unit mass or volume). the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm).

                   = the counting efficiency (as counts per disintegration).

V = the sample size (in units of mass or volume). 2.22x106 = the number of disintegrations per minute (dpm) per microcurie. Y = the fractional radiochemical yield (when applicable). A. = the radioactive decay constant for the particular radionuclide. At = the elapsed time between the 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 priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 61 OF 116 ATTACHMENT 10 (Page 4of 4) SURRY RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Note 2: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m, 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 nuclides with half lives greater than 8 days, that are measurable and identifiable at levels exceeding the LLD, together with the above nuclides, shall also be identified and :reported. Note 3: Sampling and analysis shall also be performed following shutdown, startup, and whenever a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER occurs within a one hour period, When:

a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has increased more than a factor of 3; and

b. The noble gas activity monitor shows that effluent activity has increased by more than

a factor of 3.

Note 4: 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 subsections 6.3.1, 6.3.3, and 6.3.4. Note 5: Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RA1ED THERMAL POWER in one hour and analyses shall be completed within 48 hours of charging. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if:

a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has increased by a factor of 3; and

b. Noble gas monitor shows that effluent activity has increased more than a factor of 3.

Note 6: To be representative of the quantities and concentrations of radioactive materials in gaseous effluents, composite sampling shall employ appropriate methods which will result in a specimen representative of the effluent release.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 62 OF 116 ATTACHMENT 11 (Page 1 of 3) NORTH ANNA RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Liquid Release Sampling Minimum Type of Activity Lower Limit of Analysis Detection (LLD) Type Frequency Frequency Analysis (µCi/ml), (Note 1) Prior to release. A. Waste Gas Prior to release. Principal Gamma (Each Tanlc lxl0-4 Storage Tank (Each Tanlc) Emitters (Note 2) Grab Sample) Principal Gamma lxl0-4 B. Containment Prior to release. Prior to release. Emitters (Note 2) PURGE (Each PURGE (Each PURGE) Grab Sample) H-3 lxl0-6 C. Ventilation Principal Gamma lxI0-4 Monthly Monthly Emitters (Note 2) (1) Process Vent (2) Vent. Vent A (Grab Sample) (Note 3) H-3 lxIQ-6 (3) Vent. Vent B (Notes 3,4, and 5) Continuous Weekly lxl0-12 I-131 (Note 4) (Charcoal Sample) D. All Release Continuous Weekly Principal Gamma (Note 4) Particulate Sample Emitters (Note 2) lxl0-11 Types as Monthly Continuous Composite Gross Alpha lxl0-11 listed in A, (Note 4) Particulate Sample B, and C. Quarterly Continuous Composite Sr-89, Sr-90 lxI0-11 (Note 4) Particulate Sample Noble Gases Continuous Noble Gas lxI0-6 Gross Beta (Note 4) Monitor or Gamma E. Cond. Air Principle Gamma Ejector Vent Weekly Weekly Emitters (Note 7) lxI0-4 Steam Gen. (Grab Sample) Blowdown H-3 lxI0-6 Vent F. Containment Principle Gamma lxI0-4 Vacuum Prior to release. Prior to each Emitters (Note 2) Steam (Grab Sample) release

Ejector (Hogger)

H-3 lxlQ-6

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 63 OF 116

ATTACHMENT 11 (Page 2 of 3)

NORTH ANNA RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Note 1: The LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD_ 4.66 Sb

                                    -   E

V
2.22 x 106
Y
e(-A~'t)

Where: ILD = the "a priori" (before the fact) Lower Limit of Detection as defined above (as

E

                   =
                   =

microcuries per unit mass or volume). the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm). the counting efficiency (as counts per disintegration). V = the sample size (in units of mass or volume). 2.22 x 1Q6 = the number of disintegrations per minute (dpm) per microcurie. Y = the fractional radiochemical yield (when applicable). A = the radioactive decay constant for the particular radionuclide.

             ~t    = the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y and ~t should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 64 OF 116

ATTACHMENT 11 (Page 3 of 3)

NORTH ANNA RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Note 2: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m, 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 that are measurable and identifiable, at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported. Note 3: Sampling and analysis shall also be performed following shutdown, startup, and whenever a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER occurs within a one hour period, if:

a. Analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant is greater than 1.0 µCi/gm; and

b. The noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.

Note 4: 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 subsections 6.3.1, 6.3.3, and 6.3.4. Note 5: Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RAIBD THERMAL POWER in one hour and analyses shall be completed within 48 hours of charging. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if:

a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant is greater than 1.0 µCi/gm and;

b. Noble gas monitor shows that effluent activity has increased more than a factor of 3.

Note 6: Whenever the secondary coolant activity exceeds 10-s µCi/ml, samples shall be obtained and analyzed weekly. The turbine building sump pumps shall be placed in manual operation and samples shall be taken and analyzed prior to release. Secondary coolant activity samples shall be collected and analyzed on a weekly basis. These samples are analyzed for gross activity or gamma isotopic activity within 24 hours. Note 7:

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m; Xe-135, Xe-135m, and Xe-138 for gaseous emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, at levels exceeding the LLD together with the above nuclides, shall also be identified and reported.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 65 OF 116 ATTACHMENT 12 (Page 1 of 3) GASEOUS EFFLUENT DOSE FACTORS FOR SURRY POWER STATION (Gamma and Beta Dose Factors)

                    'X/Q = 6.0E-05 sec/m3 at 499 meters N Direction Dose Factors for Ventilation Vent Kivv                 Livv               Mivv        Nivv Noble Gas    Total Body .            Skin           GammaAir      Beta Air Radionuclide   mr~mL)'.r           mr~mL~r            mradl)'.r   mradl)'.r Curie/Sec            Curie/Sec          Curie/Sec   Curie/Sec Kr-83m       4.54E+OO                  -              l.16E+03    1.73E+04 Kr-85m       7.02E+04              8.76E+04           7.38E+04    1.18E+05 Kr-85        9.66E+02              8.04E+04           1.03E+03    1.17E+05 Kr-87        3.55E+05             5.84E+05            3.70E+05    6.18E+05 Kr-88        8.82E+05              1.42E+05           9.12E+05    1.76E+05 Kr-89        9.96E+05             6.06E+05            1.04E+06    6.36E+05 Kr-90        9.36E+05             4.37E+05            9.78E+05    4.70E+05 Xe-13Im      5.49E+03              2.86E+04           9.36E+03    6.66E+04 Xe-133m      1.51E+04             5.96E+04            1.96E+04    8.88E+04 Xe-133       1.76E+04              1.84E+04          2.12E+04     6.30E+04 Xe-135m      l.87E+05             4.27E+04           2.02E+05     4.43E+04 Xe-135       1.09E+05              1.12E+05           1.15E+05    1.48E+05 Xe-137       8.52E+04             7.32E+05           9.06E+04     7.62E+05 Xe-138       5.30E+05             2.48E+05           5.53E+05     2.85E+05 Ar-41        5.30E+05             1.61E+05           5.58E+05     1.97E+05

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 66 OF 116

ATTACHMENT 12 (Page 2 of 3)

GASEOUS EFFLUENT DOSE FACTORS FOR SURRY POWER STATION (Gamma and Beta Dose Factors)

                      'X/Q = 1.0E-06 sec/m3 at 644 meters S Direction Dose Factors for Process Vent Kil3                 Lipv                Mipv        Nipv Noble Gas    Total ody               Skin            Gamma Air     Beta Air Radionuclide   mr~ml:}'.r          mr~ml:}'.r          mradl~r      mradl~r Curie/Sec            Curie/Sec           Curie/Sec   Curie/Sec Kr-83m        7.56E-02                  -              l.93E+Ol    2.88E+02 Kr-85m       l.17E+03              l.46E+03            l.23E+03    1.97E+03 Kr-85        1.61E+Ol              1.34E+03            1.72E+Ol    1.95E+03 l.03E+04 Kr-87        5.92E+03              9.73E+03            6.17E+03 Kr-88        1.47E+04              2.37E+03            1.52E+04    2.93E+03 Kr-89        l.66E+04              l.01E+04            1.73E+04    1.06E+04 Kr-90        1.56E+04              7.29E+03            1.63E+04    7.83E+03 Xe-131m      9.15E+Ol              4.76E+02            1.56E+02    1.l 1E+03 Xe-133m      2.51E+02              9.94E+02            3.27E+02    1.48E+03 Xe-133       2.94E+02              3.06E+02            3.53E+02    1.05E+03 Xe-135m      3.12E+03              7.11E+02            3.36E+03    7.39E+02 Xe-135       1.81E+03              1.86E+03            1.92E+03    2.46E+03 Xe-137       1.42E+03              1.22E+04            1.51E+03    1.27E+04 Xe-138       8.83E+03              4.13E+03            9.21E+03    4.75E+03 Ar-41        8.84E+03              2.69E+03            9.30E+03    3.28E+03

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 67 OF 116

ATTACHMENT 12 (Page 3 of 3)

GASEOUS EFFLUENT DOSE FACTORS FOR SURRY POWER STATION (Inhalation Pathway Dose Factors) Ventilation Vent X/Q =6.0E-05 sec/m3 at 499 meters N Direction Process Vent X/Q = 1.0E-06 sec/m3 at 644 meters S Direction Pivv Pipv Radionuclide mrem/yr mrem/yr Curie/sec Curie/sec H-3 6.75E+o4 l.12E+o3 Cr-51 5.13E+o3 8.55E+ol Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m 3.64E+o5 6.07E+o3 Te-129m 3.80E+o5 6.33E+o3 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND I-131 9.75E+o8 l.62E+07

ND - No data for dose factor according to Reg. Guide 1.109, Rev. 1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 68 OF 116

ATTACHMENT 13 (Page 1 of 3)

GASEOUS EFFLUENT DOSE FACTORS FOR NORTH ANNA POWER STATION (Gamma and Beta Dose Factors) XIQ =9.3E-06 sec/m3 at 1416 meters SE Direction Dose Factors for Ventilation Vent Kivv Livv Mivv Nivv Noble Gas Total Body Skin GammaAir Beta Air Radionuclide mr~ml~r mr~ml~r mradl~r mradl~r Curie/Sec Curie/Sec Curie/Sec Curie/Sec Kr-83m 7.03E-01 - l.79E+02 2.68E+o3 Kr-85m l.09E+04 l.36E+04 l.14E+04 l.83E+o4 Kr-85 L50E+02 l.25E+04 l.60E+02 l.81E+o4 Kr-87 5.51E+04 9.05E+04 5.74E+04 9.58E+o4

Kr-88 Kr-89 l.37E+05 l.54E+05 2.20E+04 9.39E+04 1.4IE+05 l.6IE+05 2.72E+o4 9.86E+04 Kr-90 1.45E+05 6.78E+04 l.52E+05 7.28E+o4 Xe-I31m 8.5IE+02 4.43E+03 1.45E+03 l.03E+o4 Xe-I33m 2.33E+03 9.24E+03 3.04E+03 l.38E+o4 Xe-133 2.73E+03 2.85E+03 3.28E+03 9.77E+o3 Xe-I35m 2.90E+04 6.6IE+03 3.I2E+04 6.87E+o3 Xe-135 l.68E+04 l.73E+04 l.79E+04 2.29E+o4 Xe-137 l.32E+04 1.I3E+05 1.40E+04 l.18E+o5 Xe-138 8.21E+04 3.84E+04 8.57E+04 4.42E+o4 Ar-41 8.22E+04 2.50E+04 8.65E+04 3.05E+o4

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 69 OF 116

ATTACHMENT 13 (Page 2 of 3)

GASEOUS EFFLUENT DOSE FACTORS FOR NORTH ANNA POWER STATION (Gamma and Beta Dose Factors)

                   'XJQ = 1.2E-06 sec/m3 at 1513 meters S Direction Dose Factors for Process Vent Ki~                  Lipv               Mipv         Nipv Noble Gas   Total ody               Skin           Gamma Air      Beta Air Radionuclide mr~ml~r              mr,ml~r             mradl~r     mradl~r Curie/Sec            Curie/Sec          Curie/Sec    Curie/Sec Kr-83m       9.07E-02                  -             2.32E+Ol     3.46E+02 Kr-85m      1.40E+03              1.75E+03           1.48E+03     2.36E+03 Kr-85       1.93E+Ol              l.61E+03           2.06E+Ol     2.34E+03 Kr-87       7.10E+03              1.17E+04           7.40E+03     1.24E+04

Kr-88 Kr-89 Kr-90 1.76E+04 1.99E+04 1.87E+04 2.84E+03 1.21E+04 8.75E+03 1.82E+04 2.08E+04 1.96E+04 3.52E+03 l.27E+04 9.40E+03 Xe-131m l.10E+02 5.71E+02 1.87E+02 1.33E+03 Xe-133m 3.01E+02 1.19E+03 3.92E+02 1.78E+03 Xe-133 3.53E+02 3.67E+02 4.24E+02 l.26E+03 Xe-135m 3.74E+03 8.53E+02 4.03E+03 8.87E+02 Xe-135 2.17E+03 2.23E+03 2.30E+03 2.95E+03 Xe-137 1.70E+03 1.46E+04 1.81E+03 1.52E+04 Xe-138 1.06E+04 4.96E+03 1.11E+04 5.70E+03 Ar-41 1.06E+04 3.23E+03 l.12E+04 3.94E+03

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 70 OF 116 ATTACHMENT 13 (Page 3 of 3) GASEOUS EFFLUENT DOSE FACTORS FOR NORTH ANNA POWER STATION (Inhalation Pathway Dose Factors) Ventilation Vent X/Q = 9.3E-06 sec/m3 at 1416 meters SE Direction Process Vent 'XJQ = l.2E-06 sec/m3 at 1513 meters S Direction Pivv Pipv Radionuclide mrem/yr mrem/yr Curie/sec Curie/sec H-3 l.05E+o4 1.35E+o3 Cr-51 7.95E+o2 1.02E+o2 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND

Zn-65 Rb-86 Sr-90 Y-91 Zr-95 ND ND ND ND ND ND ND ND ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m 5.64E+o4 7.28E+o3 Te-129m 5.88E+o4 7.59E+o3 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND I-131 l.51E+08 1.95E+o7
ND - No data for dose factor according to Reg. Guide 1.109, Rev. 1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 71 OF 116

ATTACHMENT 14 (Page 1 of 2)

SURRY RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT CHANNELS ACTION OPERABLE

1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release 1 1 (b) Iodine Sampler 1 2 (c) Particulate Sampler 1 2 (d) Process Vent Flow Rate Monitor 1 3

- (e) Sampler Flow Rate Measuring Device

2. CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor (b) Air Ejector Flow Rate Measuring Device 1

2 (one per unit) 2 (one per unit) 3 1 3

3. VENTILATION VENT SYSTEM (a) Noble Gas Activity Monitor 1 1 (b) Iodine Sampler 1 2 (c) Particulate Sampler 1 2 (d) Ventilation Vent Flow Rate Monitor 1 3 (e) Sampler Flow Rate Measuring Device 1 3

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 72 OF 116

ATTACHMENT 14 (Page 2of2)

SURRY RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ACTION 1: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this path may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. ACTION 2: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via the effected path may continue provided

samples are continuously collected within one hour with auxiliary sampling equipment as required in Attachment 8.

ACTION 3: 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 hours .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 73 OF 116

ATTACHMENT 15 (Page 1 of 2)

NORTH ANNA RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT CHANNELS ACTION OPERABLE

1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release 1 2,4 (b) Iodine Sampler 1 2, 5 (c) Particulate Sampler 1 2, 5 (d) Process Vent Flow Rate Measuring Device 1 1 (e) Sampler Flow Rate Measuring Device 1 1

2. CONDENSER AIR FJECTOR SYSIBM (a) Gross Activity Monitor 1 3 (b) Flow Rate Monitor 1 1

3. VENTILATION VENT SYSIBM (Shared with Unit 2)

(a) Noble Gas Activity Monitor 1 (Note 1) 2 (b) Iodine Sampler 1 (Note 1) 2 (c) Particulate Sampler 1 (Note 1) 2 (d) Flow Rate Monitor 1 (Note 1) 1 (e) Sampler Flow Rate Monitor 1 (Note 1) 1 Note 1: Orie per vent stack

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 74 OF 116

ATTACHMENT 15 (Page 2of2)

NORTH ANNA RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ACTION 1: With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this path may continue provided the flow rate is estimated at least once per 4 hours. ACTION 2: .With the number of channels OEPRABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity or gamma isotopic activity within 24 hours. ACTION 3 With the number of channels OPERABLE less than required by the minimum channels OEPRABLE requirement, effluent releases via this pathway may continue provided the frequency of the grab samples required by Technical Specification requirement 4.4.6.3.b is increased to at least once per 4 hours and these samples are analyzed for gross activity or gamma isotopic activity within 8 hours. ACTION 4: With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the Waste GasDecay Tanks may be released to the environment provided that prior to initiate the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Station Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of Waste Gas Decay Tank effluents.

ACTION 5: With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases from the Waste Gas Decay Tanks may continue provided samples are continuously collected with auxiliary sampling equipment as required in Attachment 11 .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 75 OF 116 ATTACHMENT 16 (Page 1 of 1) SURRY RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTIONAL TEST

1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor - Providing Alann and Automatic Termination of Release D M*

R Q (b) Iodine Sampler w N.A. N.A. N.A.

(c) Particulate Sampler w N.A. N.A. N.A. (d) Process Vent Flow Rate Monitor D N.A. R N.A. (e) Sampler Flow Rate Measuring Device D N.A. SA N.A.

2. CONDENSER AIR EJECTOR SYSIBM (a) Gross Activity Monitor D M R Q (b) Air Ejector Flow Rate Measuring D N.A. R N.A.

Device

3. VENTILATION VENT SYSIBM (a) Noble Gas Activity Monitor D M R Q (b) Iodine Sampler w N.A. N.A. N.A.

(c) Particulate Sampler w N.A. N.A. N.A. (d) Ventilation Vent Flow Rate Monitor D N.A. R N.A. (e) Sampler Flow Rate Measuring Device D N.A. SA N.A.

Prior to each Waste Gas Decay Taruc release

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 76 OF 116 ATTACHMENT 17 (Page 1 of 2) NORTH ANNA RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTIONAL TEST

1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release D p R Q (Note I)

(b) Iodine Sampler w N.A. N.A. N.A. (c) Particulate Sampler w N.A. N.A. N.A. (d) Process Vent Flow Rate Measuring Device D N.A. R Q (e) Sampler Flow Rate Monitor D (Note 3) N.A. R N.A.

2. CONDENSER AIR EJECTOR SYSIBM (a) Noble Gas Activity Monitor (b) Flow Rate Monitor D

D M N.A. R R Q (Note2) Q

3. VENTILATION VENT SYSTEM (Shared with Unit 2)

(a) Noble Gas Activity Monitor D M R Q (Note 2) (b) Iodine Sampler w N.A. N.A. N.A. (c) Particulate Sampler w N.A. N.A. N.A. (d) Flow Rate Monitor D N.A. R Q (e) Sampler Flow Rate Monitor D Note (3) N.A. R N.A.

l VIRGINIA VPAP-2103 POWER REVISIONO PAGE 77 OF 116 ATTACHMENT 17 (Page2of2) NORTH ANNA RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTE 1: The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the alann/trip setpoint.

b. Instrument controls not set in operate mode.

NOTE2: The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm setpoint.

b. Instrument controls not set in operate mode.

NOTE3: CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 78 OF 116 ATTACHMENT 18 (Page 1 of 2) CRITICAL ORGAN AND INHALATION DOSE FACTORS FOR SURRY (Critical Pathway Dose Factors) Ventilation Vent D/Q =9 .OE-10 m-2 at 5150 meters S Direction Process Vent D/Q =4.3E-10 m-2 at 5150 meters S Direction RMivv . RMipv Radionuclide mrem/yr mrem/yr Curie/sec Curie/sec H-3 7.20E+o2 3.12E+o2 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND

Sr-89 Sr-90 Y-91 Zr-95 ND ND ND ND ND ND ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m 8.06E+o4 3.85E+o4 Te-129m 1.25E+o5 5.98E+o4 1-131 6.21E+o8 2.97E+o8 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND
ND - No data for dose factor according to Reg. Guide 1.109, Rev. 1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 79 OF 116

ATTACHMENT 18 (Page 2 of 2)

CRITICAL ORGAN AND INHALATION DOSE FACTORS FOR SURRY (Inhalation Pathway Dose Factors) Ventilation Vent X/Q = 3.0E-07 se.c/m3 at 5150 meters S Direction Process Vent X/Q = 1.3E-07 se.c/m3 at 5150 meters S Direction Rlivv Rlipv Radionuclide mrem/yr mrem/yr Curie/se.c Curie/se.c H-3 1.94E+o2 8.41E+ol Cr-51 1.73E+ol 7.48E+o0 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND Sr-89 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m l.46E+o3 6.33E+o2 Te-129m l.64E+o3 7.12E+o2 I-131 4.45E+o6 l.93E+o6 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND

Ce-144 ND ND - No data for dose factor according to Reg. Guide 1.109, Rev. 1.

ND

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 80 OF 116 ATTACHMENT 19 (Page 1 of 1) CRITICAL ORGAN AND INHALATION DOSE FACTORS FOR NORTH ANNA (Critical Pathway Dose Factors) Ventilation Vent D/Q = 2.4E-OCJ m-2 at 3250 meters N Direction Process Vent D/Q = 1. lE-09 m-2 at 3250 meters N Direction Rivv Ripv Radionuclide mrem/yr mrem/yr Curie/sec Curie/sec H-3 l.73E+-03 9.36E+o2 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND Sr-89 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m l.97E+-05 9.04E+o4 Te-129m 2.95E+-05 l.35E+o5 1-131 1.45E+-09 6.72E+o8 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND

ND - No data for dose factor according to Reg. Guide 1.109, Rev. 1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 81 OF 116 ATTACHMENT 20 (Page 1 of 2) SURRY'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis

1. DIRECT RADIATION About 40 Routine Monitoring stations to be placed as follows:

1) Inner Ring in general area of site boundary with station in each sector. GAMMADOSE

2) Outer Ring 6 to 8 km Quarterly Quarterly from the site with a station in each sector 3)The balance of the 8

dosimeters should be placed in special interest areas such as population centers nearby residents, schools, and in 2 or 3 areas to serve as controls.

2. AIRBORNE Samples from 7 locations:

a) 1 sample from close to Radioiodine Cannister the SITE BOUNDARY location of the highest 1-131 Analysis Weekly calculated annual average ground level Continuous Particulate Sampler Radioiodines and DIQ. Sampler Gross beta radioactivity b) 5 sample locations 6-8 operation with analysis following filter Particulates km distance located in a sample collection change; concentric ring around weekly. Station. Gamma isotopic analysis of composite c) 1 sample from a control (by location) quarterly location 15-30 km distant, providing valid background data.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 82 OF 116

ATTACHMENT 20 (Page 2 of 2)

SURRY'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis

3. WATERBORNE Gamma isotopic analysis a) 1 sample upstream monthly; a) Sutface Monthly Sample b) 1 sample downstream Composite for tritium analysis quarterly.

Gamma isotopic and tritium b) Ground Sample from 1 or 2 sources Quarterly analysis quarterly 1 sample from downstream c) Sediment from Gamma isotopic analysis area with existing or Semi-Annually shoreline semi-annually potential recreational value 5 samples from vicinity of Gamma isotopic analysis d) Silt Semi-Annually the Station semi-annually 4 . INGESTION a) 4 samples from milking animals in the vicinity of Station. Gamma isotopic and 1-131 a) Mille b) 1 sample from milking Monthly analysis monthly animals at a control location (15-30 km distant) a) 3 sample of oysters in Bi-Monthly Gamma isotopic on edibles the vicinity of the Station b) 5 samples of clams in the vicinity of the Bi-Monthly Gamma isotopic on edibles Station. b) Fish and c) 1 sampling of crabs Invertebrates from the vicinity of the Annually Gamma isotopic on edibles Station . d) .2 samples of fish from the vicinity of the Station Gamma isotopic on edibles Semi-Annually (catfish, white perch, eel) a) 1 sample com Gamma isotopic on edible c) Food Products b) 1 sample soybean Annually portion c) 1 sample peanuts

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 83 OF 116

ATTACHMENT 21 (Page 1 of 4)

NORTH ANNA'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM(Note l) Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location(Note 2) Frequency Analysis

1. DIRECT RADIATION (Note3) 36 routine monitoring stations either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously to be placed as follows:

1) An inner ring of stations, one in each meteorological sector within the site boundary. GAMMAOOSE

2) An outer ring of Quarterly Quarterly stations, one in each meteorological sector within 8 km range from the site

3) The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 84 OF 116

ATTACHMENT 21 (Page 2 of 4)

NORTH ANNA'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location(Note 2) Frequency Analysis

2. AIRBORNE Samples from 5 locations:

a) 3 samples from close to the 3 site boundary Radioiodine Cannister locations (in different sectors) of the highest I-131 analysis, weekly calculated historical annual average ground Continuous Particulate Sam12ler levelD/Q. sampler Gross beta radioactivity Radioiodines and (2/3 running time Particulates b) 1 sample from the analysis following filter vicinity of a community cycle), operation change; (Note 4) having the highest with sample calculated annual collection weekly Gamma isotopic average ground level analysis of composite DIQ. (by location) c) 1 sample from a control quarterly (Note 5) location 15-40 km distant and in the least prevalent wind direction

3. WATERBORNE Sample off Gamma isotopic analysis upstream, monthly; (Note 5)

   . a) Surface       1 sample circulating water      downstream and discharge                       cooling lagoon. Composite for tritium Grab Monthly       analysis quarterly.

Sample from 1 or 2 sources Gamma isotopic and tritium b) Ground Grab Quarterly analysis quarterly (Note 5) only if likely to be affected. 1 sample from downstream Gamma isotopic analysis c) Sediment area with existing or Semi-Annually semi-annually (Note 5) potential recreational value

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 85 OF 116

ATTACHMENT 21 (Page 3 of 4)

NORTH ANNA'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location(Note 2) Frequency Analysis

4. INGESTION a) Samples from milking animals in 3 locations within 5 km distance having the highest dose potential. If there are none, then, 1 sample from milking animals in each of 3 areas between 5 to 8 km distant Monthly at Gamma isotopic (Note 5) a) MiJk(Note 7) where doses are calculated all times. and I-131 analysis to be greater than 1 mrem monthly.

per yr. (Note 6) b) 1 sample from milking animals at a control location (15-30 km distant) and in the least prevalent wind direction). a) 1 sample of commercially and recreationally important species (bass, sunfish,

b. Fish and catfish) in vicinity of plant Semiannually Gamma isotopic on edible Invertebrates discharge area. portions.

b) 1 sample of same species in areas not influenced by plant discharge a) Samples of an edible broad leaf vegetation grown nearest each of two different offsite locations of highest predicted historical annual average ground level D/Q if milk sampling is not Monthly if Gamma isotopic (Note 5) c) Food Products pe:rformed. available, or and 1-131 analysis. b) 1 sample of broad leaf at harvest vegetation grown 15-30 km distant in the least prevalent wind direction if milk

sampling is not performed

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 86 OF 116

ATTACHMENT 21 (Page4of 4)

NORTH ANNA'S RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Note 1: The number, media, frequency, and location of samples may vary from site to site. This table presents an acceptable minimum program for a site at which each entry is applicable. Local site characteristics must be examined to determine if pathways not covered by this table may

         .significantly contribute to an individual's dose and be included in the sampling program.

Note 2: For each and every sample location in Attachment 21, specific parameters of distance and direction sector from the centerline of the reactor, and additional description where pertinent, shall be provided in Attachment 23. Refer to Radiological Assessment Branch Technical Positions and to NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plant . Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to subsection 6.6.1. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a licensee Event Report and pursuant to subsection 6.6.2, identify the cause of the unavailability of samples for that pathway and identify the new locations for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report revised figures and tables for the ODCM reflecting the new locations. Note 3: One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously inay be used in place of or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations, e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading. Note 4: Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. Note 5: Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. Note 6: The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

Note 7: If milk sampling cannot be performed, use item 4.c (Pg. 3 of 4, Attachment 21)

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 87 OF 116

ATTACHMENT 22 (Page 1 of 4)

SURRY'S ENYIBONMENTAL SAMPLING LOCATIONS SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES) Site Boundary Air Charcoal and Surry Station (SS) 0.37 NNE Location at Sector Particulate with Highest D/Q Hog Island Reserve (HIR) 2.0 NNE Bacons Castle (BC) 4.5 SSW Alliance (ALL) 5.1 WSW Colonial Parkway (CP) 3.7 NNW Dow Chemical (DOW) 5.1 ENE Fort Eustis (FE) 4.8 ESE Newport News (NN) 16.5 ESE Control Location Environmental Control (00) Onsite ** 1LDs West North West (02) 0.17 WNW Site Boundary Surry Station Discharge 0.6 NW Site Boundary (03) North North West (04) 0.4 NNW Site Boundary North (05) 0.33 N Site Boundary North North East (06) 0.28 NNE Site Boundary North East (07) 0.31 NE Site Boundary East North East (08) 0.43 ENE Site Boundary East (Exclusion) (09) 0.31 E Onsite West (10) 0.40 w Site Boundary West South West (11) 0.45 WSW Site Boundary South West (12) 0.30 SW Site Boundary South South West (13) 0.43 SSW Site Boundary South (14) 0.48 s Site Boundary South South East (15) 0.74 SSE Site Boundary South East (16) 1.00 SE Site Boundary East (17) 0.57 E Site Boundary Station Intake (18) 1.23 ESE Site Boundary Hog Island Reserve (19) 1.94 NNE Near Resident

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 88 OF 116

ATTACHMENT 22 (Page 2of 4)

SURRY'S ENVIRONMENTAL SAMPLING LOCATIONS SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES) Environmental Bacons Castle (20) 4.45

SSW Approx. 5 miles 1LDs Route 633 (21) 3.5 SW Approx. 5 miles Alliance (22) 5.1 WSW Approx. 5 miles Surry (23) 8.0 WSW Population Center Route 636 and 637 (24) 4.0 w Approx. 5 miles Scotland Wharf (25) 5.0 WNW Approx. 5 miles Jamestown (26) 6.3 NW Approx. 5 miles Colonial Parkway (27) 3.7 NNW Approx. 5 miles Route 617 and 618 (28) 5.2 NNW Approx. 5 miles
Kingsmill Williamsburg Kingsmill North Budweiser Water Plant (29)

(30) (31) (32) (33) 4.8 7.8 5.6 5.7 4.8 N N NNE NNE NE. Approx. 5 miles Population Center Approx. 5 miles Population Center Approx. 5 miles Dow (34) 5.1 ENE Approx. 5 miles Lee Hall (35) 7.1 ENE Population Center Goose Island (36) 5.0 E Approx. 5 miles Fort Eustis (37) 4.8 ESE Approx. 5 miles Newport News (38) 16.5 ESE Population Center James River Bridge (39) 14.8 SSE Control Benn's Church (40) 14.5 s Control Smithfield (41) 11.5 s Control Rushmere (42) 5.2 SSE Approx. 5 miles Route 628 (43) 5.0 s Approx. 5 miles Milk Lee Hall 7.1 ENE Epp's 4.8 SSW Colonial Parkway 3.7 NNW

Judkin's William's 6.2 22.5 SSW s Control Location

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 89 OF 116

ATTACHMENT 22 (Page 3 of 4)

SURRY'S ENVIRONMENTAL SAMPLING LOCATIONS SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES) Well Water Surry Station Onsite*** Hog Island Reserve 2.0 NNE Bacons Castle 4.5 SSW Jamestown 6.3 NW Crops Slade's Farm 2.4 s State Split (Com, Peanuts, Soybeans) Brock's Farm 3.8 s State Split Poole's Garden 2.3 s State Split Crops Carter's Grove Garden 4.8 NE State Split

(Cabbage, Kale)

River Water (Bi-monthly) Ryan's Garden Surry Station Intake Hog Island Point Newport News 1.9 2.4 12.0 ESE NE SE Control Location (Chester, Va.) Chicahominy River 11.2 WNW Control Location Surry Station Discharge 0.17 NW River Water Surry Discharge 0.17 NW (Monthly) Scotland Wharf 5.0 WNW Control Location Sediment (Silt) Chicahominy River 11.2 WNW Control Location Surry Station Intake 1.9 ESE Surry Station Discharge 1.0 NNW Hog Island Point 2.4 NE Point of Shoals 6.4 SSE Newport News 12.0 SE

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 90 OF 116 ATTACHMENT 22 (Page4 of 4) SURRY'S ENVIRONMENTAL SAMPLING LOCATIONS SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES) Clams Chicahominy River 11.2 WNW Control Location Surry Station Discharge 1.3 NNW Hog Island Point 2.4 NE Jamestown 5.1 WNW Lawne's Creek 2.4 SE Oysters Deep Water Shoals 3.9 ESE Point of Shoals 6.4 SSE Newport News 12.0 SE Crabs Surry Station Discharge 0.6 NW

Fish Shoreline Sediment Surry Station Discharge Hog Island Reserve Burwell's Bay 0.6 0.8 7.76 NW N

SSE

 **     Onsite Location - in Lead Shield
 ***    Onsite sample of Well Water - taken from tap-water at Suny Environmental Building.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 91 OF 116

ATTACHMENT 23 (Page 1 of 4)

NORTH ANNA'S ENVIRONMENTAL SAMPLING LOCATIONS Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No .. (Miles) Frequency NAPS Sewage Treaunent 01 0.20 NE Quarterly & On-Site Environmental Plant Annually TI..Ds Frederick's Hall Quarterly & 02 5.30 SSW Annually Mineral,VA Quarterly & 03 7.10 WSW Annually Wares Crossroads 5.10 WSW Quarterly & 04 Annually Route 752 4.20 NNE Quarterly & 05 Annually Sturgeon's Creek Marina 05A 3.20 N Quarterly & Annually Levy, VA 4.70 ESE Quarterly & 06 Annually Bumpass, VA 7.30 SSE Quarterly & 07 Annually End of Route 685 21 1.00 WNW Quarterly & Exclusion Annually Boundary Route 700 22 1.00 WSW Quarterly & Exclusion Annually Boundary "Aspen Hills" 23 0.93 SSE Quarterly & Exclusion Annually Boundary Orange, VA 24 22.00 NW Quarterly & Control Annually Bearing Cooling Tower N-1/33 0.06 N Quarterly On-Site Sturgeon's Creek Marina N-2/34 3.20 N Quarterly Parking Lot "C" NNE-3/35 0.25 NNE Quarterly On-Site Good Hope Church NNE-4/36 4.96 NNE Quarterly Parking Lot "B" NE-5/37 0.20 NE Quarterly On-Site Lake Anna Marina NE-6/38 1.49 NE Quarterly

Weather Tower Fence ENE-7/39 0.36 ENE Quarterly On-Site Route 689 ENE-8/40 2.43 ENE Quarterly Near Training Facility E-9/41 0.30 E Quarterly On-Site L ___

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 92 OF 116

ATTACHMENT 23 (Page 2 of 4)

NORTH ANNA'S ENVIRONMENTAL SAMPLING LOCATIONS Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Environmental "Morning Glory Hill" E-10/42 2.85 E Quarterly TLDs Island Dike ESE-11/43 0.12 ESE Quarterly On-Site (cont) Route 622 ESE-12/44 4.70 ESE Quarterly Biology Lab SE-13/45 0.75 SE Quarterly On-Site Route 701 (Dam Entrance) SE-14/46 5.88 SE Quarterly "Aspen Hills" SSE-15/47 0.93 SSE Quarterly Exclusion Boundary Elk Creek SSE-15/47 0.93 SSE Quarterly Warehouse Compound Gate S-17/49 0.22 s Quarterly On-Site

Elk Creek Church NAPS Access Road Route 700 500KVTower S-18/50 SSW-19/51 SW-22/54 WSW-23/55 1.55 0.36 4.36 0.40 s

SSW SW WSW Quarterly Quarterly Quarterly Quarterly On-Site On-Site Exclusion Route 700 WSW-24/56 1.00 WSW Quarterly Boundary NAPS Radio Tower W-25/27 0.31 w Quarterly On-Site Route 685 W-26/58 1.55 w Quarterly End of Route 685 WNW-27/59 1.00 WNW Quarterly Exclusion Boundary H. Purcell's Private Road WNW-27/59 1.52 WNW Quarterly End of #1/#2 Intake NW-29/61 0.15 NW Quarterly On-Site Lake Anna Campground NW-30/62 2.54 NW Quarterly

               #1/#2 Intake              NNW-31/63        0.07    NNW     Quarterly   On-Site Route 208                 NNW-32/64        3.43    NNW     Quarterly Bumpass Post Office           C-1/2        7.30    SSE     Quarterly   Control Orange, VA                    C-3/4       22.00    NW      Quarterly   Control Mineral, VA                   C-5/6        7.10   WSW      Quarterly   Control Louisa, VA                   C-7/8        11.54   WSW      Quarterly   Control

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 93 OF 116

ATTACHMENT 23 (Page 3 of 4)

NORTH ANNA'S ENVIRONMENTAL SAMPLING LOCATIONS Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency NAPS Sewage 01 0.20 NE Weekly On-Site Airborne Treatment Plant Particulate Frederick's Hall 02 5.30 SSW Weekly and Mineral, VA 03 7.10 WSW Weekly Rad.ioiodine Wares Crossroads 04 5.10 WNW Weekly Route 752 05 4.20 NNE Weekly Sturgeon's Creek Marina 05A 3.20 N Weekly Levy, VA 06 4.70 ESE Weekly Bumpass, VA 07 7.30 SSE Weekly End of Route 685 21 1.00 WNW Weekly Exclusion Boundary Route 700 22 1.00 WSW Weekly Exclusion Boundary "Aspen Hills" 23 0.93 SSE Weekly Exclusion Boundary Orange, VA 24 22.00 NW Weekly Control Waste Heat Treatment Surface Facility (Second Cooling 08 1.10 SSE Monthly Water Lagoon) Lake Anna (upstream) 09 2.20 NW Monthly Control River North Anna River 11 5.80 SE Quarterly Water (downstream) Ground Water Biology Lab OlA 0.75 SE Quarterly (well water) Waste Heat Treatment Facility (Second Cooling 08 1.10 SSE Semi-Annually Aquatic Lagoon) Sediment Lake Anna (upstream) 09 2.20 NW Semi-Annually Control North Anna River 11 5.80 SE Semi-Annually (downstream) Shoreline Soil Lake Anna (upstream) 09 2.20 NW Semi-Annually NAPS Sewage 01 NE Once per 3 yrs On-Site 0.20 Treatment Plant Soil Mineral, VA 03 7.10 WSW Once per 3 yrs Wares Crossroads 04 5.10 WNW Once per 3 yrs Route 752 05 4.20 NNE Once per 3 yrs

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 94 OF 116

ATTACHMENT 23 (Page 4of 4)

NORTH ANNA'S ENVIRONMENTAL SAMPLING LOCATIONS Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Levy, VA 06 4.70 ESE Once per 3 yrs Soil Bumpass, VA 07 7.30 SSE Once per 3 yrs (cont) End of Route 685 21 1.00 WNW Once per 3 yrs Exclusion Boundary Route 700 22 1.00 WSW Once per 3 yrs Exclusion Boundary "Aspen Hills" 23 0.93 SSE Once per 3 yrs Exclusion Boundary Orange, VA 24 22.00 NW Once per 3 yrs Control Holladay Dairy 12 8.30 NW Monthly Milk (R.C. Goodwin) Terrell's Dairy 13 SSE Monthly 5.60 (Frederick's Hall)

Fish Waste Heat Treatment Facility (Second Cooling Lagoon)

Lake Anna (upstream) Route 713 08 09 14 1.10 2.20 varies SSE NW NE Quarterly Quarterly Control Food Products Route 614 15 varies SE Monthly

  . (Broad Leaf   Route 629/522               16      varies   NW     if available, or Control vegetation)    Route 685                  21       varies  WNW     at harvest "Aspen Hills" Area         23       varies  SSE

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 95 OF 116 ATTACHMENT 24 (Page 1 of 2) DETECTION CAPABILITIES FOR SURRY STATION ENVIRONMENTAL SAMPLE ANALYSJS(Note 1) LOWER LIMIT OF DETECTION (LLD)(Note 4) Airborne Particulate Food Sediment Analysis Water Fish Milk (pCi/kg) (pCi/1) or Gases (pCi/kg) (pCi/1) Products (Note2) (pCi/kg) (wet) (pCi/m3) (wet) (wet) Gross beta 4 0.01 H-3 2,000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 (Note 3) 1 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-140 60 60 La-140 15 15 Note 1: Required detection capabilities for thennoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13. Note 2: 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. Note 3: LLD fo_r the Ground (drinking) Water Samples. The LLD for the Surface (non-drinking Water Samples is 10 pCi/1.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 96 OF 116

ATTACHMENT 24 (Page 2 of 2)

DETECTION CAPABILITIES FOR SURRY STATION ENVIRONMENTAL SAMPLE ANALYSis<Note t) LOWER LIMIT OF DETECTION (LLD)(Note 4) Note 4: Acceptable detection capabilities for radioactive materials in environmental samples are tabulated in terms of the lower limits of detection (LLDs). LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD= 4.66 Sb E

V
2.22 x 1 Q6
Y
exp (-AA)

Where:

l.LD Sb

                   = the "a priori" (before the fact) Lower Limit of Detection as defined above (as microcuries per unit mass or volume).
                   = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm).

E the counting efficiency (as counts per disintegration). V = the sample size (in units of mass or volume). 2.22 x 106 = the number of disintegrations per minute (dpm) per microcurie. Y = the fractional radiochemical yield (when applicable). A = the radioactive decay constant for the particular radionuclide. At = the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples). Typical values ofE, V, Y and At should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 97 OF 116

ATTACHMENT 25 (Page 1 of 2)

DETECTION CAPABILITIES FOR NORTH ANNA STATION ENVIRONMENTAL SAMPLE ANALYSIS(Note t) LOWER LIMIT OF DETECTION (LLD)(Note 3) Airborne Particulate Food Sediment Analysis Water Fish Milk Products (pCi/kg) (pCi/1) or Gases (pCi/kg) (pCi/1) (Note2) (pCi/kg) (wet) (pCi/m3) (wet) (wet) Gross beta 4 0.01 H-3 2,000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 I-131 (Note 3) 1 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 Note 1: This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.8. Note 2: This LLD value is for drinking water samples.

II VIRGINIA VPAP:-2103 I POWER REVISIONO I PAGE 98 OF 116

ATTACHMENT 25 (Page2 of2)

DETECTION CAPABILITIES FOR NORTH ANNA STATION ENVIRONMENTAL SAMPLE ANALYSIS(Note 1) LOWER LIMIT OF DETECTION (LLD)(Note 3) Note 3: Acceptable detection capabilities for radioactive materials in environmental samples are tabulated in terms of the lower limits of detection (LLDs). LLD is defined, for purposes of this requirement, as the smallest concentration of radioactive material in a sample that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD= 4.66 Sb E

V
2.22 x 106
Y
exp (-AL\)

Where:

ILD = the "a priori" (before the fact) Lower Limit of Detection as defined above (as Sb microcuries per unit mass or volume).

                      = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm).

E = the counting efficiency (as counts per disintegration). V = the sample size (in units of mass or volume). 2.22 x 1()6 = the number of disintegrations per minute (dpm) per microcurie. Y = the fractional radiochemical yield (when applicable). A = the radioactive decay constant for the particular radionuclide.

                .L\t  = the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).

Typical values ofE, V, Y and L\t should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an a "posteriori" (after the fact) limit for a particular measurement.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 99 OF 116

ATTACHMENT 26 (Page 1 of 1)

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES AT SURRY STATION Airborne Water Particulate Fish Milk Food Analysis (pCi/1) or Gases (pCi/I) Products (pCi/kg, wet) (pCi/kg, wet) (pCi/m3) H-3 30,000 Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000

Zn-65 Zr-Nb-95 1-131 300 400 (Note 1) 2 0.9 20,000 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 Note 1: Reporting Level for the Ground (drinking) Water Samples required by Attachment 20. The Reporting Level for the Surface (non-drinking) Water Samples required by Attachment 20 is 20 pCi/1 .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 100 OF 116 ATTACHMENT 27 (Page 1 of 1) REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES AT NORTH ANNA STATION Airborne Water Particulate Fish Milk Food Analysis (pCi/1) or Gases (pCi/kg, wet) (pCi/1) Products (pCi/kg, wet) (pCi/m3) H-3 20,()()()(1) Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000

Zn-65 Zr-Nb-95 1-131 300 400 2 0.9 20,000 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 Note 1: For drinking water samples .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 101 OF 116

ATTACHMENT 28 (Page 1 of 8)

SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS

1. 0 METEOROLOGICAL ANALYSIS 1.1 Purpose The purpose of the meteorological analysis was to determine the annual average X/Q and D/Q values at critical locations around the Station for ventilation vent (ground level) and process vent (mixed mode) releases. The annual average X/Q and D/Q values were used in performing a dose pathway analysis to determine both the maximum exposed individual at SITE BOUNDARY and MEMBER OF TIIE PUBLIC. The 'XJQ and D/Q values resulting in the maximum exposures were incorporated into the dose factors in Attachments 12 and 18.

1. 2 Meteorological Data, Parameters, and Methodology Onsite meteorological data for the period January l, 1979, through December 31, 1981, was used in calculations. This data included wind speed, wind direction, and differential

- temperature for the purpose of determining joint frequency distributions for those releases characterized as ground level (i.e., ventilation vent), and those characterized as mixed mode (i.e., process vent). The portions of release characterized as ground level were based on

         ~T1ss.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were based on ~T158.9ft-28.2ft and 158.9 ft wind data.
         'X/Q's and D/Q's were calculated using the NRC computer code "XOQDOQ - Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations",

September, 1977. The code is based upon a straight line airflow model implementing the assumptions outlined in Section C (excluding Cla and Clb) of Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light - Water- Cooled Reactors". The open terrain adjustment factors were applied to the 'X/Q values as recommended in Regulatory Guide 1.111. The site region is characterized flat terrain such that open terrain correction factors are considered appropriate. The ground level ventilation vent release calculations included a building wake correction based on a 1516 m2 containment minimum cross-sectional area. The effective release height used in mixed mode release calculations was based on a process vent release height of 131 ft, and plume rise due to momentum for a vent diameter of 3 in. with plume exit velocity of 100 ft/sec.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 102 OF 116 ATTACHMENT 28 (Page 2 of 8) SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Ventilation vent, and vent releases other than from the process vent, are considered ground level as specified in Regulatory Guide 1.111 for release points less than the height of adjacent solid structures, terrain elevations were obtained from Surry Power Station Units 1 and 2 Virginia Electric and Power Company Updated Final Safety Analysis Report Table 1 lA-11. X/Q and D/Q values were calculated for the nearest SITE BOUNDARY, resident, milk cow, and vegetable *garden by sector for process vent and ventilation vent releases. X/Q values were also calculated for the nearest discharge canal bank for process and ventilation vent releases. According to the definition for short term in NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Stations", October, 1978, some gaseous releases may fit this category, primarily waste gas decay tank releases and containment purges. However, these releases are considered long term for dose calculations as past releases were both random in time of day and duration as evidenced by reviewing past release reports . Therefore, the use of annual average concentrations is appropriate according to NUREG-0133. 1.3 Results The X/Q value that resulted in the maximum total body, skin and inhalation exposure for ventilation vent releases was 6.0E-05 sec/m3 at a SITE BOUNDARY location 499 meters N sector. For process vent releases, the SITE BOUNDARY X/Q value was l.OE-06 sec/m3 at a location 644 meters S sector. The discharge canal bank X/Q value that resulted in the maximum inhalation exposure for ventilation vent releases was 7 .8E-05 sec/m3 at a location 290 meters NW sector. The discharge canal bank X/Q value for process vent was 1.6E-06 sec/m3 at a location 290 meters NW sector. Pathway analysis indicated that the maximum exposure from I-131, and from all radionuclides in particulate form with half-lives greater than 8 days was through the grass-cow-milk pathway. The D/Q value from ventilation vent releases resulting in the maximum exposure was 9.0E-10 per m2 at a location 5150 meters S sector. For process vent releases, the D/Q value was 4.3E-10 per m2 at a location 5150 meters S sector. For tritium, the X/Q value from ventilation vent releases resulting in the maximum exposure for the milk pathway was 3.0E-07 sec/m3, and l.3E-07 sec/m3 for process vent releases at a location 5150 meters S sector. The inhalation pathway is the only other pathway existing at this location. Therefore, the X/Q values given for tritium also apply for the inhalation pathway.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 103 OF 116

ATTACHMENT 28 (Page 3 of 8)

SURRY METEOROLOGICAL, LIOUID AND GASEOUS PATHWAY ANALYSIS 2.0 . LIQUID PATHWAY ANALYSIS 2.1 Purpose The pmpose of the liquid pathway analysis was to determine the maximum exposed MEMBER OF THE PUBLIC in UNRES1RICTED AREAS as a result of radioactive liquid effluent releases. The analysis includes a determination of most restrictive liquid pathway, most restrictive age group, and critical organ. This analysis is required for subsection 6.2, Liquid Radioactive Waste Effluents.

2. 2 Data, Parameters, and Methodology Radioactive liquid effluent release data for the years 1976, 1977, 1978, 1979, 1980, and 1981 was compiled from the Surry Power Station effluent release reports. The data for each year, along with appropriate site specific parameters and default selected parameters, was entered into the NRC computer code LADTAP as described in NUREG-0133 .

Liquid radioactive effluents from both units are released to the James River via the discharge canal. Possible pathways of exposure for release from the Station include ingestion of fish and invertebrates and shoreline activities. The irrigated food pathway and potable water pathway do not exist at this location. Access to the discharge canal by the general public is gained two ways: access for bank fishing is controlled by the Station and is limited to Virginia Power employees or guests of employees, and boating access is open to the public as far upstream as the inshore end of the discharge canal groin. It has been estimated that boat sport fishing would be performed a maximum of 800 hours per year, and that bank fishing would be performed a maximum of 160 hours per year. For an individual fishing in the discharge canal, no river dilution was assumed for the fish pathway. For an individual located beyond the discharge canal groins, a river dilution factor of 5 was assumed as appropriate according to Regulatory Guide 1.109, Rev. 1, and the fish, invertebrate, and shoreline pathways were considered to exist. Dose factors, bioaccumulation factors, and shore width factors given in Regulatory Guide 1.109, Rev. 1, and in LADTAP were used, as were usage terms for shoreline activities and ingestion of fish and invertebrates. Dose to an individual fishing on the discharge bank was determined by multiplying the annual dose calculated with LADTAP by the fractional year the individual spent fishing in the canal.

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ATTACHMENT 28 (Page4of 8)

SURRY METEOROLOGICAL, LIOUID AND GASEOUS PATHWAY ANALYSIS 2.3 RESULTS For the years 1976, 1977, 1979, 1980, and 1981, the invertebrate pathway resulted in the largest dose. In 1978 the fish pathway resulted in the largest dose. The maximum exposed

MEMBER OF THE PUBLIC was determined to utilize the James River. The critical age group was the adult and the critical organ was either the thyroid or GI-LLI. The ingestion dose factor, Ai, in subsection 6.2.3, Liquid Effluent Dose Limit, includes the fish and invertebrate pathways. Ai dose factors were calculated for the total body, thyroid, and GI-LLI organs.

3.0 GASEOUS PATHWAY ANALYSIS 3.1 Purpose A gaseous effluent pathway analysis was performed to determine the location that would result in the maximum doses due to noble gases for use in demonstrating compliance with

subsections 6.3.1.a and 6.3.3.a. The analysis also included a determination of the location, pathway, and critical organ, of the maximum exposed :MEMBER OF THE PUBLIC, as a result of the release of 1-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for use in demonstrating compliance with subsection 6.3.4.a. In addition, the analysis includes the determination of the critical organ, maximum age group, and sector location of an exposed individual through the inhalation pathway from 1-131, tritium, and particulates for use in demonstrating compliance with subsection 6.3.1.a.

3. 2 Data, Parameters, and Methodology Annual average 'X/Q values were calculated, as described in subsection 1 of this attachment, for the nearest SITE BOUNDARY in each directional sector and at other critical locations accessible to the public inside SITE BOUNDARY. The largest 'X/Q value was determined to be 6.0E-05 sec/m3 at SITE BOUNDARY for ventilation vent releases at a location 499 meters N direction, and l.OE-06 sec/m3 at SITE BOUNDARY for process vent releases at a location 644 meters S direction. The maximum doses to total body and skin, and air doses for gamma and beta radiation due to noble gases would be at these SITE BOUNDARY locations. The doses from both release points are summed in calculations to calculate total maximum dose .

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 105 OF 116

ATTACHMENT 28 (Page 5 of 8)

SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Step 6.3.1.a.2 dose limits apply specifically to the inhalation pathway. therefore, the locations and 'X/Q values determined for maximum noble gas doses can be used to determine the maximum dose form 1-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway. The NRC computer code GASPAR, "Evaluation of Atmospheric Releases", Revised 8/19n7, was run using 1976, 1977, 1978, 1979, 1980 and 1981 Surry Power Station gaseous effluent release report data. Doses from I-131, tritium, and particulates for the inhalation pathway were calculated using the 6.0E-05 sec/m3 SITE BOUNDARY X/Q. except for the source term data and the X/Q value, computer code default parameters were used. Results for each year indicated that the critical age group was the child and the critical organ was the thyroid for the inhalation pathway. In 1979, the teen was the critical age group. However, the dose calculated for the teen was only slightly greater than for the child and the doses could be considered equivalent The gamma and beta dose factors Kivv, Livv, Mivv, and Nivv in Attachment 12 were obtained by performing a units conversion of the appropriate dose factors from Table B-1, Regulatory Guide 1.109, Rev. 1, to mrem/yr per CiJm3 or mrad/yr per CiJm3, and multiplying by the ventilation vent SITE BOUNDARY X/Q value of 6.0E-05 sec/m3. The same approach was used in calculating the gamma and beta dose factors Kipv, Lipv, Mipv, and Nipv in Attachment 12 using the process vent SITE BOUNDARY XIQ value of l.OE-06 sec/m3. Inhalation pathway dose factors Pivv and Pipv in Attachment 12 were calculated using the following equation: Pi= K' (BR) DFAi (X/Q (mrem/yr per Curie/sec) where: K' = a constant of unit conversion, IE+ 12 pCi/Ci BR = the breathing rate of the child age group, 3700 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev.I DFAi = the thyroid organ inhalation dose factor for child age group for the ith radionuclide, in mrem/pCi, from Table E-9, Regulatory Guide 1.109, Rev. 1

'XJQ_ = the ventilation vent SITE BOUNDARY X/Q, 6.0E-5 sec/m3, or the process vent SITE BOUNDARY 'X/Q, 1.0E-06 sec/m3 as appropriate.

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ATTACHMENT 28 (Page 6of8)

SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Subsection 6.3.4.a, requires that the dose to the maximum exposed MEMBER OF THE PUBLIC from I-131, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days be less than or equal to the specified limits. Dose calculations were performed for an exposed MEMBER OF TIIB PUBLIC within SITE BOUNDARY UNRESTRICfED AREAS, discharge canal bank, and to an exposed MEMBER OF THE PUBLIC beyond SITE BOUNDARY at real residences with the largest 'X/Q values using the NRC computer code GASPAR. Doses to MEMBERS OF THE PUBLIC were also calculated for the vegetable garden, meat animal, and milk-cow pathways with the largest D/Q values using the NRC computer code GASPAR. It was determined that the MEMBER OF THE PUBLIC within SITE BOUNDARY would be using the discharge canal bank for fishing a maximum of 160 hours per year. The maximum annual X/Q at this location was determined to be 7.8E-05 sec/m3 at 290 meters NW direction.

After applying a correction for the fractional part of year an individual would be fishing at this location, the dose was calculated to be less than an individual would receive at SITE BOUNDARY.

The MEMBER OF THE PUBLIC receiving the largest dose beyond SITE BOUNDARY was determined to be located 5150 meters S sector. The critical pathway was the grass-cow-milk, the maximum age group was the infant, and the critical organ the thyroid. For each year 1976, 1977, 1978, 1979, 1980 and 1981 the dose to the infant from the grass-cow-mild pathway was greater than the dose to the MEMBER OF THE PUBLIC within SITE BOUNDARY, nearest residence, vegetable or meat pathways. Therefore, the maximum exposed MEMBER OF THE PUBLIC was determined to be the infant, exposed through the grass-cow-milk pathway, critical organ thyroid, at a location 5150 meters S sector. The only other pathway existing at this location for the infant is the inhalation.

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ATTACHMENT 28 (Page 7 of 8)

SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS The RMivv and RMipv dose factors, except for tritium, in Attachment 18 were calculated by multiplying the appropriate D/Q value with the following equation: RMi = K' Qp (Uap) Fm (r) (DFLi) [fpfs + ( 1-fpfs) e-Aith] e-Aitf A'+ 1 AW Yp Ys .* where: K' = a constant of unit conversion, lE+ 12 pCi/Ci Qp = cow's consumption rate, 50, in Kg/day (wet weight) Uap = infant milk consumption rate, 330, liters/yr Yp = agricultural productivity by unit area of pasture feed grass, 0. 7 Kg!m2 Ys = agricultural productivity by unit area of stored feed, 2.0, in Kg!m2

Fm r

DFLj

            =  stable element transfer coefficients, from Table E-1, Regulatory Guide 1.109, Rev. 1
            = fraction of deposited activity retained on cow's feed grass, 1.0 for radioiodine, and
            =

0.2 for particulates thyroid ingestion dose factor for the ith radionuclide for the infant, in mrern/pCi, from Table E-14, Regulatory Guide 1.109, Rev.I Ai = decay constant for the ith radionuclide, in sec-I Aw = decay constant for removal of activity of leaf and plant surfaces by weathering, 5.73E-07 sec-I (corresponding to a 14 day half-life) tf = transport time from pasture to cow, to milk, to receptor, 1.73+05, in seconds tii = transport time from pasture, to harvest, to cow, to milk, to receptor, 7 .78E+06, in seconds fp = fraction of year that cow is on pasture, 0.67 (dimensionless), 7.78E+06 in seconds fs = fraction of cow feed that is pasture grass while cow is on pasture, 1.0, dimensionless Parameters used in the above equation were obtained from NUREG-0133 and Regulatory Guide *1.109,.Rev.1 .

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ATTACHMENT 28 (Page 8 of 8)

SURRY METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Since the concentration of tritium in milk is based on the airborne concentration rather than the deposition, the following equation is used: RH-3 = K' K"' Fm Qp Uap (DFLH-3) [0.75 (0.5/H)] x X/Q where: K"' = a constant of unit conversion 1E+o3 gm/kg H = absolute humidity of the atmosphere, 8.0, gmlm3 0.75 = the fraction of total feed that is water 0.5 = the*ratio of the specific activity of the feed grass to the atmospheric water X/Q = the annual average concentration at a location 5150 meters S sector, 3.0E-07 sec/m3 for ventilation vent releases, and 1.3E-07 sec/m3 for the process vent releases Other parameters have been previously defined.

The inhalation pathway dose factors Rlivv and Rlipv in Attachment 18 were calculated using the following equation:

where: Rli = K' (BR) DFAi (X/Q) (mrem/yr per Curie/sec) K' = a constant of unit conversion, lE+ 12 pCi/Ci BR = breathing rate of the infant age group, 1400 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev .1 DFAi = thyriod organ inhalation dose factor for infant age group for the ith radionuclide, in mrem/pCi, from Table R-10, Regulatory Guide 1.109, Rev.1 XIQ = ventilation vent 'X/Q, 3.0E-07 sec/m3, or the process vent SITE BOUNDARY 'X/Q, 1.3E-07 sec/m3, at a location 5150 meters S sector. TheGASPARcomputerrunsusing 1976, 1977, 1978, 1979, 1980and 1981 Surry effluent release data were reviewed to determine the percent of total dose from the cow milk and inhalation pathways for 1-133. I-133 contributed less than 1% of the total dose to an infant's thyroid except for the year 1977 when the percent 1-133 was 1.77. The calculations indicate that I-133 is a negligible dose contributor and it's inclusion in a sampling and analysis program, and dose calculation is unnecessary.

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ATTACHMENT 29 (Page 1 of 8)

NORTH ANNA METEOROLOGICAL, LIOUID AND GASEOUS PATHWAY ANALYSIS

1. 0 METEOROLOGICAL ANALYSIS 1.1 Purpose The purpose of the meteorological analysis was to determine the annual average 'X/Q and D/Q values at critical locations around the Station for ventilation vent (ground level) and process vent (mixed mode) releases. The annual average 'X/Q and D/Q values were used in performing a dose pathway analysis to determine both the maximum exposed individual at SITE BOUNDARY and :tvffiMBER OF THE PUBLIC. The 'X/Q and D/Q values resulting in the maximum exposures were incorporated into the dose factors in Attachments 13 and 19.

1.2 Meteorological Data, Parameters, and Methodology Onsite meteorological data for the period January 1, 1981, through December 31, 1981, was used in calculations. This data included wind speed, wind direction, and differential temperature for the purpose of determining joint frequency distributions for those releases characterized as ground level (e.g., ventilation vent), and those characteriz.ed as mixed mode (i.e., process vent). The portions of release characterized as ground level were based on L\T1s8.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were based on AT1s8.9ft-28.2ft and 158.9 ft wind data. X/Q's and D/Q's were calculated using the NRC computer code "XOQDOQ - Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations", September, 1977. The code is based upon a straight line airflow model implementing the assumptions outlined in Section C (excluding C1a and C 1b) of Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light - Water - Cooled Reactors". The open terrain adjustment factors were applied to the 'X/Q values as recommended in Regulatory Guide 1.111. The site region is characterized by gently rolling terrain such that open terrain correction factors are considered appropriate. The ground level ventilation vent release calculations included a building wake correction based on a 1516 m2 containment

minimum cross-sectional area.

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ATTACHMENT 29 (Page 2of 8)

NORTH ANNA METEOROLOGICAL, LIOUID AND GASEOUS PATHWAY ANALYSIS The effective release height used in mixed mode release calculations was based on a process vent release height of 157 .5 ft, and plume rise due to momentum for a vent diameter of 3 in. with plume exit velocity of 100 ft/sec. Ventilation vent, and vent releases other than from the process vent, are considered ground level as specified in Regulatory Guide 1.111 for release points less than the height of adjacent solid structures, terrain elevations were obtained from North Anna Power Station Units 1 and 2 Virginia Electric and Power Company Final Safety Analysis Report Table 11 C.2-8. X/Q and D/Q values were calculated for the nearest SITE BOUNDARY, resident, milk cow, and vegetable garden by sector for process vent and ventilation vent releases at distances specified from North Anna Power Station Annual Environmental Survey Data for 1981. X/Q values were also calculated for the nearest lake shoreline by sector for the process vent and

ventilation ventreleases .

According to the definition for short term in NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Stations", October, 1978, some gaseous releases may fit this category, primarily waste gas decay tank releases and containment purges. However, these releases are considered long term for dose calculations as past releases were both random in time of day and duration as evidenced by reviewing past release reports. Therefore, the use of annual average concentrations is appropriate according to NUREG-0133. The X/Q and D/Q values calculated from 1981 meteorological data are comparable to the values presented in the North Anna Power Station UFSAR. 1.3 Results The X/Q value that resulted in the maximum total body, skin and inhalation exposure for ventilation vent releases was 9.3E-06 sec/m3 at a SITE BOUNDARY location 1416 meters SE sector. For process vent releases, the SITE BOUNDARY XIQ value was 1.2E-06 sec/m3 at a . location 1513 meters S sector. The shoreline X/Q value that resulted in the maximum inhalation exposure for ventilation vent releases was 1.0E-04 sec/m3 at a location 241 meters NNE

sector. The shoreline X/Q value for process vent was 3.7E-06 sec/m3 at a location 241 meters NNE sector.

 .VIRGINIA                                                                                     VPAP-2103 POWER                                                                                 REVISIONO PAGE 111 OF 116

ATTACHMENT 29 (Page 3 of 8)

NORTH ANNA METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Pathway analysis indicated that the maximum exposure from I-131, and from all radionuclides in particulate form with half-lives greater than 8 days was through the grass-cow-milk pathway. The D/Q value from ventilation vent releases resulting in the maximum exposure was 2.4E-09 per m2 at a location 3250 meters N sector. For process vent releases, the D/Q value was 1. lE-09 per m2 at a location 3250 meters N sector. For tritium, the X/Q value from ventilation vent releases resulting in the maximum exposure for the milk pathway was 7.2E-07 sec/m3, and 3.9E-07 sec/m3 for process vent releases at a location 3250 meters N sector. 2.0 LIQUID PATHWAY ANALYSIS 2.1 Purpose The purpose of the liquid pathway analysis was to determine the maximum exposed MEMBER OF TIIE PUBLIC in UNRESTRICI'ED AREAS as a result of radioactive liquid effluent releases. The analysis includes a determination of most restrictive liquid pathway, most restrictive age group, and critical organ. This analysis is required for subsection 6.2, Liquid Radioactive Waste Effluents.

2. 2 Data, Parameters, and Methodology Radioactive liquid effluent release data for the years 1979, 1980, and 1981 was compiled from the North Anna Power Station semi-annual effluent release reports. The data for each year, along with appropriate site specific parameters and default selected parameters, was entered into the NRC computer code LADTAP as described in NUREG-0133.

Reconcentration of effluents using the small lake connected to larger water body model was selected with the appropriate parameters determined from Table 3.5.3.5, Design Data for Reservoir and Waste Heat Treatment Facility from Virginia Electric and Power Company, Applicant's Environmental Report Supplement, North Anna Power Station, Units 1 and 2, March 15, 1972. Dilution factors for aquatic foods, shoreline, and drinking water were set to one. Transit time calculations were based on average flow rates. All other parameters were defaults selected by the LADTAP computer code.

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ATTACHMENT 29

                                                    -(Page 4 of 8)

NORTH ANNA METEOROLOGICAL, LIOUID AND GASEOUS PATHWAY ANALYSIS 2.3 RESULTS For each year, the fish pathway resulted in the largest dose. The critical organ each year was the liver, and the adult and teenage age groups received the same organ dose. However, since the adult total body dose was greater than the teen total body dose for each year, the adult was selected as the most restrictive age group. Dose factors in Attachment 7 are for the maximum exposed MEMBER OF THE PUBLIC, an adult, with the critical organ being the liver. 3.0

GASEOUS PATHWAY ANALYSIS
3 .1 Purpose A gaseous effluent pathway analysis was performed to determine the location that would result in the maximum doses due to noble gases for use in demonstrating compliance with
subsections 6.3.1.a and 6.3.3.a. The analysis also included a determination of the critical pathway, location of maximum exposed MEMBER OF THE PUBLIC, and the critical organ
for the maximum dose due to 1-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for use in demonstrating compliance with requirements in step 6.3.1.a.1 and subsection 6.3.3.a. The Analysis also included a determination of the critical pathway, location of maximum exposed MEMBER OF THE PUBLIC, and the critical organ for the maximum dose due to 1-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for use in demonstrating compliance with step 6.3.1.a.2 and subsection 6.3.4.a.

3. 2 Data, Parameters, and Methodology Annual average 'XJQ values were calculated, as described in subsection 1 of this attachment, for the nearest SITE BOUNDARY in each directional sector and at other critical locations beyond

            *the SITE BOUNDARY. The largest X/Q value was determined to be 9.3E-06 sec/m3 at SITE BOUNDARY for ventilation vent releases at a location 1416 meters SE direction, and l.2E-06 sec/m3 at SITE BOUNDARY for process vent releases at a location 1513 meters S direction.

The maximum doses to total body and skin, and air doses for gamma and beta radiation due to

- noble gases would be at these SITE BOUNDARY locations. The doses from both release points are summed in calculations to calculate. total maximum dose.

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ATTACHMENT 29 (Page 5 of 8)

NORTH ANNA METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Step 6.3.1.a.2 dose limits apply specifically to the inhalation path.way. therefore, the locations and 'X/Q values determined for maximum noble gas doses can be used to determine the maximum dose form 1-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway. The NRC computer code GASPAR, "Evaluation of Atmospheric Releases", Revised 8/19fi7, was run using 1979, 1980 and 1981 North Anna Power Station Gaseous Effluent Release Report data. Doses from 1-131, tritium, and particulates for the inhalation pathway were calculated using the 9.3E-06 sec/m3 SITE BOUNDARY 'X/Q. Except for the source term data and the 'X/Q value, computer code default parameters were used. Results for each year indicated that the critical age group was the child and the critical organ was the thyroid for the inhalation pathway.

The gamma and beta dose factors Kivv, Livv, Mivv, and Nivv in Attachment 12 were obtained by performing a units conversion of the appropriate dose factors from Table B.:.1, Regulatory Guide 1.109, Rev. 1, to mrem/yr per Cifm3 or mrad/yr per Cifm3, and multiplying by the ventilation vent SITE BOUNDARY 'X/Q value of9.3E-06 sec/m3. The same approach was used in calculating the gamma and beta dose factors Kipv, Lipv, Mipv, and Nipv in Attachment 13 using the process vent SITE BOUNDARY X/Q value of 1.2E-06 sec/m3.

The inhalation pathway dose factors Pivv and Pipv in Attachment 13 were calculated using the following equation: Pi= K' (BR) DFAi ('X/Q) (mrem/yr per Curie/sec) where: K' = a constant of unit conversion, lE+ 12 pCi/Ci BR = the breathing rate of the child age group, 3700 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev.I DFAi = the thyroid organ inhalation dose factor for child age group for the ith radionuclide, in mrem/pCi, from Table E-9, Regulatory Guide 1.109, Rev. 1

      'XIQ    = the ventilation vent SITE BOUNDARY 'X/Q, 9.3E-06 sec/m3, or the process vent

SITE BOUNDARY 'X/Q, l.2E-06 sec/m3 as appropriate.

 ----~-- --

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 114 OF 116

ATTACHMENT 29 (Page 6of 8)

NORTH ANNA METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Subsection 6.3.4.a, requires that the dose to the maximum exposed MEMBER OF THE PUBLIC from 1-131, tritium, and from allradionuclides in particulate form with half-lives greater than 8 days be less than or equal to the specified limits. Dose calculations were performed for an exposed MEMBER OF THE PUBLIC within SITE BOUNDARY UNRESTRICfED AREAS, and to an exposed MEMBER OF THE PUBLIC beyond SITE BOUNDARY at locations identified in the North Anna Power Station Annual Environmental Survey Data for 1981. It was determined that the MEMBER OF THE PUBLIC within SITE BOUNDARY would be using Lake Anna for recreational purposes a maximum of 2232 hours per year. It is assumed that this MEMBER OF THE PUBLIC would be located the entire 2232 hours at the lake shoreline with the largest annual 'X/Q of 1.0E-04 at a location 241 meters NNE sector. The

NRC computer code GASPAR was run to calculate the inhalation dose to this individual. The GASPAR results were corrected for the fractional year the MEMBER OF THE PUBLIC would be using the lake.

Using the NRC computer code GASPAR and annual average 'X/Q and D/Q values obtained as described in subsection 1 of this attachment the MEMBER OF THE PUBLIC receiving the largest dose beyond SITE BOUNDARY was determined to be located 3250 meters N sector. The critical pathway was the grass-cow-milk, the maximum age group was the infant, and the critical organ the thyroid. For each year 1979, 1980 and 1981 the dose to the infant from the grass-cow-milk pathway was greater than the dose to the MEMBER OF THE PUBLIC within SITE BOUNDARY. Therefore, the maximum exposed MEMBER OF THE PUBLIC was determined to be the infant, exposed through the grass-cow-milk pathway, critical organ thyroid, at a location 3250 meters N sector.

------VIRGINIA                                                                                     VPAP-2103 POWER                                                                                   REVISIONO PAGE 115 OF 116

ATTACHMENT 29 (Page 7 of 8)

NORTH ANNA METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS The Rivv and Ripv dose factors, except for tritium, in Attachment 19 were calculated by multiplying the appropriate D/Q value with the following equation: Ri = K' Qp (Uap) Fm (r) (DFLi) [ ~ + (1-fpfs) e-Aith] e-Aitf Ai+ Aw Yp Ys where: K' = a constant of unit conversion, IE+ 12 pCi/Ci Qp = cow's consumption rate, 50, in Kg/day (wet weight) Uap = infant milk consumption rate, 330, liters/yr Yp = agricultural productivity by unit area of pasture feed grass, 0. 7 Kg!m2 Ys = agricultural productivity by unit area of stored feed, 2.0, in Kg!m2 Fm = stable element transfer coefficients, from Table E-1, Regulatory Guide 1.109, Rev. 1 r = fraction of deposited activity retained on cow's feed grass, 1.0 for radioiodine, and 0.2 for particulates DFLi = thyroid ingestion dose factor for the ith radionuclide for the infant, in mrem/pCi, from Table E-14, Regulatory Guide 1.109, Rev.I Ai = decay constant for the ith radionuclide, in sec-1 Aw = decay constant for removal of activity of leaf and plant surfaces by weathering, 5.73E-07 sec-1 (corresponding to a 14 day half-life) tr = transport time from pasture to cow, to milk, to receptor, 1.73E+o5, in seconds lb = transport time from pasture, to harvest, to cow, to milk, to receptor, 7.78E+o6, in seconds fp = fraction of year that cow is on pasture, 0.58 (dimensionless), 7 months per year from NUREG-0597 fs = fraction of cow feed that is pasture grass while cow is on pasture, 1.0, dimensionless Parameters used in the above equation were obtained from NUREG-0133 and Regulatory

Guide 1.109, Rev. I.

VIRGINIA VPAP-2103 POWER REVISIONO PAGE 116 OF 116 ATTACHMENT 29 (Page 8 of 8) NORTH ANNA METEOROLOGICAL, LIQUID AND GASEOUS PATHWAY ANALYSIS Since the concentration of tritium in milk is based on the airborne concentration rather than the deposition, the following equation is used: RH-3 = K' K"' Fm Qp Uap (DFLH-3) [0.75 (0.5/H)] x X/Q where: K"' = a constant of unit conversion IE+o3 gm/kg H = absolute humidity of the atmosphere, 8.0, gm/m3 0.75 = the fraction of total feed that is water 0.5 = the ratio of the specific activity of the feed grass to the atmospheric water XIQ = the annual average concentration at a location 3250 meters N sector, 7 .2E-07 sec/m3 for ventilation vent releases, and 3.9E-07 sec/m3 for the process vent releases Other parameters have been previously defined.

Attachment 5 Process Control Program Virginia Electric and Power Company

Station Administrative VIRGINIA POWER Procedure

Title:

Radioactive Waste Process Control Program (PCP) Lead Department: Radiological Protection Procedure Number: ,Revision Number: Effective Date: VPAP-2104 0 05/31/90 Surry Power Station North Anna Power Station Approved by: Approved by: 1/t-~/fo 1;;;1c* Approved by: Date c/~J- 7 .- rJ Date

  ~lif~*

L

VIRGINIA VPAP-2104 POWER REVISIONO PAGE20F 16 TABLE OF CONTENTS Section Page 1.0 PURPOSE 3 2.0 SCOPE 3

3. 0 REFERENCE/COMMITMENT DOCUMENTS 3 4.0 DEFINITIONS 4

5. 0 RESPONSIBILITIES 6 6.0 INSTRUCTIONS 7 6 .1 General Descriptions and Requirements 7 6.1.1 Types of Wet Radioactive Waste 7 6.1.2 Waste Sources 7

6.1.3 Requirements for Processing Wet Radioactive Waste 6.1.4 Process Control Program Implementing Procedures 6.1.5 Requirements For Use of Contractor Services 6.2 Solidification of Wet Waste 8

8 9 10 6.2.1 Solidification Parameters 10 6.2.2 Adverse Chemical Reactions During Solidification 10 6.2.3 Sampling, Analysis, and Process Surveillance 11 6.2.4 Processing Acceptance Criteria 12

6. 3 Dewatering and Encapsulation of Filter Elements 12 6.3.1 General Requirements 12 6.3.2 Filter Elements to be Disposed of as Class A Waste 13 6.3.3 Filter Elements to be Disposed of as Class B or C Waste 13

6. 4 Reporting Requirements 14

           " ...6.4.l....Major.Changes to Radioactive_solid Waste.TreatmentSystems.       14 6.4.2 Changes to the Process Control Program (PCP)                        15 7 .0 RECORDS                                                                          16

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 30F 16 1.0 PURPOSE This procedure establishes Virginia Power's PROCESS CONTROL PROGRAM (PCP) including associated requirements and responsibilities. The PCP provides instructions for processing and packaging of wet radioactive wastes* to assure compliance with applicable Federal and State regulations for disposal of solid radioactive waste. 2.0 SCOPE This procedure is applicable to the processing and packaging of wet radioactive waste performed at or by the Station. Systems and procedures used for implementing the PCP, including vendor supplied systems and procedures, shall be considered a part of the PCP.

3. 0

REFERENCES/COMMITMENT DOCUMENTS 3 .1 References 3.1.1 10 CFR 20, Standards for.Protection Against Radiation 3.1.2 10 CFR50,.Domestic.Licensingof.Production and Utilization.Facilities - .

3.1.3. 10 CFR 61, .Licensing Requirements for Land Disposal of Radioactive Waste 3.1.4 10 CFR 71, Packaging and Transportation of Radioactive Material 3.1.5 49 CFR Parts 170 to 189, Department of Transportation Regulations for . Transportation of Hazardous Materials 3.1.6 USNRC Low-Level Waste Licensing, Branch Technical Position on Radioactive

               "' Waste Classification and Technical Position on Waste Form, May 1983, Rev 0 3.1.7     INPO 88-010, Guidelines for Radiological Protection at Nuclear Power Stations 3.1.8     NUREG-0800, USNRC, Standard Review Plan 11.4, Solid Waste Management Systems, Rev 2, July 1981 3.1.9

NRC Generic Letter 89-01, Implementation ofProgrammatic Controls for Radiological Effluent Technical Specifications (RETS)'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 3 .1.10 Surry and North Anna Technical Specifications 3.1.11 NODS-HP-01, Radiation Protection Plan 3.1.12 VPAP-0102, Station Nuclear Safety and Operating Committee

VIRGINIA VPAP-2104 POWER REVISIONO PAGE40F 16 3.1.13

VPAP-2101, Radiation Protection Plan 3.1.14 VPAP-2103, Offsite Dose Calculation Manual (ODCM) 3 .1.15 VPAP-3001, Safety Evaluations ( when issued) 3 .1.16 Chem-Nuclear Systems, Inc. Letter (concerning limitation of package void space),

October 6, 1989, GAR-196-89, [4605g]

3. 2 Commitment Documents None 4.0 DEFINITIONS NOTE: Terms which are defined in Surry and North Anna Technical Specifications appear as all capitalized letters in the text of this procedure for identification.

4.1 Batch A quantity of waste that kor may be mixed to produce a homogeneous mixture for the, *

    .. *.** purposes of sampling,.,testing, and processing.".Different samples,of a .homogeneous mixture are expected to exhibit similar.chemical and physical properties.

4. 2 Composite A mixture of samples, proportional by volume to the individual transfers making up a batch, that creates a test specimen representative of the batch.

4.3 Free Liquid Free liquid is the liquid still visible after solidification or dewatering is complete, or is drainable from the low point of a punctured container (NRC SRP 11.4, ETSB 11-3).

4. 4 High Integrity Container A container designed to provide long-term structural stability to contained waste during the required disposal period. May be used as an alternative to waste solidification. See section C.4 of NRC BTP C:Waste Form) for more details. High integrity containers must be approved by the appropriate agency.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE4AOF16 SUPPLEMENTAL REFERENCE PAGE This Supplemental Reference Page is provided to aid the procedure user in determining the appropriate procedures to use until such time that procedures referenced in the References Section, which reflect "When Issued", are approved and issued.

a. Upgraded Procedure Reference VPAP-3001, Safety Evaluations (When Issued)

The following existing procedures shall be used with respect to Safety Evaluations

           . until .such time that the new referenced procedure is approved and issued:

a. Surry

1. SUADM-LR-12, Safety Analysis/10CFR50.59/10CFR72.48 Safety Evaluations and Justifications for Continued Operations

b. North Anna

1. ADM-3.9, 10CFR 50.59 Safety Evaluation and JCOs (North Anna)

2. ADM-3.15, Tracking of Justifications for Continued Operation (JCO)

NOTE: This Supplemental Reference Page shall be removed and processed as directed upon notification from Records Management.

VIRGINIA VPAP-2104 POWER REVISIONO PAGES OF 16

4. 5

Non-Corrosive Liquid In lieu of specific tests, a liquid may be considered non-corrosive if it has a pH between 4 and 11 (based on section C.2.h of NRC BTP (Waste Form)).

4. 6 Process Control Program

        . The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling,

analyses, tests and determinations 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 a way that assures compliance with 10 CFR Parts 20, 61 and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

4. 7 Site Boundary The SITE BOUNDARY is that line beyond which the land is not owned, leased, or otherwise controlled by Virginia Power.

4. 8 Solidification Solidification is the conversion of wet waste into a form that meets shipping*and burial ground requirements.

4. 9 Spent Ion Exchange Material Organic resins and other ion exchange material are considered spent when decontamination

       * *factors-decrease significantly or when activity levels reach a pre-determined level.

4 .10 Stabilization or Stability A structurally stable waste form will generally maintain its physical dimensions and its form under the expected disposal conditions. Structural stability can be provided by the waste form itself, processing the waste to a stable form (e.g, solidify), or placing the waste in a disposal container or structure that provides stability after disposal (10 CFR 61.56(b)). 4 .11 Test Specimen A sample obtained from a batch of waste to be processed (solidified or absorbed), or a

         .. simulated sample of similar chemical and physical characteristics, on which a test can be performed to verify the intended process will perform satisfactory.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE60F 16 4 .12 Unrestricted Area

UNRESTRICTED AREA is defined as any area at or beyond the SITE BOUNDARY where access is not controlled by Virginia Power for purposes of protection of individuals from
exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional or recreational purposes.

4.13 Wet versus Dry Wastes (from NRC SRP 11.4, BTP . ETSB .11-3) Radioactive waste is generated in the forms of *~wet".. and "dry" wastes. Wet wastes, including spent ion exchange material, filter sludge, evaporator concentrates, and spent cartridge filter elements, normally are byproducts from liquid processing systems. Dry wastes, including activated charcoal, HEPA filters, rags, paper, and clothing, normally are byproducts from

        * * '* *'ventilation air and gaseous waste processing systems; and maintenance and refueling operations.

5. 0 RESPONSIBILITIES 5 .1 Health Physics Health Physics (HP) is responsible for:

5 .1.1 Implementing the PROCESS CONTROL PROGRAM as a part of the Radiation Protection Program. 5.1.2

Ensuring thatvendors*broughton site by Health Physics to perform waste processing

                        * *--are-cognizant of responsibilities in accordance with this procedure.

5.1.3 Maintaining procedures necessary for implementing the PCP.

5. 2 Operations Department The Operations Department is responsible for:

5.2.1 Implementing the PROCESS CONTROL PROGRAM as*part of normal Station** operations. 5.2.2 Ensuring that vendors brought on site by Operations to perform waste processing are cognizant of responsibilities in accordance with this procedure. 5.2.3 Maintaining procedures necessary for implementing the PCP . L_

VIRGINIA VPAP-2104 POWER REVISIONO PAGE70F 16 6 .0 *INSTRUCTIONS 6 .1 General Descriptions and Requirements 6.1.1 Types of Wet Radioactive Waste Wet radioactive wastes produced at the Station which must be processed for disposal include:

Resin
Filter elements
Waste oil
Liquid waste 6.1.2 Waste Sources a/ Station systems which normally process radioactive liquids with the subsequent
generation of *spent radioactive ion exchange bead resin and/or filter elements which must be processed for disposal are:
Primary Coolant System
Boron Recovery System
Spent Fuel Pit Purification System
Vent and Drain System
Liquid Waste Processing System

b. If primary to secondary leakage occurs while the Condensate Polishing System is processing secondary condensate, resin and filter elements used in the system may become radioactive. If so, they shall be processed for disposal.

              *c: If lubricatinglcooling*oil becomes contaminated *with radioactive material, and if the oil is to be disposed of as radioactive waste in* a* licensed land disposal facility, the oil shall be considered and processed as wet radioactive waste.

d. If liquid wet waste is produced which must be disposed of (e.g., evaporator bottoms or decontamination solutions) it shall be treated as wet radioactive waste .

VIRGINIA VPAP-2104 POWER REVISION 0 PAGE 8 OF 16

                  *6.1.3      Requirements for Processing Wet Radioactive Waste

a. Liquids which are to be processed as radioactive waste shall be processed by solidification.

b. Resins shall be processed by dewatering and/or-solidification.

c. Filter elements shall be processed by.dewatering or encapsulation in a solidification binder.

d. Waste oil shall be processed by solidification or transferred to a licensed waste processor for disposal .

                         . e .. Class B and Class C waste shall be stabilized prior to disposal (10 CFR 61).

f

Certain categories of Class* Awaste:shall be stabilized prior to disposal as. required
by the disposal *site and/or the disposal site license conditions.

6.1.4 Process Control Program Implementing. Procedures

a. Health Physics shall maintain procedures necessary to implement the PCP.

Procedures shall include acceptable methods for:

1. -Radioactive waste sampling, analysis and waste classification. Waste

                                              .. classification shallbe performed per 10 CFR 61.55, Waste Classification, and

methods set forth in NRC BTP on Radioactive Waste Classification.

2. Radioactive waste processing including waste solidification and stabilization.

Acceptance criteria shall meet criteria set forth in:

10 CFR 61.56, Waste Characteristics
NRCBTPonWasteForm
Disposal site criteria

3. Radioactive waste packaging and shipping. Acceptance criteria shall meet requirements set forth in:

10 CFR 20.311, Transfer for Disposal and Manifests Y,; *.,,:"',. : .. * -'"';, .* :

1 10 *CFR7 l";Packaging*and <Transporting ofRadioactive Material

                                        '." 49 CFR 170- 189,.Transportation of Hazardous Materials

VIRGINIA VPAP-2104 POWER REVISIONO PAGE90F 16

b. Operations Department shall maintain procedures necessary to implement the PCP.

Procedures shall include acceptable methods for dewatering ion exchange resin. 6.1.5 Requirements For Use of Contractor Services The following actions shall be taken before a contractor-supplied waste processing system is used on site:

a. Obtain the following, as a minimum, for review and evaluation:

A detailed system description, which may be included in a topical report or equivalent documentation
System operating procedures, which include process control parameters
A list of required physical interfaces and Station materials/services
A list of chemicals to-be brought on.site, quantity.to be used and.material safety data sheets for each chemical

A list of expected utility/contractor responsibilities including disposal of unused and contaminated chemicals
Vendor's document control procedures/manual to ensure controls are in place which prohibits use of procedures not approved by Station Nuclear Safety Operating Committee (SNSOC)

b. Compare the system description and operating procedures to the requirements

              ... provided in Subsection 6.2, Solidification of Wet Waste. Ensure that the system can be operated within requirements.

c. Submit system operating procedures to SNSOC for review and approval in accordance with VPAP-0102, Station Nuclear Safety and Operating Committee.

Processing of radwaste shall not be performed without approved operating procedures.

d. Ensure the contractor provides a system as proposed, described, and approved for use at the Station.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE lOOF 16

6. 2

Solidification of Wet Waste

        *Procedures used for wet waste solidification shall incorporate the following requirements:

6.2.1 Solidification Parameters

a. As appropriate; parameters used when performing solidification may include, but are not limited to:

Waste type Q WastepH
Ratios of waste/liquid to solidification agent/catalyst
Waste oil content
Waste principal chemical constituents
Mixing and curing times

b. Once established, solidification parameters shall provide-boundary conditions to ensure that:

Solidification is complete
Requirements* for waste form stability are met
There are no detectable free standing liquids 6.2.2 Adverse Chemical Reactions During Solidification Adverse chemical reactions between waste contaminants and solidification agents may not be noticeable during specimen tests performed to develop solidification parameters.

To preclude such adverse chemical reactions, the following shall be performed prior to initial solidification of wet radioactive waste : NOTE: Performance of this subsection is not required if solidification is to be performed.by a vendor and results of such testing performed by the vendor was included in a technical report describing the proposed solidification methodology.

a. Prepare large volume (e.g., 1 or 2 gallons) non-radioactive mixtures of the waste stream chemicals potentially present (e.g., resin beads, boric acid, acids, bases, detergents, decontamination solutions) .

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 11 OF 16

b. Solidify the mixture.

1. The mixture shall be solidified using solidification procedure and parameters prepared for specified waste stream.

2. The solidification shall be performed.within an insulated container to simulate the restricted heat removable capability of larger containers.

c. Ensure the mixture solidifies without*generatingexcessive temperatures or gases.

6.2.3 Sampling, Analysis, and Process Surveillance Wet radioactive waste shall be processed strictly in accordance with the approved

solidification procedure for the specific waste stream substances to be solidified.

Waste shall be sampled, analyzed, and compared to solidification parameters.

                     *a~. Results of sampkanalysis*shallbe recorded on waste processing.data sheets.

b. A representative test specimen from at least every tenth batch of each type of waste to be solidified*shall.be used to verify solidification.. If any test specimen fails to solidify:

1. Solidification of the batch under test shall be suspended until such time as:

Additional samples can be obtained
Alternative solidification parameters can be determined
Subsequent tests verify solidification

2. Solidification of the batch may then be resumed using the alternative solidification parameters determined.

3. A representative test specimen shall be obtained from each subsequent batch of the same type of waste and test solidification performed.

4. Collection and testing ofrepresentative testspecimens from each consecutive batch shall continue until three consecutive initial test specimens demonstrate solidification.

      * .......... -C~- .. Jf.necessary,  procedures .shall.be revised to.ensure:solidification of subsequent batches of waste.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 120F 16

                           . d: Ifprc5visions*oflhe-PROCESS CONTROL PROGRAM cannot be satisfied,
                                 * ,,suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

6.2.4 Solidification Acceptance Criteria NOTE: The following are general considerations. Specific site disposal criteria must be addressed based on the site to be used. Procedures for wet radioactive waste solidification shall incorporate the following requirements:

a. Containers for processed waste shall be filled to at least 85% of capacity. If a

                                    -Container is processed to.less than 85% of capacity, it shall not be shippedfor
                                   . disposal prior to approval from the disposal site.

b. Solid waste that contains liquid shall have as little free-standingJiquid as is...

reasonably achievable, but in no case shall the liquid exceed 1% of the volume. The liquid shall be noncorrosive.

1. If a high integrity*container is not used, the maximum free liquid is 0.5% of the waste volume.

2. If a high integrity container is used, the maximum free liquid is 1.0% of the waste volume.

6. 3 Dewatering and Encapsulation of Filter Elements NOTE: Filter elements are normally mechanical filters .with-wound fiber cartridges used for removing particulates from liquid systems. This procedure is only applicable. to filter elements which are of the cartridge type.

6.3.1 General Requirements

    -. ---,,, -- .... ----" "a ......Spentfilter.dements.removed from systems.shall ..be_~placedin .appropriate storage to await processing and shipment.

b. Processing of spent filter elements shall be based on waste classification of filter.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 13 OF 16 NOTE: The following are general considerations. Specific site disposal criteria must be addressed based on the site to be used. 1; If filter media-is classified.as Class A\waste and does not contain nuclides with half-lives greater than 5 years which have a total specific activity of 1 µCi/cc or greater, it may be disposed of as Class A waste.

2. If filter media is classified as Class B or Class C waste (per 10 CFR 61.55), it shall be encapsulated in a solidification media prior to disposal or disposed of in a high integrity container (NRC BTP, C.5 (Waste Form)).

6.3.2

Filter Elements to be Disposed of as Class A Waste

a. Filters should be allowed to drain dry in such a manner that any liquid trapped in

                 ** voids is allowed to drain.
             . b. Filters shall not be compacted unless they.are first allowed to dry essentially free of moisture .

c. If moist filters are to be packaged without compaction:

1 . There shall be no indication of moisture on the filter in the form of drops or surface wetness .

                   . 2. Place filters in a container or plastic bag to which absorbent material has been placed to absorb unintentional and incidental amounts of liquids. The amount of absorbent material should be equal to at least one-fourth the volume of filter.

d. Ensure documentation indication package contents describes the presence of filters.

6.3.3 Filter Elements to be Disposed of as Class B or C Waste

a. If filters are to be solidified by being encapsulated in a solidification media:

1. Place filters in a suitable container such that filters will be completely surrounded by the solidification media when added. A basket type

                      "* arrangement of thin wire is recommended to hold filters in a fixed geometry.

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 140F 16 NOTE: The solidification media, including absence of free liquid, must be tested and documented in a manner required for solidification described in subsection 6.2, Solidification of Wet Waste.

2. Fill container with solidification media until filters are completely covered and container is filled to at least 85% of capacity.

3. Place solidified filter container in container appropriate for shipping and disposal at specified disposal site. A high integrity container is recommended to ensure compliance with all requirements.

b ..* If an encapsulated filter is to be disposed of in a high integrity container, properly place the container with the encapsulated filter in a high integrity container.

c. If anun-encapsulated filter-is to'be disposed of in a high integrity container:

1. Place filters in container such that fi1 ters will be held in a fixed *geometry and such that liquids will not be trapped within filters. A basket type arrarigement of thin wire is recommended to hold filters provided container's Cof C will not be violated.

                               '2. If resin will be added;proceed with resin addition as appropriate.

3. Dewater the container, as applicable.

6. 4 Reporting *Requirements 6.4.1 Major Changes to Radioactive Solid Waste Treatment Systems NOTE: Information required by this subsection to be reported to the NRC may be submitted as part of the annual FSAR update.

Major changes to the radioactive solid waste systems:

a. Shall become effective upon review and acceptance by SNSOC.

b. Shall be reported to the NRC in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by SNSOC. The discussion of each change shall contain:

   .*:(*f .

VIRGINIA VPAP-2104 POWER REVISIONO PAGE 15 OF 16

1. A summary of the evaluation* that led to the determination that the change could

be made in accordance with 10 CFR Part 50.59; Such evaluations shall be made in accordance with VPAP-3001, Safety Evaluations.

2. Detailed information sufficient to totally support the reason for-the change without benefit of additional or supplemental information.

3. A detailed description of equipment, components, and processes involved and interfaces with other plant systems.

4. An evaluation of the change, in quantity of solid waste differing from that previously predicted in the license application and amendments to the application.

5.- An evaluation of the change, which shows the expected maximum exposures to individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license and amendments.

6. A comparison of the predicted releases of radioactive materials, in solid waste, to the actual releases.for-the*period prior to the changes.

7. An estimate of the exposure to plant operating personnel as a result of the.

change.

8. Documentation of SNSOC review and approval.

6.4.2 Changes to the Process Control Program (PCP)

        * * * *°Changes to the PCP shall be:

a. Documented; reviews shall be retained as Station records. Documentation shall include:

1. Information to support the change together with.the appropriate analyses or evaluations justifying the changes.

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.

              ,h..,.Reviewed and.approved by SNSOC and-Plant-Manager prior to implementation.

v .I VIRGINIA VPAP-2104 POWER REVISIONO PAGE 160F 16 7.0 RECORDS The following individual/packaged documents and related correspondence completed as a result of the performance or implementation of this procedure are records. Records shall be transmitted to Records Management in accordance with VPAP-1701, Records Management. PROCESS CONTROL PROGRAM records shall include, but are not limited to:

System description of any contractor's temporary processing system. Such-a description may be provided in a topical report or other equivalent documentation
Approved solidification system operating procedures
Data sheets used to record solidification data, including test specimen data

            * ..Records of reviews performed for changes made to the PROCESS CONTROL PROGRAM

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

Sena\ . , ~ . ~ -

L.: - 09 3 January 31, 1989 Rec'd~ ffB O 91989 Nuclear Operations Licensing Supervisor TO ALL POWER REACTOR LICENSEES ANC APPLICANTS

SUBJECT:

1MPLEMEP1T1'TI0N OF PROGRAMMATIC CONTROLS FOR ((ADIOLOGlC'Al EFFLUENT TECHNICAL SPEC1FICAT10NS IN THE ADMINISTRATIVE CONTROLS SECTION OF THE TEChNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF ~ETS TO THE OFFSITE DOSE CALCULATION MANUAL OR TO THE PROCESS CO~TROL PROGRAM (GENERIC LETTER 89-01) . The NRC staff has examined the contents of the Radiological Effluent Technical Specifications (RETS) in relation to the Corrmission's Interim Polit\* Statement on Technical Specification Improvements. The staff has detennined that pro-gramr.1atic controls can be implemented in the Administrative Controls section of the Tcthnical Specifications (TS) to sati~tv existing regulatorv reauirements for RETS. At the same time, the procedural d1::tails of the current TS on radio-

active effluents and radio1ogical environmental a1onitoring can be relocated to the Offsite Dose Calculation Manual (CDCM). Like~ise. the procedural details of the current TS on ~clid radioactive wastes can be relocated to the Process Control Prograrr, (PCP). lhese actior.s simplih the RETS meet the regulatorv 1

reauirements for radioactive effluents and ,*adiological environmental monitor-ing, and are provideo a~ a line-item irnprovem=nt of the TS, co11sistent with the gcals of the Policy Statement. New prograrrmatic controls fer radioactive effluents and radiological environ-mental monitoring are incorporated in the TS to conforr., to the regulatorv reouirements of 10 CFR 20.10~. 40 CFR Part ]90 1 10 CFP. 50.36a. and Appendix I tc 10 CFR Part 50. Existing programatic recuirements for the PCP are being retained in the TS. The procedura1 details i~cluded in licensees' present TS on radioactive effluents, solid radioactive wastes, environmental monitoring, and associated reporting recuirements will be relocated to the ODCM or PCP as appropriate. Licensees will handle future changes to these procedural details in the OOCM and the PCP under the administrative controls for changes to the O~CH or PCP. Finally, the definitions of the ODC~ and PCP are updated to reflect these changes. Enclosure 1 provides guidance for the preparation of a license amendment re-ouest to implement these alternatives for RETS. Enclosure 2 provides a list-ing of existing RETS and a description of h0\11 each is addressed. Enclosure 3 pr~vides model TS for progrinnatic controls for RE1S and its associated report-ing reauirements. Finallv, Enclosure 4 provides model specifications for retairiing existing reauirements for exp1osive gas monitoring instrumentation recuirements that apply on a plant-specific basis. licensees are encouraged to propose changes to rs* that are consistent with the guidance provided in the enclosures. Cor.fonning atr,endment recuests will be expediticuslv revie~ed bv

Generic Letter 89-01 2 January 31, 1989 the NRC Project Manager for the facility. Proposed amendments that deviate from this guidance will require a longer, more detailed review. Please contact the appropriate Project Manager if you have questions on this matter. Sincerely,

Enclosures:

                                   ~~~or                      for Projects Office of Nuclear Reactor Regulation l through 4 as stated

Generic Letter 89-01 ENCLOSURE 1

GUIDANCE FOR THE IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RETS IN THE ADMINISTRATIVE CONTROLS SECTION OF TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF CURRENT RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR PROCESS CONTROL PROGRAM INTRODUCTION This enclosure provides guidance for the preparation of a license amendment request to implement progranvnatic controls in Technical Specifications (TS) for ~adioactive effluents and for radiological environmental monitoring con-forming to the applicable regulatory requirements. This will allow the reloca-tion of existing procedural details of the current Radiological Effluent Technical Specifications (RETS) to the Offsite Dose Calculation Manual (ODCM).

Procedural details for solid radioactive wastes will be relocated to the Process Control Program (PCP). A proposed amendment will (1) incorporate pro-grammatic controls in the Administrative Controls section of the TS that sat-isfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a. and Appendix I to 10 CFR Part 50, (2) relocate the existing procedural details in current s~ecifications involving radioactive effluent monitoring instrumenta-tion, the control of liquid and gaseous effluents, equipment requirements for liquid and gaseous effluents, radiological environmental monitoring, and radio-logical reporting details from the TS to the ODCM, (3) relocate the definition of solidification and existing procedural details in the current specification on solid radioactive wastes to the PCP, (4) simplify the associated reporting requirements, (5) simplify the administrative controls for changes to the ODCM and PCP, (6) add record retention requirements for changes to the ODCM and PCP, and (7) update the definitions of the ODCM and PCP consistent with these changes. The NRC staff's intent in recommending these changes to the TS and the reloca-tion of procedural details of the current RETS to the ODCM and PCP is to ful-fill the goal of the Commission Policy Statement for Technical Specification Improvements. It is not the staff's intent to reduce the level of radiol~gical effluent control. Rather, this amendment will provide progranvnatic controls for RETS consistent with regulatory requirements and allow relocation of the procedural details of current RETS to the OOCM or PCP. Therefore, future changes to these procedural details will be controlled by the controls for changes to the OOCM or PCP included in the Administrative Controls sectiCM'l of the TS. These procedural details are not required to be included in TS by 10 CFR 50.36a. DISCUSSION. Enclosure 2 to Generic Letter 89-01 provides a summary listing of specifica-tions that are included under the heading of RETS in the Standard Technical Specifications (STS) and their disposition. Most of these specifications will be addressed by programmatic controls in the Administrative Controls section of the TS. Some specifications under the heading of RETS are not covered by the new progranvnatic controls and will be retained as requirements in the existing plant TS. Examples include requirements for explosive gas monitoring instru-

mentation, limitations on the quantity of radioactivity in liquid or gaseous holdup or storage tanks or in the condenser exhaust for BWRs, or limitations on explosive gas mixtures in offgas treatment systems and storage tanks.

- Generic Letter 89- 01 Enclosure 1 licensees with nonstandard TS should follow the guidance provided in Enclo-sure 2 for the disposition of similar requirements in the format of their TS.

Because solid radioactive wastes are addressed under existing programmatic controls for the Process Control Program, which is a separate program from the new programmatic controls for liquid and gaseous radioactive effluents, the requirements for solid radioactive wastes and associated solid waste reporting requirements in current TS are included as procedural details that will be relocated to the PCP as part of this line-item improvement of TS. Also, the staff has concluded that records of licensee *reviews performed for changes made to the ODCM and PCP should be documented ~nd retained for the duration of the unit operating license. This approach is in lieu of the current requirements that the reasons for changes to the ODCM and PCP be addressed in the Semiannual

  .Effluent Release Report.

The following items are to be included in a license amendment request to imple-ment these changes. First, the model specifications in Enclosure 3 to Generic Letter 89-01 should be incorporated into the TS to satisfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50. The definitions of the ODCM and PCP should be updated to reflect these changes. The programmatic and reporting requirements are general in nature and do not contain plant-specific details. Therefore, these changes to the Administrative Controls section of the TS are to replace corresponding requirements in plant TS that address these items. They should be proposed for incorporation into the plant's TS without change in substance to replace existing requirements. If necessary, only changes in format should be proposed. If the current TS include requirements for explosive gas monitoring instrumentation as part of the gaseous effluent monitoring instrumentation requirements, these require-ments should be retained. Enclosure 4 to Generic Letter 89- 01 provides model specifications for retaining such requirements. Second, the procedural details covered in the licensee's current RETS, consist-ing of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM or PCP as appropriate and in a manner that ensures that these details are incorporated in plant operating pro-cedures. The NRC staff does not intend to repeat technical reviews of the re-located procedural details because their consistency with the applicable regula-tory requirements is a matter of record from past NRC reviews of RETS. If licensees make other than editorial changes in the procedural details being transferred to the ODCM, each change should be identified by markings in the margin and the requirements of new Specification 6.14a.(l) and (2) followed. Finally, licensees should confirm in the amendment request that changes for relocating the procedural details of current RETS to either the ODCM or PCP have been prepared in accordance with the proposed changes to the Administra-tive Controls section of the TS so that they may be implemented immediately upon issuance of the proposed amendment. A complete and legible copy of the

revised ODCM should be forwarded with the amendment request for NRC use as a reference. The NRC staff will not concur in or approve the revised ODCM.

Enclosure 1 Generic Letter 89-01 Licensees should refer to NGeneric Letter 89-0lu in the Subject line of license amendment reauests implementing the guidance of this Generic Letter. This will facilitate the staff's tracking of licensees' responses to this Generic Letter. SUMMAR't' The license amendment reauest for the li~e-item improvements of the TS relative to the RETS will entail (1) the incorporation of progrannatic controls for radioactive effluents and radiological environmental ~onitoring in the Admin-istrative Controls section of the TS, (2) incorporatation of the procedural details of the current RETS in the OOCM or PCP as appropriate, and (3) confirm-ation that the guidance of this Generic Letter has been f~llowed

DISPOSITION OF SPECIFICATIONS AND ADMINISTRATIVE CONTROLS INCLUDED UNDER fkt HEADING OF RETS lN T~[ STANDARfi lECHkICAL SPECIFIC~TIONS SPEC IF ICATION TITLE ~1S~OSITIO~ OF EXISTl"G SPECIFICATION -'*

n OFFSITE DOSE CALCULATION l-1ANUAL Defir.1tion is updated to reflect the change in scope r-1.17 of the ODCM. .... 111

..+

I'll 1.22 PROCESS CONTROL PROGRAM Definition is updated t~ reflect the change in scope ""' OD of the PCP. '° I 0 1.32 SOLIDIFICAlJON Def 1nit1on is relocated to the PCP. 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT Progranmatic controls are included in 6.8.4 g. lte11 1). MONITORING IHSTRUMENTATION Existing specificatfon procedural details are relocated to the ODCH. 3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT Progrannatic controls are included in 6.8.4 g. Item 1). MOMITORING INSTRUMENTATION Existing specification procedural details are relocated to the OOCM. Existing reouirements for explosive gas monitoring instrumentation should be retained. Model specifications for these requirements are provided in Enclosure 4. 3/4.11.1.1 LIQUID EFFLUE~TS: CONCENTRATION Progrannatic controls are included in 6.8.4 g. Items 2) and 3). Existing specification procedural details are relocated tu the ODCM. 3/4.11.1.2 LIQUID EFFLUENTS: DOSE Progrannatic controls are included fn 6.8.4 g. Items 4) and 5). Existing specification procedural details are relocated to the ODCM. 3/4.11.1.3 LIQUID EFFLUENTS: LIQUID Progrannatic controls are included in 6.8.4 g. Jtetn 6). R~DWASTE TREATMENT SYSTEM Existinq specification pro~edural details are relocated to the ODCM. ,,, Existing specification requfretnents to be retained. n...,. 3/4.11.1.4 LIQUID HOLDUP TANKS 0 C I'll N

tJSPOSITION OF SPECIFICATIONS AND ADMINISTRATIVf CONTROLS INCLUDED UNDER THE HEADING OF RETS IN THE STANDARD TECHNICAL SPECIFICATIC~S !Cont.) SPECIFICATION TITLE DISPOSITION OF EXISTING SPECIFICATION 3/4.11.2.1 GASEOUS EFFLUENTS: DOS£ RATE Progrannatic controls are included in 6.8.4 g. Items 3) and 7). Existing specification procedural details are relocated to the ODCM. 3/4.11.2.2 GASEOUS EFFLUENTS: DOSE-HOBLE Progra111J1atic controls are incl~ded in 6.8.4 g. Items 5l GASES and 8). Existing specification procedural details are. relocated to the OOCM. 3/4.11.2.3 GASEOUS EFFLUENTS: DOSE--IODINf- Progra11111atic controls are included in 6.8.4 g. Items 5) 131. IODJNE-133. TRITIUM. ANO and 9). Existing specification procedural details are RADIOACTIVE MATERIAL IN PARTICU- relocated to the OOCM. LATE FORM 3/4.11.2.4 GASEOUS EFFLUENTS: GASEOUS Progrannatic controls are incfoded in 6.8.4 g. Item 6). RADWASTE TREATMENT or Existing specification procedural details Bre relocated VENTILATION EXHAUST TREtTM£NT to the OOCM. N SYSTEM 3/4.11.2.5 EXPLOSIVE GAS MIXTURE Existing specification requirements should be retained. 3/4.11.?.6 GAS STORAGE TANKS Existing specification requirements should be retained. 3/4.11.2.7 MAIN CONDENSER (8~Pl Existing specification reouirements should be retained. 3/4.11.2.8 PURGING AND VENTING (BWR Mark II Pro9rannatic controls are included in 6.8.4 g. Item 10). containments) Existing specification procedural details are relocated tu the ODCM. 3/4.11.3 SOLID RADIOACTIVE WASTES Existing specification procedural details are relocated tu the PCP. ,,, 3/4.11.4 RA~IOACTIVE EFFLUENTS: TOTAL Progra111J1atic controls are included in 6.8.4 g. Item Ill. n

                                                                                                              ..J DOSE                               Existing specification procedural details are relocated      0 Vt to the OOCM.                                                ..,

C It)

DISPOSITION OF SPECIFICATIONS AND ADMINISTRATIVE tONi'ROLS . er, INCLUDED UNDEff TR£ ~[ADING OF RETS IN THt STANDARD TECHNICAL SPECIFICATIONS (Cont.) n, n, SPECIF I CAT ION TITLE DISPOSITION OF EXISTING SPECIFICATION n r-3/4.12.1 RADIOLOGICAL ENVIRONMENTAL PrograR111atic controls are included tn 6.8.4 h. Item 1). ffl MONITORING: MONITORING PROGRAM Existing specification procedural details are relocated ..... to the OOCM. .., n, 0, U) 3/4.lL.2 RADIOLOGICAL ENVIRONMENTAL Progranmatic controls are included in 6.8.4 h. Item 2l. I MONITORING: LANO USE CENSUS Existing specification procedural details are relocated C> to the OOCM. 3/4.12.3 RADIOLOGICAL ENVIRONMENTAl Pro~ralllllatic controls are included tn 6.8.4 h. Item 3). MONITORING: INTERLABORATORY Existing specification procedural details are relocated COMPARISON PROGRAM to the OOCMo 5.1.3 DESIGN FEATURE~: SITE - MAP Existing specification reouirements should be retained. DEFINING UNRESTRICTED AREAS ANO SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS w fi.9.1.3 REPORTING REQUIREMENTS: ANNUAL Specification simplified and existing reporting details RADIOLOGICAL ENVIRONMENTAL are relocated to the ODCM. OPERATUIG REPORT 6.9.1.4 REPORTING REQUIREMENTS: SEMI- Specification simplified and existing reporting details ANNUAL RADIOACTIVE EFFLUENT are relocated to the OOCM or PCP as appropriate. RELEASE REPORT 6 .13 PROCESS CONTROL PROGRAM Spee if i cat ion rieou irement s are* *s i 111p 11 fi ed. 6.14 OFFSITE OOSE CALCULATION MANUAL Specification reouirements are simplified. 6.15 MAJOR CHANGES TO LIQUID, GASEOUS. Existing procedural details are relocated to the OOCM or ,.., AND SOLID RADWASTE TREATMENT PCP as appropriate. ::, n SYSTEMS

                                                                                                                      .,,0 C
                                                                                                                   /:

Generic Letter 89-01 Enclosure 3 TECHNICAL SPECIFICATIONS TO BE REVISED 1.17 DEFINITIONS: OFFSITE POSE CALCULATION MANUAL 1.22 DEFINITIONS: PROCESS CONTROL PROGRAM 6.8.4 g. PROCEDURES AND PROGRAMS: RADIOACTIVE EFFLUENT CONTROLS 6.8.4 h. PROCEDURES AND PROGRAMS: RADIOLOGICAL ENVIRONMENTAL MONITORING 6.9.1.3 REPORTING REQUIREMENTS: ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.4 REPORTING REQUIREMENTS: SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.10 RECORD RETENTION 6.13 PROCESS CONTROL PROGRAM (PCP) 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)

1.0 DEFINITIONS MODEL TECHNICAL SPECIFICATION REVISIONS (To supplement or replace existing specifications)

OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (OOCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Pro-grams required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4. 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ing 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.

Generic Letter 89- 01 Enclosure 3

6.0 ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS

( 6.8.4 The following programs shall be established, implemented, and maintained:

g. Radioactive Effluent Controls Program A program shall be provided confoMT\ing with 10 CFR S0.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall in-clude remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Limitations on the operability o~ radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to

3) 10 CFR Part 20, Appendix B, Table II, Column 2, M6nitoring, sampling, and analysis of ~adioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conform-ing to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in .accordance with the methodology and parameters in the ODCM at least every 31 days, *

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of ~hese systems are used to reduce releases of radio-activity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) . Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY

conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column l,

Generic Letter 89-01 Enclosure 3

ADMINISTRATIVE CONTROLS 6.8.4 g. Radioactive Effluent Controls Program (Cont.)

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on venting and purging of the Mark II containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable (BWRs w/Mark II containments),

and

11) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190 .

h. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radio-nuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental expo-sure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the method-ology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and

3) Participation in* a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance pro-gram for environmental monitoring .

Generic Letter 89-0l Enclosure 3

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.3 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May l *of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the OOCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part SO. SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT** 6.9.1.4 The Semiannual Radioactive Effluent Release Report covering the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July l of each year. The report shall in-clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in con-formance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR Part SO .

6.10 RECORD RETENTION 6.10.3 The following records shall be retained for the duration of the unit o.

Operating License: Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM. 6.13 PROCESS CONTROL PROGRAM (PCP) Changes to the PCP:

a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall contain:

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

 *A single submittal may be made for a multi-unit station.
 **A single submittal may be made for a multi-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.

Generic Letter 89-01 Enclosure 3

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (Cont.)

2) A determination that the change will maintain the overall con-formance of the solidified waste product to existing require-ments of Federal, State, or other applicable regulations.

b. Shall become effective after review and acceptance by the [URG] and the approval of the Plant Manager.

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall contain:

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 level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after review and acceptance by the [URGJ and the approval of the Plant Manager.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented .

Generic Letter 89- 01 Enclosure 4

INSTRUMENTATION MODIFICATION OF THE SPECIFICATION FOR RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TO RETAIN REQUIREMENTS FOR EXPLOSIVE GAS MONITORING INSTRUMENTATION EXPLOSIVE RABi9AfiVE GASE8ij5-EFftijENf MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION explosive 3.3.3.11 The rad;oact;ve gaseoas-effiaent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Specifications-37117£71-and 3.11.2.5 are not exceeded. ;he-A1armffr;p-Setpo;nts-cf-these-channe1s-meet;ng-Speeifieation 97 H. 7 f-: i- she, ,-be-determined-and-adj t1sted-; n-accordance-w; th-the-methodo, ogy and-parameters-;n-the-8BM-:

APPLICABILITY: As shown in Table 3.3-13 ACTION: explosive

a. With an rad;oact;ve gaseoas-effiaent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the

b.

above specification;-immediate,y-saspend-the-reiease-of-radioective geseot1s-eff1t1ents-mon;tored-by-the-affected-channe1;-cr declare the channel inoperable and take the ACTION shown in Table 3.3-13. explosive With less than the minimum number of redioect;ve gaseot1s-eff1t1ent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful expta;n-;n-the-next-Semi-annt1a1-Radioactive-Eff1aent-Re1ease-Report prepare and submit a Special Report to the Commission pursuant to Specification 6-:9-:i-:4 6.9.2 to explain why this inoperability was not corrected in a timely manner.

c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS explosive 4.3.3.11 Each radioact;ve gaseoas-effiaent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, 58ijRE EHEK; CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9 .

Sample STS 3/4 3-(n)

TABLE 3.3-13 EXPtO~lVl RAQJQA,tl¥~ GAS~QUG ~~~bY~Nl MONITORING JNSTRU~[NTATION MlNIMll~ CllANf~[l c; INSTRUMENT OPERABLE APPL 1cie1L 1TY ACTtON

1. (Not 1.1sefil LA. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring S)stem (for systems designed to withstand the effects of a hydrogen explosion)

a. Hydrogen Monitor (Automatic Control) 1 49

b. Hydrogen or Ox_ygen Moriitor (Process) 1 ** 49 w

w

. i:,.

I t:8. WASTE GA5 HOLDUP SYSTlM Explosive Gas Monitoring S.vstcm (for s.vstems not designed to withstand the effects of a hvdrogen I'>

 +          e,cplosion)

a. Hydrogen Morlitors (Automatic Control.

redundant) 2 ** 50. 52

h. Hydrogen or Oxygen Monitors (Process. 2 ** 50 dual) n_.

0

                                                                                                            "'C""1 It'
                                                                                                             ~

Generic Letter 89-01 Enclosure 4

~

*~

(Not used) TABLE 3.3-13 (Continued) During WASTE GAS HOLDUP SYSTEM operation. ACTION STATEMENTS ACTION 45 - {Not used) ACTION 46 - {Not used) ACTION 47 - {Not used) ACTION 48 - {Not used) ACTION 49 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement. operation of this WASTE GAS HOLDUP SYSTEM may continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. ACTION SO - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement. operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours. With both channels inoperable, operation may continue provided grab samples are taken and analyzed at least once per 4 hours during degassing operations and at least once per 24 hours during other operations. ACTION 51 - (Not used) ACTION 52 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner .

Sample STS 3/4 3-(n+2)
TABLE 4.3-9 Vt a, 0, n, EXPLOSIVE 3 ::3

",:J n, RAbl9AGlJV~ GAS~QY, ~FF~Y~Nl MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .., n, n V,

V, 11) CHANtiEl MODES FOR WHICH r+ r+ CHAtlNEl SQIIRG~ CHANNEL OPE RAT IOtJAL SURVEILLANCE .., n, INSTRUMENT CHECK GM~GK CALI BRATIOf' TEST _JS REquIRrL. (X)

1. (Not used)

                                                                                                                        "'I 0

_.j 2A. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring Svstem (for svstems designed to withstand the effects of a hydrogen explosion)

a. Hydrogen Monitor D N.. A... 0(4) M **

w (Automatic Control)

w h. Hydrogen or Oxygen Monitor 0 N.. A... Q(4) or Q(5) ** ~

-~

w I

(Process)

28. WASTE GAS HOLDUP SYSTEM E1plosive Gas Monitoring Svstem (for systems not designed to withstand the effects of a hydrogen e,plosion)

a. Hydrogen Monitors D N... A... 0(4) M **

(Automatic Control. redundant)

b. Hvdrogen or Oxygen Monttors 0 N... A... Q(4) or Q(5) M **

(Process. dual)

3 n

0 1/t C l'I)

Generic Letter 89- 01 Enclosure 4 TABLE 4.3-9 (Co~tinued) TABLE NOTATIONS (Not used) During WASTE GAS HOLDUP SYSTEM operation. (1) (Not used) (2) (Not used) (3) (Not used) (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

a. One volume percent hydogen, balance nitrogen, and

c. Four volume percent hydrogen, balance nitrogen.

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

a.

b. One volume percent oxygen, balance nitrogen, and Four volume percent oxygen, balance nitrogen .

Sample STS 3/4 3-(n+4)

L

LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Date of

Letter No. Sub.iect Issuance Issued To 88-20 INDIVIDUAL PLANT 11/23/88 ALL LICENSEES HOLDING EXAMINATION FOR SEVERE OPERATING LICENSES ACCIDENT VULNERABILITIES - AND CONSTRUCTION 10 CFR 50.54(f) PERMITS FOR NUCLEAR POWER REACTOR FACILITIES 88-19 USE OF DEADLY FORCE BY 10/28/88 ALL FUEL CYCLE. FACILITY LICENSEE GUARDS TO PREVENT LICENSEES WHO POSSESS, THEFT OF SPECIAL NUCLEAR USE, IMPORT, EXPORT, MATERIAL OR TRANSPORT FORMULA OUA~TITIES OF STRATEGIC SPECIAL NLlCLEAR MATERIAL 88-18 PLANT RECORD STORAGE ON 10/20/88 ALL LICENSEES OF OPTICAL DISKS OPERATING REACTOP.~

ANC HOLDERS OF CONSTRUCTION PERMITS 88-17 LOSS OF DECAY HEAT REMOVAL 10/17/88 ALL HOLDERS OF 10 CFR 50.54(f) OPERATI~G LICENSES OR CONSTRUCTIOt-: PERMITS FOR I PRESSURIZED WATER RE.ACTORS 88-16 REMOVAL OF CYCLE-SPECIFIC 10/04/88 ALL POWER REACTO~ PARAMETER LIMITS FROM LICENSEES ANG TECHNICAL SPECIFICATIONS APPLICANTS

     .88-15      ELECTRIC POWER SYSTEMS -       09/12/88       ALL POWER REACTOP INADEQUATE CONTROL OVER                       LICENSEES AND DESIGN PROCESSES                              APPLICANTS 88-14       INSTRUMENT AIR SUPPLY          08/08/88       ALL HOLDERS OF SYSTEM PROBLEMS AFFECTING                      OPERATING LICENSES SAFETY-RELATED EQUIPMENT                       OR CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-13      OPERATOR LICENSING              08/08/88       ALL POWER REACTOK EXAMINATIONS                                  LICENSEES AND APPLICANTS FOR AN OPERATING LICENSE

88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS FROM TECHNICAL LlCENSEES ANu
SPECIFICATIONS APPLICANTS I}}

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ML18151A257: Difference between revisions

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