ML20062N080

Revisions 1 to Proposed Tech Spec Changes,Enhancing Radiological Effluent & Environ Monitoring

ML20062N080
Person / Time
Site:	Prairie Island
Issue date:	08/13/1982
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20062N069	List: ML20062N067 ML20062N076 ML20062N080 ML20062N083
References
NUDOCS 8208200298
Download: ML20062N080 (63)

Text

TS-1 REV TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE

__ PAGE 1.0 Definitions TS.1-1 2.0 Safety Limits and Limiting Safety System Settings TS.2.1-1 2.1 Safety Limit, Reactor Core TS.2.1-1 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 Conditions for Operation TS.3.1-1 3.1 Reactor Coolant System Ts.3.1-1 3.2 Chemical amd Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 3.4 Steam and Power Conversion System TS.3.4-1 3.5 Instrumentation System TS.3.5-1 3.6 Containment System TS.3.6-1 3.7 Auxiliary Electrical Systems TS.3.7-1 3.8 Refueling and Fuel Handling TS.3.8-1 3.9 Radioactive Effluents TS.3.9-1 3.10 Control Rod and Power Distribution Limits TS.3.10-1 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 3.14 Fire Detection and Protection Systems TS.3.14-1 3.15 Event Monitoring Instrumentation TS.3.15-1 4.0 Surveillance Requirements TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Inservice Inspection Requirements TS.4.2-1 4.3 Reactor Coolant System Pressure Isolation TS.4.3-1 Valves

~ 4.4 Containment System Tests TS. 4. 4-1 4.5 Engineered Safety Features TS.4.5-1 4.6 Periodic Testing of Emergency Power System TS.4.6-1 4.7 Main Steam Stop Valves

.TS.4.7-1 4.8 Steam and Power Conversion Systems TS.4.8-1 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring-Program

.TS.4.10-1 4.11 Radioactive Source Leakage Test TS.4.11-1 4.12 Steam Generator Tube Surveillance TS.4.12-1 4.13 Snubbers TS.4.13-1 4.14 Centrol Room Air _ Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Spacial Ventilation System TS.4.15-1

-4.16 Fire Detection and Protection Systems-TS.4.16-1.

4.17 Radioactive Effluents TS.4.17-1 8208200298 820813 PDR ADOCK 05000282 P

PDR L__

TS-iii REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data 3.1-2 Unit 2 Reactor Vessel Toughness Data 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument operating conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instramentation 3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.12-1 Safety Related Snubbers 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests

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4.2-1 Special Inservice Inspection Requirements 4.4-1 Unit I and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Radiation Environmental Monitoring Program (REMP)

Sample Collection and Analysis 4.10-2 REMP - Maximum Values for the Lower Limits of Detection 4.10-3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tribe Inspectica 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation :

Surveillance Rquirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1.

Anticipated Annual. Release of Radioactive Material in Liquid Effluents From PrairieLIsland_ Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating i

Plant (Per Unit).

6.1-1 Minimum Shif t Crew Composition 6.7-1 Special. Reports-

TS-iv REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit I and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RT NDT Reactor Vessel Steels Exposed to 550* Temperature 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity, Limit Versus Percent of RATED THERMAL POWER with Primary Coolant Specific Activity >1.0-Ci/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island. Nuclear Generating Plant Site Boundary for Liquid '

Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Caseous Effluents 3.10-1

_ Required Shutdown Reactivity Vs Reactor-Boron Concentration.

3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope For F = ~2.21 9

3.10-6 Deviation from Target Flux Difference as a Function of Thermal

-Power 3.10-7 Normalized Exposure Dependent Function BU(E ) - for Exxon Nuclear Company Fuel:

3.10 V(Z) as a function of core height 4.4-1 Shield Building Design In-Leakage, Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Environmental:

Monitoring Program (Sample Location Map) 4.10-2

-Prairie Island Nuclear Generating Plant' Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-site Operating.

. _ Organization 6.1-2.

' Prairie Island-Nuclear Generating _ Plant l Functional Organization for-On-site Operating Group

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l 1

TS.1-7 I

EV W.

Process Control Program (PCP) 5 The PCP is the manual containing the current formula, sampling, analysis, test and determinations to be made to ensure that the processing and pack-aging 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 10CFR20,10CFR71, and Federal and State regulations and i

other requirements governing the disposal of radioactive wastes.

X.

Solidification Solidification is the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

I Z.

Offsite Dose Calculation Manual (ODCM)

The ODCM is the manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous effluents, in the calculation of liquid and gaseous effluent monitoring instrumentation alarm and/or trip setpoints, and in the conduct of environ-mental radiological environmental monitoring.

A.A.

Source Check 3

A source check is the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

A.B.

Caseous Radvaste Treatment System The 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.

A.C.

Ventilation Exhaust Treatment System f

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 enrironment (such a system is not considered to have any effect on noble gas effluents). Engineered safety feature atmospheric cleanup systems are not considered to be Ventilation Exhaust Treatment l

System Components.

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TS.1-8 REV A.D.

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

A.E.

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.

A.F.

Members of the Public Means all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors, or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational occupational, or other purposes not associated with the plant.

A.G.

Site Boundary Means a line within which the land is owned, leased, or otherwise controlled by the licensee. The site boundary for liquid releases of radioactive material is defined in Figure 3.9-1.

The site boundary for gaseous releases of radfoactive material is defined in Figure 3.9-2.

A.H.

Unrestricted Areas Means any area at or beyond the site boundary to which access is not con-trolled by the licensee for purposes of protection of individuals.from ex-posure to radiation and radioactive materials or any area within the' site-boundary used for residential quarters or industrial, commercial,~institu-tional and/or recreational purposes.

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TS.3.9-1 REV 3.9 RADIOACTIVE EFFLUENTS Applicability Applies at all times to the liquid and gaseous radioactive effluents from the plant and the solidification and packaging for shipment of solid radioactive waste.

Obj ective To implement the requirements of 10CFR20,10CFR71,10CFR50 Section 50.36a, Appendix A and Appendix I to 10CFR50, 40CFR141, and 40CFR190 pertaining to radioactive effluents.

Specifications A. Liquid Effluents 1.

Concentration a.

The concentration of liquid radioactive material released at any time ' from the site (Figure 3.9-1) shall be limited to the concentrations specified in 10 CFR Part 20,~ Appendix B, Table' II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained ngble gases, the concentration shall be limited to 2 x 10 uci/ml total activity.

b.

When the concentration of radioactive material _ in-liquid released from~ the site exceeds -the limits in (a) above, immediately restore the concentration within acceptable. limits and provide prompt notification with written followup to the Commission.

2.

Dose a.

.The dose or dose commitment to an individoal from radioactive materials in liquid effluents released from the site (Figure 3.9.1) for each unit shall be limited:

-1.

During -any calendar quarter. to $1.5 mrem to the total-body and tos 5 mrem to any-organ, and

.2.

During any calen'dar year;tos 3 mrem'tofthe total body and to $10 mrem to~any organ.-

TS.3.9-2 REV b.

When the calculated dose from the release of radioactive materials in liquid released from the site to unrestricted areas exceeds the limits in (a) above, within 30 days submit to the Commission a special report which identifies the cause(s) for exceeding the limit (s) and defines the correc-tive actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.

3.

Liquid Radwaste System a.

The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose due to liquid effluent released from the site (Figure 3.9-1) when averaged over one month would exceed 0.12 mrem to the total body or 0.4 mrem to any organ, b.

With radioactive liquid waste being discharged without treat-ment and in excess of the limits in (a) above, within 30 days submit to the Commission a special report which includes the following information:

1.

Identification of the inoperable equipment or sub-systems and the reason for inoperability.

2.

Action (s) to be taken to restore equipment to operable status, and 3.

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

4.

Liquid Storage Tanks The quantity of radioactive material. contained in each of the a.

following tanks st all be limited to 10 curies, excluding tritium and dissolved or entrained noble gases:

Condensate storage tanks Outside temporary tanks r

I TS.3.9-3 REV b.

With the quantity of radioactive material in any of the above listed tanks exceeding the limit in (a) above, immediately suspend all additions of radioactive materials to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

B. Gaseous Effluents

1. Dose Rate

a. The dose rate at any time due to radioactive materials released in gaseous effluents from the site (Figure 3.9.2) shall be limited to the following values:

1. The dose rate limit for noble gases shall be $500 mrem / year to the total body and s 3000 mrem / year to the skin, and

2. The dose rate limit for I-131, tritium, and radioactive particulates with half-lives greater than eight days shall be s 1500 mrem / year to any organ

b. With the dose rate (s) exceeding the limits in (a) above, immediately decrease the release rate within acceptable limits and provide prompt notification with written followup to the Commission.

2. Dose from Noble Gases i

a. The air dose in unrestricted areas due to noble gases released in gaseous effluents from the site (Figure 3.9-2) from each unit shall be limited to the following values:

1.

During any calendar quarter, to s5 mrad for gamma radiation and $10 mrad for beta radiation, and 2.

During any calendar year, tos10 mrad for gamma radiation and s20 mrad for beta radiation.

b. With the calculated' air dose from radioactive noble gases in gaseous effluent exceeding any of the above limits, within 30 days submit to the Commission a special report which identifies the cause(s) for exceeding the limit (s) and defines the corrective action (s) taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.

1 TS.3.9-4 REV I

3. Dose from I-131 Tritium, and Radioactive Particulates With Half-Lives Creater Than Eight Days j

a. The dose to any organ of an individual due to I-131, tritium, and radioactive particulates with half-lives greater than eight days released in gaseous effluents from the site (Figure 3.9-2) from each unit shn11 be limited to the following:

r 1.

During any calendar quarter to 7.5 mrem, and 2.

During any calendar year to 15 mrem

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b. With the calculated dose from the release of 1-131, tritium, and radioactive particulates with half-lives greater than eight days in gaseous effluents exceeding the limit (s) in (a) above, within 30 days submit to the Commission a special report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.

4.

Caseous Radwaste Treatment System and Ventilation Exhaust Treatment Systems

a. The Caseous Radwaste Treatment System and Ventilation Exhaust Treatment Systems shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected dose-due to gaseous effluents released from the site (Figures 3.9-2) when averaged over one month would exceed 0.4 mrad for gamma j

air dose, 0.8 mrad for beta air dose, or 0.6 mrem organ dose.

b. With gaseous waste being discharged without full treatment and in excess of the limits in (a) above, within 30 days submit to the Commission a special report which includes.the following 3

information:

f

TS.3.9-5 REV 1.

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

2.

Action (s) taken to restore the inoperable equipment to operable status, and 3.

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

c. Except as provided for in (d) below, the concentration of oxygen at the outlet of each operating recombiner shall be limited tos2% by volume.

i

d. With the concentration of oxygen measured at the outlet of L

operating recombiner(s) >2% by volume but < 4% by volume, restore the concentration of oxygen tos 2% by volume within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

e. With the concentration of oxygen at the outlet of operating recombiner(s) >4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen tos2% within one hour.

f. The quantity of radioactivity contained in each gas storage tank shall be limited to g78,800 curies of noble gases (con-sidered as dose equivalent Xe-133).

g. With the goantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

h. The radioactive gas contained in the vaste gas holdup system shall not be deliberately discharged to the environment during unfavorable wind conditions when the cooling towers are in operation.

For the purposes of this specification,-

unfavorable wind conditions are defined as wind from 5* west-of north to 45* east of north at 10 miles per hour or less.

5.

Containment Purging

a. Containment purge and vent releases shall be treated during power operation through the charcoal and particulate filters of the in-service purge system or shield building ventilation system.

b. Prior to purging containment during power operation or immediately.after.shutdownfif the containment is_to be purged, the sampling and analysis specified in Table 4.17-4 shall be completed.

TS.3.9-6 REV C. Solid Radioactive Waste 1.

A solid radwaste system shall be operable and used, as applicable in accordance with a Process Control Program for the Solidification and packaging of radioactive wastes to ensure meeting the require-ments of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.

2.

With the packaging requirements of 10 CFR Part 20 or 10 CFR Part 71 not satisfied, suspend. shipments of defectively packaged solid radioacive wastes from the site.

TS.3.9-7 REV D. Dose from All Uranium Fuel Cycle Sources

a. The dose or dose commitment to a member of the general public from all uranium fuel cycle sources is limited to 25 mrem to the total body or any organ (except for the thyroid, which is limited to 75 mrem) over a period of 12 consecutive months.

b. With the calculated dose from the release of radioactive materials in liquid or gaseous ef fluents exceeding twice the limits of Specifications 3.9.A.2.a.1, 3.9.A.2.a.2, 3.9.B.2.a.1, 3.9.B.2.a.2, 3.9.B.3.a.1, or 3.9.B.3.a.2, submit within 30 days a special report to the Commission which calculates the highest radiation exposure to any member of the general public from all uranium fuel cycle sources (including all effluent pathways and. direct radiation).

Unless this report shows that exposures are less than the 40 CFR Part 190 standard, either apply to the Commission for a variance to continue releases which exceed the 40 CFR Part 190 standard or reduce subsequent releases to permit the standard to be met.

E. Radioactive Liquid Effluent Monitoring Instrumentation

a. The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.9-1 shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.9.A.l.a are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM).

b. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setroint less conservative than required by the above specification, immediately suspend the release of radioactive liquid ef fluents monitored by the af fected channel or declare the channel inoperable.

c. With less than the minimum required radioactive liquid effluent monitoring instrumentation channels operable, take the action shown in Table 3.9-1.

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TS.3.9-8 REV F.

Radioactive Gaseous Effluent Monitoring Instrumentation a.

The radioactive gaseous ef fluent monitoring instrumentation charinels shown in Table 3.9-2 shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.9.B.I.a are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the ODCM.

b.

With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, innediately suspend the release of radioactive gaseous ef fluenta monitored by the affected channel or declare the channel inoperable, c.

With less than the minimum required radioactive gaseous ef fluent monitor-ing instrumentation channels operable, take the action shown in Table 3.9-2.

TS.3.9-9 REV Bases:

A.

Liquid Effluents l

Specification 3.9.A.1 is p ovided to ensure that the concentration of radio-active materials celeased in liquid waste ef fluents from the sita to unre-stricted areas will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures exceeding (1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(2) to the population.

The concentration limit for disaolved or entrained noble gases is based upon the assumption that Xc-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in Inter-national Commission on Radiological Protection (ICRP) Publication 2.

Specification 3.9. A.2.a is provided to implement the requirements of Sections 4

II.A, III.A and IV.A of Appendix I, 10 CFR Part 50.

The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I.

Action required by Specification 3.8.A.2.B provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achicvable", consider-ing that the nearest drinking water supply using the river for drinking water is more than 300 miles downstream, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the drinking water that are in excess of the requirements of 40 CFR 141.

Specification 3.9.A.3 provides assurance that the liquid redwaste treatment system will be available for use whenever liquid effluents require treatment prior to relcase to the environment.

The requirements that the appropriate portinns of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective,given in Section II.D of Appendix 1 to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I,;10 CFR Part 50, for liquid effluents.

The liquid radwaste treatment system is shared by both units. EIt is not practical to determine the contribution from each unit to liquid radwaste releases. For this reason, liquid radwaste releases will be allocated equally to each unit.

Restricting the quantity of radioactive. material' contained in the specified

. tanks provides assurance that in the event ofian uncontrolled release of the contents of _ the tank, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, in an unrestricted area.

TS.3.9-10 REV B.

Gaseous Effluents Specification 3.9.B.I.a is provided to ensure that the dose rate at anytime at the site boundary from gaseous ef fluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B. Table II Column 1.

These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either inside or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary.

The specified release rate limits restrict, at all tiacs, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to $ 500 mrem / year to the total body or to $3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to $1500 mrem / year at the site boundary.

Specification 3.9.B.2.a is provided to implement the requirements of Sections II.B. III.A and IV.A of Appendix I, 10 CFR Part 50.

The Limiting Conditions for Operation implement the guides set forth in Section II.B of Appendix I.

Action required by Specification 3.9.B.2.b provides the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in-gaseous effluents will be kept "as low as is reasonably achievable."

.. = _ _

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1 TS.3.9-11 REV h

Specification 3.9.B.3.a is provided to implement the requirements of Sections l

II.C, III.A and IV.A of Appendix I, 10 CFR Part 50.

The Limiting Conditions t

for Operation are the guides set forth in Section II.C of Appendix I.

The I

action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept i

"as low as is reasonably achievable." The release rate specifications for I-131, tritium and radioactive particulates with half-lives greater than eight l

3 days are dependent on the existing radionuclide pathways to man in the unre-i stricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposi-tion of radionuclides onto green Icafy vegetation with subsequent consumption a

by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground wth subsequent exposure of man.

Specification 3.9.B.4.a provides assurance that the Waste Cas Treatment System and the Ventilation Exhaust Treatment Systems will be available for use i

whenever gaseous wastes are released to the environment. The requirement that

{'

the appropriate portions of the Waste Gas Treatment System be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as ic reasonably achievable."

This specification implements the requirements of 10 CFR 50.36a, General

~

Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

i l

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I i

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-c

TS.3.9-12 REV Specification 3.9.B.4.c, 3.9.B.4.d and 3.9.B.4.e are provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. Maintaining the concentrations below the flammability limit 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.

The waste gas system is a pressurized systen with two potential yources of oxygen:

1) oxygen added for recombiner operation, and 2) placing tanks vented for maintenance back on the system. The system is operated with flow through the recombiners and with excess hydrogen in the system. By verifying that oxygen is $2% at the recombiner outlet, there will be no explosive mixtures in the system.

If the required gas analysers are not operable, the oxygen to the recombiner will be isolated to prevent oxygen from entering the system from this source. Tanks that may undergo maintenance are normally purged with nitrogen before placing them in service to eliminate this as a source of oxygen.

Specification 3.9.B.4.f is provided to limit the radioactivity which can be stored in one decay tank.

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.

Specification 3.9.B.5.a requires the containment to be purged, during reactor operation, through the inservice purge system. This provides for iodine and particulate removal from the purge release. During outages when the containment is opened for maintenance, the containment ventilation exhaust is directed to the monitored reactor building vent.

The cooling towers at Prairie Island are located to the south of the plant and are within the 50*-arc described in this specification. At low wind, velocities (below 10 mph) the gaseous activity released from the gaseous radwaste systen could be at or near ground level near the cooling towers and remain long enough to be drawn into the circulating water in the tower. This specification minimizes the possibility of releases from the gaseous radwaste system from entering the river from tower scrubbing.

The Waste Gas Treatment System, containment purge release vent, and spent fuel pool vent are shared by both units. Experience has also shown that contributioas from both units are released from each auxiliary building vent. For this reason, it is not practical to allocate releases to any specific unit. All releases will be allocated equally in determining conformance to the design objectives of 10 CFR Part 50, Appendix I.

s TS.3.9-13 REV i

C.

Solid Radioactive Waste The Operability requirements placed on the solid radwaste system ensure i

that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite.

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.

The process parameters included in establishing the Process Control Program may include, but are not limited to waste type, war:c pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

D.

Dose from All Uranium Fuel Cycle Sources This specification is provided to meet the dose limitations of 40 CFR 190.

)

The specification requires the preparation and submittal of a special report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I.

Submittal of the report is con-sidered a timely request, and a variance is granted until Staff action on this request is complete.

For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a real individual will exceed 40 CFR 190 if the individual reactors remain within the reporting requirement level.

For the purposes of the special report, it may be assumed that the dose commitment to the real individual from other uranium fuel cycle sources is negligible, with the exception that. dose contributions from other nuclear.

fuel cycle facilities at the same site or within a radius of 5 miles must-be considered.

E. & F. Effluent Monitoring Instrumentation These specifications are provided to assure that effluent releaseL points are continuously monitored.

I 1

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-+.

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SUB. ';

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-- t Fidi RETURN PIPE Z

RADWASTE p/

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Z DISCHARGE s

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- TOWERS,

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• 5 STT4UGTURE Eo

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!ii s

s m

d LEGEND y

- - - -SITE BOUNDARY FOR LIQUID EFFLUENTS i

i i

i i

O 500' IOOO' I590' 2000' FIGURE TS.3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents

-

• q \\'_s__i11' __-11

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'a i'

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I SHO AE LINE 8

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i TURBINE BLOG

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AEACTOR BLOG'S. "

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VENT ELEV. 7*79'MSL CSRADE I

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- - - Site Boundary for g

Gaseous Effluents i '

8

- w

• Fence W

'\\s

- - Property Line h

' si

' [gggg 7g 7__ _ ;

7 1[

L*D*3 6

sm-em.

,sm -

FIGURE TS.3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents

. _ =.

t I

TABLE TS.3.9-1 (Pg 1 of 2',

i REV I

TABLE TS.3.9 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION 1.

Gross Radioactivity Monitors Providing Automatic Termination of Release a.

Liquid Radwaste Effluent Line 1

During releases 1

b.

Steam Generator Blowdown 1/ Unit During releases 2

Effluent Line 2.

Flow Rate Measurement Devices P.

Liepid Rndwaste Ef fluent Line 1

During releases 4

requiring throt-tling of flow b.

Steam Generator Blowdown Flow 1/ Gen During releases 4

c.

Discharge Canal Flow 1

At all times 4

3.

Continuous Composite Samplers a.

Each Turbine Building 1/ Unit During releases 3

Sump Effluent Line

-4.

Discharge Canal Monitor 1

At all times:

3 5.

Tank Level Monitor a.

Condensate Storage Tanks 1/ Unit When tanks 5

are in use b.

Temporary Outdoor Tanks 1/ Tank Uhen tanks 5

Holding Radioactive Liquid are in use.

TABLE TS.3.9-1 (Pg 2 of 2)

REV TABLE TS.3.9-1 TABLE NOTATION ACTION 1 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases may continue for up to 14 days provided that prior to each release:

a.

At least two independent samples are analyzed in accordance with Specification 4.17. A.1.d, and b.

At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are saalyzed for gross radjoactivity(betaorgamma)atalimit of detection of at least 10 uCi/ gram:

1.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the specific activity of the secondary coolant is > 0.01 uCi/ gram dose equivalent I-131.

2.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is < 0.01 uCi/ gram dose equivalent I-131.

ACTION 3 With the numbers of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, grab samples are collected and saved for weekly compositing nnd analysis.

ACTION 4 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathwsy may continue 'sr up;to 30 days provided the flow rate is estimated at least once.per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.

Pump curves may be used to estimate flow.

ACTION 5 With the number of channels Operable less than required by the Minimum Channels Operable requirement, liquid additions to the tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions.

L

TABLE TS.3.9-2 (Pg 1 of 2)

REV TABLE TS.3.9-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMEWT OPERABLE APPLICABILITY ACTIO'd 1.

Waste Gas Holdup System 2

During system 2

Explosive Gas operation (Oxygen) Monitors 2.

Effluent Release Points (Unit No. 1 Reactor Bldg, Unit No. 1 Aux Bldg, Unit No. 2 Reactor Bldg, Unit No. 2 Aux Bldg, Spent Fuel Pool, Radwaste Bldg) a.

Noble Gas Activity 1

During releases 4,5,7 Monitor

• b.

Iodine Sampler 1

During releases 3

Cartridge c.

Particulate Sampler 1

During releases 3

Filter d.

Sampler Flow 1

During releases 1

Monitor 3.

Air Ejector Noble Gas 1

During power 6

Monitors (Each Unit)-

operation

• Noble gas activity monitors providing automatic termination of releases (except the Radwaste Building which has no automatic isolation function).

TABLE TS.3.9-2 (Pg 2 of 2)

REV TABLE TS.3.9-2 TABLE NOTATION ACTION 1 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2 With the number of channels Operable less than required by the Minimum Channels Operable requirement, operating of this system may continue for up to 14 days. With two channels inoperable, manually isolate the oxygen addition line.

ACTION 3 With the numbers of channels Operable less than required by the Minimum Channels Operable requirenent, effluent releases via this pathway may continue for up to 30 days, provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.17-4.

ACTION 4 With the number of channels Operable less than required by the Minimum Channela Operable requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and these sanples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 5 With the number of channels Operable less than required by the }Enimum Channels Operable requirement, immediately suspend Purging of radio-active effluents via this pathway (applicable to Reactor Building vents).

ACTION 6 With the number of channels Operable less than required by the Minimum Channels Operable requirement, air ejector operation may continue for up to 30 days provided grab samples are taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l ACTION 7 With the number of channels operable less than required by the Minimum l

Channels operable requirement, the contents of the waste gas decay tanks j

may be released to the environment for up to 14 days provided that prior i

to initiating 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 Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway (applicable to Unit 2 Auxiliary Building Vent).

L

TABLE TS.4.1-1 (Page 4 of 5) 4 Channel.

Functional

Response

Description.

Check Calibrate Test Test Remarks 19.

Deleted 20.

Boric Acid Make-up Flow NA R

NA NA Channel-21.

Containment Sump Level NA R

R NA Includes Sumps A, B, and C

- 22. LAccumulator Level.

S R

R NA and Pressure 23.

Steam Generator Pressure S

R M

NA 24.

Turbine First Stage Pressure S

R M

NA 25.

Emergency Plan Radiation

• M R

M NA Includes those named in the emergency Instruments procedure (referenced in Spec.

6.5 A.6.)

26.. Protection Systems

. NA '

NA M

NA Includes auto load sequencers Logic. Channel Testing 27.

Turbine Overspeed NA R

M NA

^

-Protection Trip Channel 28.

Deleted 29.

Deleted 30.

Deleted 31.

Seismic Monitors R

R NA NA Includes those reported in Item 4 of Table TS.6.7-1 32.

Coolant' Flow - RTD S

'R.

M NA Bypass Flowmeter

- 33.

CRDM Cooling Shroud

'S NA R

NA FSAR page 3.2-56 Exhaust Air Temperature

34.. Reactor Gap Exhaust S

NA R

NA FSAR page 5.4-2 Air Temperature

~

_%-4------.#.._e

-.r e

r-

-= *-

TS.4.10-1 REV 4.10 RADIATION ENVIRONMENTAL MONITORING PROGRAM Applicability Applies at all times to the periodic monitoring and recording of radioactive effluents found in the plant environs.

Objective To provide for measurement of radiation levels and radioactivity in the site environs on a continuing basis.

Specification A.

Sample collection and Analysis 1.

The Radiation Environmental Monitoring Program described in Table 4.10-1 shall be conducted. Radioanalysis shall be conducted meeting the requirements of Table TS.4.10-2.

A map and a table identifying the locations of the sampling shall be provided in the Offsite Dose Calculation Manual (0DCM).

2.

Whenever the Radiation Environmental Monitoring Program is not.being conducted as specified in. Table TS.4.10-1 the Annual Radiation Environmental Monitoring Report shall-include a description of the reasons for not conducting the program as required and plans for preventing a recurrence.

(

3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due ' to hazardous conditions, seasonable unavailability, or to malfunction of automatic sampling equipment. If the latter occurs, every ef fort shall be made to complete corrective action prior to the end of the next sampling period.

4.

With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4.10-3 when averaged over any calendar quarter, in lieu of any other' report, prepare and submit

(

to the Commission within 30 days'from'the end of the.affected calendar

('

quarter a Report pursuant to Specification 6.7.C.2(a).

When more than

~

i one of the radionuclides in Table 3.12-2'are detected lLn the sampling l

medium, this report shall be submitted if:

concentration (1), concentration (2)

...> 1.0 l

limit level (1)

-limit level (2)

When radionuclides other than-those--in Table 4.10-2 are detected and are the. result of plant effluents, this report shal1~be submitted:if

~

the potential annual _ dose to an individual;is equal to or greater than

-the calendar year limits'of~ Specifications 3.9.A.2,13.9.B.2, or 3.9.B.3..

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 Monitoring Report.

- - ~. -

TS.4.10-2 REV 1

i 5.

Although deviations from the required sampling schedule are permitted under Paragraph 3 above, whenever milk or leafy green vegetables can no longer be obtained from the designated sample locations required by Table 4.10-1, the Semiannual Radioactive Effluent Release Report for this period shall explain why the samples can no longer be obtained and will identify the new locations which will be added to and deleted from the monitoring program as soon as practicable.

B.

Land Use Census 1.

A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. This census shall be conducted at least once per 12 months between the dates of May 1 and October 31 by door to door survey, aerial survey, or by consulting local agricultural authorities or associations.

2.

With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 4.10-A.1, the Semiannual Radioactive Effluent Release Report for this-period shall identify the new-location. The new location shall be added to the radiological environmental monitoring program within 30 days.

The sampling location, excluding the control station location, having the lowest calculated dose or dose commit-ments (via the same exposure pathway) may be deleted from-this monitoring program af ter October 31 of the year in which this land use census was conducted.

C.

Interlaboratory-Comparison Program 1.

Analyses shall be performed on radioactive materials supplied as part of an NRC approved interlaboratory comparison program as described in the ODCM.

2. -The results of analyses performed as a part of the above required program shall be included in the Annual Radiation Environmental Monitoring Report.

When required' analyses are not performed, corrective action shall'be reported in the Annual Radiation Environmental Monitoring Report.

TS.4.10-3 REV Basis A.

Sample Collection & Analysis The Radiation Environmental Monitoring Program required by this specification provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the plant operation.

This program thereby supplements the radiological effluent monitoring by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The detection capabilities required by Table 4.10-2 are state-of-the-art for routine environmental measurements in industrial laboratories, and the LLD's for drinking water meet the requirement of 40CFR141 B.

Land Use Census This specification is provided to ensure that changes in the use of off site areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information f rom door-to-door, aerial or consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter.

C.

Interlaboratory Comparison Progrm The requirement for participation in an interlaboratory comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as a part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

TABLE TS.4.10-1 (Page 1 of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLAfTr RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis 1.

AIRBORNE Samples from 5 locations:

Continuous Sampler Radioiodine analysis Radioiodine and 3 samples from offsite operation with weekly for I-131 Particulates locations (in different sample collection sectors) of the highest weekly Particulate:

calculated annual average -

Gross beta activity on ground level'D/Q, each filter weekly *.

1 sample from the vicinity Analyses shall be per-of a community having the formed more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> highest calculated annual following filter change.

average ground-level D/Q, and Perform gamma isotopic 1 sample from a control loca-analysis on composite tion 8-20 miles distance and (by location) sample in the least prevalent wind.

quarterly.

direction 2.

DIRECT 32 TLD stations established Quarterly Gamma dose

. RADIATION with duplicate dosimeters quarterly placed at the following locations:

1.

Using the 16 meteoro-logical wind sectors

e g as guidelines, an. inner E to ring of stations in the N

general area of the. site y

boundary is established

.and an outer ring of stations in the 4 to 5 mile o

distance.from the plant M

site is established.

Because g

of inaccessibility, seven g

sectors in the inner and a

outer rings.are not covered o

If Gross beta activity in any indicator sample exceeds 10 times the yearly average of the control sample, a gamma

[

u isotopic ~ analysis is required..

we CO Sample locations - are given on the figure and table in the ODCM.

TABLE TS.4.10-1 (Page 2 of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLANI RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS Number of Samples Exposure Pathway and.

Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis

'2.

DIRECT RADIATION (con't) 2.

Seven dosimeters are established at special interest areas and a control station.

3.

WATERBORNE

a. Surface Upstream & downstream Monthly Composite of Gamma isotopic analysis locations weekly samples (water of each monthly composite

& ice conditions permitting)

Tritium analysis of quarterly composites of monthly corposites

b. Ground 3 samples from wells Quarterly Camma isotopic and tritium within 5 miles of the analyses of each sample plant site and I cample from a well greater than 10 miles'from the plant site gg

<m

c. Drinking

1. sample from the City Monthly Composite of I-131 Analysis and Gross Q

of Red. Wing water supply weekly samples beta and gamma isotopic g

analyses of each monthly P

composite f'

5 Tritium analysis of quarterly L

composites of monthly composites s

a 2

C* Sample locations are given on the figure and table in the ODCM.

m S

S

~

TABLE TS.4.10-1 (Page 3 of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT

. RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE' COLLECTION AND ANALYSIS

Number of Samples.

Exposure Pathway and Sampling and Type and Frequency

.and/or Sample

' Sample Locations **

Collection Frequency of Analysis 3.

WATERBORNE (continued) i

d. Sediment'and' L.One sample. upstream o'f Semiannually Gamma isotopic analysis shoreline

-plant, one sample.down-of each sample

. stream of plant, and one

'from~ shoreline of-

-recreational' area.

l 4.

INGESTION

a. Milk:

one. sample from dairy Monthly or biweekly Gamma isotopic and I-131 farm having higest D/Q, if animals are on analysis of each sample one sample from each-of pasture

.three dairy farms cal-culated to have doses from I-131 > l mrem /yr, Land one sample from 10-20

. miles

b.. Fish and One sample of one, game Semiannually Gamma irotopic analysis f

Invertebrates -specie of fish located.

on each sample (edible upstream and downstream portion only on-fish) h of"the plant ~ site' f

w o

One sample of. Invertebrates' 4

' upstream-and' downstream of.

m the' plant site g

I

• • Sample-locations are given on the figure and table in the ODCM

.y

?ss S.

TABLE TS.4.10-1 (Page 4 of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLAhi RADIATION ENVIRON'iENTAL MONITORING PROGRAM

. SAMPLE COLLECTION AND ANALYSIS Number of Samples

. Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis c.

Food Products' One sample of corn At time of harvest Gamma isotopic analysis from highest D/Q of edible portion of farm and one sample each sample from 10-20 miles.

-One sample of broad At time of harvest I-131 analysis of edible leaf vegetation from portion of each sample highest D/Q~ garden and one sample from 10-20 miles e

o EY

< to E

d

• • Sample locations are' given on the figure and table in the ODCM.

Q k

S

TABLE TS.4.10-2.

MAXIMUM VALUES FOR TIIE LOWER LIMITS OF DETECTION (LLD) a,e Airborne. Particulate Water-or Gag Fish Milk Food Products Sediment Analysis (pCi/1)

(pCi/m )

(pCi/kg, wet)

(pCi/1)

(pCi/kg, wet)

(pCi/kg, dry) gross beta 4

1 x 10-1I 2000(1000b) 3 5.4,15-130 3

59,-

30 260 7

58,60Co 15 130 0'Zn 30.

260 E@

95 C~

Zr-Nb'

.IS

<g 131-b.d

-2 d

7 g

7 x 10 g

60 o

134,137 b

-2 Cs 15(10 ), 18 1 x 10 130 15 80 150 Q

I

'140 e

Ba-La 15 15 c 3

TABLE TS.4.10-2 TABLE NOTATION a - The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

E. V;. 2.22. Y. exp(-> at)

Where:

LLD is the apriori lower limit of detection as defined above (as picoeurie per unit mass or volume),

s is the standard deviation of the background counting rate or of the counting rate of a blank sample b

as appropriate (as counts per minute).

In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributing of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples). Typical values of E, V, Y and at shall be used in the calculations.

E is the counting efficiency (as counts per transformation),

2.22 is the~ number of transformation per minute per picoeurie,

{

Si Y is:the fractional radiochemical yield (when applicable),

d A is the radioactive decay constant for the particular radionuclide, and

).

~

at is the elapsed time between sample collection (or end of the sample collection period) and time of

?

counting.

]#

m b - LLD for drinking water.

E*

ci-Total for parent and daughter

-d - Applies to special isotope analysis-not to gamma' spectrum analyses o - Other. peaks which are measurable.and identifiable, together with the radionuclides in Table 4.10-2 shall Oi be' identified and reported.

to G

T'

~

T r-v v

"+V-

TABLE TS.4.10-3 REPORTING LEVELS FOR RADIOACTIVITY CONCEPTERATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Airborne Particulate Water or Gag Fish

. Milk Vegetables Analysis (pCi/1)

(pCi/m )

(pCi/kg, wet)

(pCi/1)

(pC1/kg, wet)

H-3 2 x 10 (*)

3 4

Mn-54 1 x 10 3 x 10 2

Fe-59 4 x 10 1 x 10' 3

Co-58 1 x 10 3 x 10 2

4 Co-60 3 x 10 1 x 10 2

4 Zn-65

'3 x 10 2 x 10 Zr-Nb-95

'4 x 10 (b) 2 2

1-131 2

0.9 3

1 x 10 Cs-134 30 10 1 x 10 60 1 x 10 E en 3

3 M

Cs-137 50 20-2 x 10 70 2 x 10 d

Ba-La-140.

2 x 10 (b)'

3 x 10 (b) 2 2

L w

a - For drinking water samples b - Total.for parent and' daughter 4

T-1 t-"

w*

TS.4.17-1 REV 4.17 RADIOACTIVE EFFLUENTS SURVEILLANCE Applicability:

Applies at all times to the periodic monitoring and recording of liquid and gaseous radioactive effluents, verification of solidification, and verification of equipment operability.

Objective:

To implement the requirements of 10CFR20,10CFR71,10CFR50 Secton 50.36a, Appendix A and Appendix I to 10CFR50, and 40CFR190 pertaining to radio-active effluents.

3 i

Specification:

A. Liquid Effluents

1. Dose Rate Monitoring and Calculations

a. 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 Table 4.17-1.

i

b. Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.17-3.

c. The results of radioactive analysis shall be used in accordance with the methods of the ODCM to assure that-the concentrations at the point of releases are maintained within the limits of Speci-fication 3.9. A.l.a.

2.

Dose Calculations

~i a.

Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined _in~accordance with the.

ODCM monthly.

a T

.~.

TS.4.17-2 REV 3.

Liquid Radwaste System a.

Doses due to liquid releases shall be projected at least once each month in accordance with the ODCM.

4.

Liquid Storage Tanks a.

The quantity of radioactive material contained in each of the tanks listed in Specification 3.9.A.4.a shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

B. Gaseous Effluents

1. Dose Rate Monitoring and Calculations

a. Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated Operable by performance of the Channel Check, Source Check, Channel Calibration, and Channel Functional Test operations at the frequencies shown in Table 4.17-2.

b. The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Specification 3.9.B.1.a in accordance with the methods and procedures of the ODCM.

c. The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the limits of Specification 3.9.B.l.a in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and-analysis program, specified in Table 4.17-4.

2. Dose Calculations for Noble Gases

a. Cumulative dose contributions for the current calendar quarter and current year shall be determined in accordance with the ODCM monthly.

TS.4.17-3 REV

3. Dose Calculations for Radioiodines, Radioactive Particulates, and Radionuclides Other Than Noble Gases With Half-Lives Greater Than Eight Days.

a. Cumulative dose contributions for the current calendar quarter and current calendar year determined in accordance with the ODCM shall be monthly.

4. Waste Gas Treatment System

a. Doses due to gaseous releases to unrestricted areas shall be projected monthly.

Releases considered in the projection should include all potentially radioactive plant gaseous effluents from all gaseous radioactive waste management systems and ventilation exhaust systems that are planned to be operated at the projected capacity and efficiency of each. A projected dose in excess of the limits specified in 3.9.B.4.a indicates that additional components or subsystems of the waste gas treatment system must be placed in service to reduce radioactive materials in gaseous effluents.

b. The concentration of oxygen in the waste gas holdup system shall be determined to be within the limits of Specification 3.9.B.4.c by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required Operable by Table 3.9-2.

c. The quantity of radioactive material in each gas storage tank in use.

shall be determined to be within the limit specified in 3.9.B.4.f monthly.

If the inventory of any tank exceeds 10,000 Curies, daily sampling when making additions shall be performed.

5. Atmospheric Steam Dump Monitoring-a.

The I-131 activity in the secondary side of each steam generator shall be. determined as required by Table TS.4.1-2B.

b.

Each time the atmospheric steam dump is used with detectable I-131 activity in the secondary coolant, the total amount of I-131 released shall be calculated based on the most recent activity measurements of the secondary water..

I i

c.

If the total amount of.I-131 released in one steam dump is greater than twice the limit of 3.9.B.3.a.2, the milk ' f rom dairy cows grazing in the downwind area shall be-analyzed for a' period of 5 days following the release. The downwind area shall_ include the 22-1/2-degree sector of a circle having its center at the plant and a 2-mile radius. The I-131 in the milk'shall be determined each day following the dump, using instrumentation with a minimum I-131 detection limit of 1.0 pCi/1.

~

TS.4.17-4 REV C. Solid Radioactive Waste

1. Verification of Solidification a.

The Process Control Program (PCP) shall be used to verify the Solidification of at least one representative test specimen-from at least every tenth batch of each type of wet radioactive waste (e.g. filter sludges, spent resins, evaporator bottoms, and chemical solutions) b.

If any test specimen fails to verify Solidification, the Solidification of the batch under test shall be suspended until such time as additicnal test specimens can be obtained, alternative Solidification parameters can be determined in accordance with the PCP, and a subsequent test verifies Solidification.

Solidification of the batch may then be resumed using the alternative Solidification parameters determined by the PCP..

c.

If the initial test specimen-from a batch of waste fails to verify Solidification, the PCP shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate Solidification. The PCP shall be modified as required, as provided for in Section 6 of.the Technical Specifications.

D. Dose from All Uranium Fuel Cycle Sources a.

Cumulative dose contributions from' all plant liquid and gaseous effluents shall'be determined in accordance with Specifications-4.17.A.2.a, 4.17.B.2.a 4.17.B.3.a, and the methods in the ODCM2 Basis i

Radioactive liquid ef fluent instrumentation is provided = to monitor and.-

I control, as applicable, the releases"of. radioactive materials in' liquid effluents during actual'or potential releases of liquid effluents. =The alarm setpoints for these instruments are calculated in accordance with -

the. methods in the ODCM to ensure that_the alarm will occur prior to exceeding the limits of 10 CFR Part 20.

The operability requirements and use of this instrumentation are consistent with ' the requirements of 4

General Design Criteria 60, 63 and 64 of. Appendix A to 10 CFR Part'50.

i j

l

TS.4.17-5 REV Rsdioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The operability requirements for this instrumentation are consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

The dose calculations for liquid effluents in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", Revision 1, April 1977. NUREG-0133, October, 1978, provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

The dose calculations for gaseous effluents in the ODCM also implement the requirements of Section III.A that conformance with the guides be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the_ actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation-of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating

(

Compliance with 10 CFR Part 50, Appendix I," Revision 1, October '1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determin-ing the air doses at the site boundary, will be based upon the historical average atmospheric conditions. NUREG-0133, October, 1978 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

TABLE TS.4.17-1 (Page 1 of 2)

REV TABLE TS.4.17-1

- RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

~

Source Channel Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequencg (4) l Liquid Radwaste Effluent Daily during Prior to Quarterly (1)

At least once every Line Gross Radioactivity releases Each Release 18 months (3)

Monitor Liquid Radwaste Effluent Daily during Quarterly At least once every Line Flow Instrument releases 18 months Steam Generator Blowdown Daily during Monthly Quarterly (1)

At least once every Gross Radioactivity releases 18 months (3)

Monitors Steam Generator Blowdown Daily during Quarterly At least once every Flow Releases 18 months Turbine Building Sump Daily during Quarterly At least once every Continuous Composite releases 18 months Samplers Discharge Canal Daily during Monthly Quarterly (2)

At least once every Monitor releases 18 months Discharge Canal P111y during Quarterly At least once every Flow Instruments releases 18 months Condensate Storage Daily Quarterly At least once every Tank Level Monitors 18 months Level Monitors for Daily when Quarterly At least once every Temporary Outdoor in use when in use 18 months when in Tanks Holding use Radioactive Liquid

TABLE TS.4.17-1 (Page 2 of 2)

REV TABLE TS.4.17-1 TABLE NOTATION 1.

The Channel Functional Test shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure (if provided).

3.

Instrument indicates a downscale failure (if provided).

4.

Instrument controls not set in operate' mode (if provided).

2.

The Channel Functional Test shall also demonstrate that alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure (if provided).

3.

Instrument indicates a downscale failure (if provided).

4.

Instrument controls not set in operate mode (if provided).

3.

The initial Channel Calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using sources traceable to NBS standards. These standards shall permit.

calibrating the system over its intended range of energy.and measurement range. For subsequent Channel Calibrations, sources that have been related to the initial-calibration shall be used.

~

4.

7he Channel Check shall consist of verifying indication of flow during periods of' release. A Channel Check shall be made at least once daily on any day on which continuous, periodic, or batch releases are-made.

TABLE TS.4.17-2 (Page 1 of 2)

REV TABLE TS.4.17 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIRE'fENTS Source Channel Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency Waste Gas Holdup System Daily During Monthly (2)

Quarterly (5)

Explosive Gas System (Oxygen) Monitors Operation Effluent Release Points (Unit No. 1 Reactor Bldg, Unit No. 1 Aux Bldg, Unit No. 2 Reactor Bldg, Unit No. 2 Aux Bldg, Spent Fuel Pool, Radwaste Bldg)

Noble Gas Activity Daily During Monthly

• Quarterly (1)

At least once every Monitor (4)

Releases 18 months (3)

(Except Radwaste Building)

Noble Gas Activity Daily During Monthly Quarterly (2)

At least once every Monitor Radwaste Releases 18 months (3)

Building (4)

Iodine and Weekly Particulate Samplers Sampler Flow Rate Weekly Quarterly At least once every Monitor 18 months Air Ejector Noble Gas Daily During Monthly Quarterly (2)

At least once every Monitors (Each Unit)

Releases 18 months (3)

O A source check of the applicable noble gas monitor shall be conducted prior to each waste gas decay tank or containment purge release.

TABLE TS.4.17-2 (Page 2 of 2)

REV TABLE TS.4.17-2 TABLE NOTATION 1.

The Channel Functional Test shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following exists.

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure (if provided).

3.

Instrument indicates a downscale failure (if provided).

4.

Instrument controls not set in operate mode (if provided).

2.

The Channel Functional Test shall also demonstrate that alarm annunication occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure (if provided).

3.

Instrument indicates a downscale failure (if provided).

4.

Instrument controls not set in operate mode (if provided).

3.

The initial Channel Calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards or' using: sources traceable to NBS standards.. These standards 'shall permit calibrating the. system over its intended range of energy and measurement range. For subsequent Channel Calibrations, sources that have been related to the initial calibration shall be used.

4.

Noble gas monitor in the Radwaste Building vent not provided with automatic isolation trip.

5.

The Channel Calibration shall include-the use of a nitrogen zero gas _and'an oxygen span gas with a nominal concentration suitable for the' range of the.

instrument.

s C

t 4

1

TABLE TS.4.17-3 (Page 1 of 4)

REV TABLE TS.4.17 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Type of Activity Lower Limit of Liquid Release Type Frequency Analysis Analysis DetectionfLLD)

Frequency (uCi/ml)**

E Batch Tank Release Each Batch Each Batch Principal Gamma 5 x 10~

Emitters

-6 I-131 1 x 10 One Batch One Batch Dissolved and 1 x 10~

Each Month Each Month Entrained Gases

~

Each Batch Monthly H-3 1 x 10 b

Composite Gross alpha 1 x 10~

Each Batch Quarterly

1. ~

1. ~

b Composite Fe-55 1 x 10 Continuous Releases Continuous Weekly PrincipagGamma 5 x 10~

g Turbine Building Composite Emitters Sumps

-0 I-131 1 x 10 Grab Sample Each Sample Principle Gamma -

5 x 10~

Each Day

. Emitters with- 0.5

-6 gpm steam I-131_

1 x 10 generatgr leakage

-5 Grab Sample-Each Sample

-Dissolved and'-

l'x'10 Each Month Entrained Gases

-5 Continuous" Monthly.

H-3' 1 x 10 g.

Composite Gross Alpha 1 x:10~

~3 Continuous"

-Quarterly Sr-89, Sr-90 5 x.10 g

Composite

. Fe -55 l'x-10-6

TABLE TS.4.17-3 (Page 2 of 4)

REV TABLE TS.4.17 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Liquid Release Type Sampling Minimum Type of Activity Lower Limit of Frequency Analysis Analysis DetectionfLLD)

Frequency (uCi/ml ^'

-7 Continuous Releases

• Weekly Grab Each Principal Gamma 5 x 10 Steam Generator Sample Dyring Sample Emitters Blowdown Releases I-131 1 x 10

-5 Grab Sample Each Sample Dissolved and 1 x 10 Each Month Entrained Cases During Releases

-5 Weekly Grab Monthly 11 - 3 1 x 10 b

Sample Dyring Composite Releases

_7 Gross Alpha 1 x 10

-8' Weekly Grab Quarterly Sr-89, Sr-90 5 x 10 b

Sample Dyring Composite Releases

-6 Fe-55 1 x 10 i

t l

l t

l r*

i i

i TABLE TS.4.17-3 (Page 3 of 4)

REV TABLE TS.4.17-3 TABLE NOTATION a.

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical sepa ra tion) :

4.66 s b

E. V. 2.22. Y. exp (-X At)

Miere :

LLD is the apriori lower limit of detection as defined above (as picocurie per unit mass or volume),

i s is the standard deviation of the background counting rate or of the b

counting rate of a blank sample as appropriate (as counts per minute, E is the counting ef ficiency (as counts 'per transformation),

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

t 2.22 is the number of transformations per minute per picoeurie, i

' Y is the fractional radiochemical yield (when applicable),

h is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between addpoint of sample collection and time of counting.

ti b'

t e *$

y-

=*

s-tgL-

I TABLE TS.4.17-3 (Pg 4 of 4)

REV TABLE TS.4.17-3 TABLE NOTATION (Continued)

Notes; b.

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 which is representative of the liquids released.

c.

The principal gamma emitters for which the LLD specification will apply are exclusively 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 which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

d.

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level.

When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Ef fluent Release Report..

e.

A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g., from a volume of system that has an input flow during the continuous release, f.

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 release.

g.

A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, to assure representative sampling.

s h.

Daily grab samples from. the turbine building sumps shall be collected

'+

and qualyzed for principal gamma emitters, including I-131, whenever primary to secondary leakage exceeds 0.5 gpm in any steam generator.

- This sampling is provided in lieu of continuous monitoring with auto-matic isolation.

1.

Grab samples shall be collected at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when steam generator blowdown releases are being made and the specific activity of the secondary coolant is > 0.01 uC1/ gram dose equivalent I-131.

i TABLE TS.4.17-4 (Page 1 of 3) l REV

?

l TABLE TS.4.17 RADI0 ACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

{

P Sampling Minimum Type of Lower Limit of

{

Gaseous Reicase Frequency Analysis Activity Detection {LLD) i Type Frequency Analysis (uCi/ml)*'

I Prior to 2

~4 Each Tank Ea r *, 9 mple Principal Gamma 1 x 10 Waste Gas Release Emitters" l

Storage Tank Grab Sample

-6 H-3 1 x 10 I

g Containment Purge Prior to 4

j Each Purge Each Sample Principal Camma 1 x 10 l

Grab Emitters" l

Sample

-6 H-3 1 x 10 t-i b

I' Effluent Release Monthly Monthly Principal Gamma Points (Unit No. 1 Grab Emitters" 1 x 10

~

Reactor Bldg. Unit Sample No. 1 Aux Bldg, t

i Unit No. 2 Reactor 8

-12 Bldg, Unit No. 2 Continuous Weekly

• I-131' 1 x 10 Aux Bldg, Spent Charcoal

-10 Fuel Pool, Sample I-133

- 1 x 10 Radwaste Bldg)

~11 8

Weekly"'

Principal Gamma 1 x 10 Continuous i

Particulate Emitters

• Sample c

-6 Monthly h H-3 1 x'10 Continuous8

)

Silica Je1 Sample 8

~

Continuous Monthly Gross-1 x 10 Composite Alpha Particulate Sample 8

~1I Continuous Quarterly

.Sr-89, Sr-90 1 x 10 Composite Particulate Sample 8

Continuous Noble Gas Noble Gases 1 x 10 Monitor ~

Gross' beta and l

gamma l

d e

-,a n

e r--m,

.-v-

,,,p s

y

TABLE TS.4.17-4 (Page 2 of 3)

REV TABLE TS.4.17-4 TABLE NOTATION a.

The LLD is the smallest concentration of radioactive, material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

E. V. 2.22. Y. exp (-) A t)

Where:

LLD is the apriori lower limit of detection as defined above (as picoeurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per transformation),

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

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

h is the radioactive decay constant for the particular radionuclide, and a t is the elapsed time between midpoint of sample collection and time of counting.

1 1

TABLE TS.4.17-4 (Pg 3 of 3)

REV TABLE TS.4.17-4 TABLE NOTATION (Continued) b.

Grab samples taken at the ventilation exhausts are generally below minimum detectable levels for most nuclides with existing analytical equipment.

If this is the case, PWR CALE Code noble gas isotopic ratios may be assumed.

c.

With 1 uCi/gm Dose Equivalent I-131 in either Unit 1 or Unit 2 reactor coolant system, the iodine and particulate collection devices for all release points shall be removed and analyzed daily until it is shown that a pattern exists which can be used to predict the release rate. Sampling may then revert to weekly. When samples collected for one day are analyzed, the corresponding LLD's may be increased by a factor of 10.

Samples shall be analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter removal.

d.

To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent streams.

e.

The principal gamma emitters for which the LLD specification will apply are exclusively the following~ radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. ' This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

f.

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.

g.

The ratio of the sample flow rate to _ the sampled stream flow rate shall be known for the time period sampled. Design flow rates may be used for building exhaust vent flow rates,

h.. Releases are made via the reactor building vents only during purging, or operation of.the shield building ventilation system, lor operation of_the auxiliary building special ventilation system. In lieu of weekly removal and analysis of iodine and particulate collectionL devices, these devices may be removed and analyzed following-each release. Removal and analysis of collection devices is not required if releases are not being made.-

TS.6.2-3 REV f.

Investigation of all events which are required by regulation or technical specifications (Appendix A) to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

g.

Revisions to the Facility Emergency Plan, Facility Security Plan, and the Fire Protection Program, h.

Operations Committee minutes to determine if matters considered by that Committee involve unreviewed ur unresolved safety questions.

1.

Other nuclear safety matters referred to the SAC by the Operations Committee, plant management or company management.

j. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures systems, or components.

k.

Reports of special inspections and audits conducted in accordance with specification 6.3.

1.

Changes to the Of fsite Dose Calculation Manual (0DCM).

t m.

Review of investigative reports of unplanned releases of radioactive material to the environs.

j 6.

Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to assure safe facility operation.

a.

Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years. The audits shall be performed in accordance with appropriate written instructions and procedures.

b.

Audits of aspects of plant radioactive effluent treatment and radio-logical environmental monitoring shall be performed as follows:

1. Implementation of the Offsite Dose Calculation Manual at least once every two years.

4

2. Implementation of the Process Control Program for solidification-of radioactive wastes at least once every two -years.

~

3. The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at least once every year.

c.

Periodic review of the audit program should be performed by the SAC at least twice a year to assure its adequacy.

d. - Written reports of the audits shall be _ reviewed by the Vice President Power Production,- by the SAC at a scheduled meeting, and by members of management having responsibility in the areas audited.'

TS.6.2-4 REV 7.

Authority The SAC shall be advisory to the Vice President - Power Production 8.

Records Minutes shall be prepared and retained for all scheduled meetings of the Safety Audit Committtee.

The minutes shall be distributed within one month of the meeting to the Vice President - Power Production, the General Manager Nuclear Plants, each member of the SAC and others designated by the Chairman. There shall be a formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group, b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions for use of specir.iists or subgroups, e.

Authority to obtain access to the nuclear power plant operating record files and operating personnel when assigned audit functions.

f.

Requirements for distribution of reports and minutes prepared by the group to others in the NSP organization.

TS.6.2-6 REV f.

All events which are required by regulations or Technical Specifications to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Drills on emergeacy procedures (including plant evacuation) and adequacy g.

of communication with of f site support groups.

h.

All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan, shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.

1.

Special reviews and investigations, as requested by the Safety Audit Committee.

j. Review of investigative reports of unplanned releases of radioactive material to the environs.

k.

All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM).

5.

Authority The OC shall be advisory to the Plant Manager.

In the event of a disagree-ment between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed.

A written summary of the disagreement will be sent to the General Manager Nuclear Plants and the Chairman of the SAC for review.

6.

Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed.

The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Manager Nuclear Plants and others designated by the OC Chairman or Vice Chairman.

7.

Procedures A written charter for the OC shall be prepared that contains:

Responsibility and authority of the group a.

b.

Content and method of submission of presentations to the Operations Committee c.

Mechanism for scheduling meetings d.

Provision for meeting agenda

TS.6.5-1 REV 6.5 PLANT OPERATING PROCEDURES Detailed written procedures, including the applicable checkof f lists and instructions, covering areas listed below shall be prepared and followed. These procedures and changes thereto, except as specified in TS 6.5.D., shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager.

A.

Plant Operations 1.

Integrated and system procedures for normal startup, operation and shutdown of the reactor and all systems and components involving nuclear safety of the facility.

2.

Fuel handling operations 3.

Actions to be taken to correct specific and foreseen potential or actual malfunction of systems or components including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up actions required after plant protective system actions have initiated.

4.

Surveillance and testing requirements that could have an effect on nuclear safety.

5.

Implementing procedures of the security plan.

6.

Implementing procedures of the emergency plan, including procedures for coping with emergency conditions involving potential or actual releases of radioactivity.

7.

Implementing procedures of emergency plans for coping with earthquakes and floods.

The flood emergency plan shall require plant shutdown for water levels at the site higher than 692 feet above MSL.

8.

Implementing procedures of the fire protection program.

9.

Implementing procedures for the Process Control Program and offsite Dose Calculation Manual including quality control measures.

Drills on the procedures specified in A.3 above, shall be conducted as a part of the retraining program.

Drills on the procedures specified in A.6. above, shall be conducted at least semiannually, including a check of communications with offsite support groups.

B.

Radiological Radiation control procedures shall be maintained and made available to all plant personnel. These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10CFR20. This radiation protection program shall be organized to meet the requirements of 10CFR20.

TS.6.5-3 REV C.

Maintenance and Test The following maintenance and test procedures vill be developed to satisfy routine inspection, preventive maintenance programs, and operating license requirements.

1.

Routine testing of Engineered Safeguards and equipment as required by the facility License and the Technical Specifications.

2.

Routine testing of standby and redundant equipment.

3.

Preventive or corrective maintenance of plant equip-i ment and systems that could have an effect on nuclear sa fe ty.

4.

Calibration and preventive maintenance of instrumentation that could af fect the nuclear safety of the plant.

5.

Special testing of equipment for proposed changes to

[

operational procedures or proposed system design changes.

D.

Process Control Program (PCP) i The PCP shall be approved by the Commission prior to initial implementation. Changes to the PCP shall satisfy the following requirements:

1.

A description of changes shall be submitted to the Commission with the Semi-Annual Radioactive Effluent Release. Report. for the period in which the change (s) were made. This submittal shall contain:

a.

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

a determination that the change did not reduce the overall conformance of the solidified waste product to existing '

criteria for solid wastes; and c.

documentation of the fact that the change has been reviewed I

and found acceptable by the Operations Committee. -

i 2.

Shall become effective upon review and acceptance by the Opera-tions Committee.

TS.6.5-4 REV E.

Offsite Dose Calculation Manual (ODCM)

The ODCM shall be approved by the Commission prior to initial implementation.

Changes to the ODCM shall satisfy the following requirements:

1.

Shall be submitted to the Commission with the Semi-Annual Radio-l active Effluent Report for the period in which the change (s) were made effective. This submittal shall contain:

a.

sufficiently detailed information to totally support the ra tionale for the change without benefit of additional or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).

b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.

documentation of the fact that the change has been reviewed -

and found acceptable by the Operations Committee.

2.

Shall become effective upon review and acceptance by the -

Operations Committee.

F.

Temporary Changes to Procedures Temporary changes to procedures described in A,B,C,D, and E above, which do not change the intent of the original' procedure may be made with the concurrence of two individuals holding senior operator licenses.

Such changes shall be documented, reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager within one month.

TS.6.7-2 REV Occupational Exposure Report.(

An annual report of 2.

occupational exposure covering the previous calendar year shall be submitted prior to March 1 of each year.

The report should tabulate on an annual basis the number of station, utility and other personnel (including con-tractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g., reactor opertions and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80%

of the total whole body dose received from external sources shall be assigned to specific major work functions.

3.

Monthly Operating Report. A monthly report of operating statistics and shutdown experience covering the previous month shall be submitted by the 15th of the following month to the Office of Management Information and Program Control, U S fluclear Regulatory Commission, Washington, DC 20555.

4.

Steam Generator Tube Inservice Inspection.

The results of steam generator tube inservice inspections shall be reported within 90 days of January 1 for all inspections completed during the previous calendar year. These reports shall in-clude; (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection, and (3) identification of tubes plugged.

1/ This report supplements the requirements of 10CFR20, section 20.407.

If 13CFR20, Section 20.407 is revised to include such information, this Specification is unnecessary.

TS.6.7-3 REV 5.

Semiannual Radioactive Effluent Release Report.

Routine radioactive effluent release reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days af ter January 1st and July 1st of each year.

The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents as outlined in Appendix B of Regulatory guide 1.21, Revision 1, June, 1974, with data summarized on a quarterly basis.

The report to be submitted 60 days af ter January 1 of each year shall include an assessment of the radiation doses from radioactive effluents released from the plant during the previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to individuals due to their activities inside the site boundary (Figures 3.9-1 and 3.9-2) during the report pe riod. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM) or standard NRC computer codes.

The report to be submitted 60 days af ter January 1 of each year shall also include an assessment of radiation doses to the likely most exposed number of the general public from reactor releases and other nearby uranium fuel cycle sources (including doses froma primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

The radioactive effluent release reports shall include the following information for solid waste shipped offsite during the report period, a.

container volume, b.

total curie quantity (specify whether determined by measurement or estimate).

c.

principal radionuclides (specify whether determined by measure-ment or estimate),

d.

type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e.

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

solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent releases reports shall include unplanned releases from the site of radioactive materials in gaseous and liquid effluents on a quarterly basis, changes to the ODCM, a description of changes to the PCP, a report of when milk or vegetable samples cannot be obtained as required by Table 4.16.1, and cha'nges in land use resulting in significant increases in calculated doses.

6.

Annual Summaries of Meteorological Data.

An annual summary of meteoro-logical data shall be submitted for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability at the request of the Commission.

L__

TS.6.7-4 REV B.

Reportable Occurrences 4

Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of occurrence.

In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date. Paless explicitly stated, the requirements of this section do not apply to the fire protection systems and measures contained in Sections 3.14/4.16, the radiological effluent limitations and measures in Sections 3.9/4.17, or the radiological environmental monitoring program in Section 4.10.

Reporting requirements have been separately specified in those sections.

1.

Prompt Notification With Written Followup.

The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph,

[

mailgram, or facsimile transmission to the Administrator of the appropriate Regional NRC Office, or his designate no later than the first working day following the event, with a written followup report within two weeks.

The written followup report shall include, as a minimum, a completed copy of a licensee l

event report form.

Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

(a) Failure of the reactor protection system or other systems subject to limiting. safety system settings to L

initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.

Note: Instrument drift discovered as a result of testing need not be reported _under this item but may be reportable under items B.1(e), B.l(f), or B.2(a) below.

(b) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications.

Note: If specified action is taken when t system-is found to be operating between the-most conservative and the least conservative aspects of a limiting condition for operation listed in the technical specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this item, but it may be reportable under item.B.2(b) below.

(c)' Abnormal degradation discovered'in fuel cladding, reactor coolant pressure boundary, or primary comtainment..

Note: Leakage of valve packing or. gaskets within the limits for identified leakage set forth:in technical spec-ifications need not be-reported under this item.

TS.6.7-6 REV (j) Release of radioactive material in liquids from the site to the

. unrestricted areas in excess of the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radio-nuclides other than dissolved or entrained noble gases or in excess of 2 x 10-4 uci/ml for total dissolved and entrained noble gases.

(k) Release of radioactive material in gases from the site to un-restricted areas at a rate which exceeds the following dose rates:

For noble gases --

500 mrem / year to the total body or 3000 mrem / year to the skin For radioiodines - 1500 mrem / year to any organ and particulates with half-lives greater than eight days (1) Exceeding the limits for the storage of radioactive materials in outside tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.

-2.

. Thirty Day Written Reports. The reportable occurrences discussed below.shall be the subject of written reports to_the Director of-the appropriate Regional Office within thirty days of occurrence of the event.

The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report shall be supplemented, as

.needed, by additional: narrative material to provide complete explanation of the. circumstances surrounding the event.-

J.

l (a)- Reactor protection system or' engineered safety feature l

instrument, settings.which are found to be less conserva-tive than: those established by the technical specifications but which do not prevent the. fulfillment.of the functional i

requirements-of affected systems.'

-(b) Conditions leading to operation in a' degraded mode per--

- mitted by a limiting condition-for operation or plant -

j.

-shutdown required by a limiting condition for operation.-

Note: Routine surveillance' testing,-instrument calibra--

tion, or preventative. maintenance which require 1

system-configurations as - described in items 'B.2(a) and B.2(b) need.not im reported except where test results themselves. reveal a degraded mode as described above.

. (c) Observed inadequacies in the implementation of admin-istrative or procedural controls which threaten to cause reduction of. degree of redundancy provided JLn reactor protection. systems gor engineered safety : feature systems.

TS.6.7-7 REV (d)

Abnormal degradation of systems other than those specified in item B.1(c) above designed to _contain radioactive material resulting from the fission process.

Note:

Sealed sources or calibration sources are not included under this item.

Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

(e)

An unplanned offsite release of 1) more than one curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:

1.

A description of the event and equipment involved.

2.

Cause(s) for the unplanned release.

3.

Actions taken to prevent recurrence.

4.

Consequences of the unplanned release.

C.

Environmental Reports The reports listed below shall be submitted to the Administrator of the appropriate Regional NRC Office. or his designate:

1.

Annual Radiation Environmental Monitoring Report i

(a) Annual Radiation Environmental Monitoring Reports cover-ing the_ operation of the program during the -previous calendar year shall be submitted prior to May 1 of each -

year.

(b) The Annual Radiation Environmental Monitoring Reports shall.

-include summaries, interpretations, and an analysis of trends.

of the results of the radiological environmental surveillance activities for the report period, including a comparison with -

preoperational studies, operational controls (as ' appropriate),

and previous environmental surveillance reports and an assess-ment of the observed impacts of the plant operation on the -

environment._ The reports shall also include the results of

, land use censuses required by' Specification.4.-10.B.l.

_ If harmful effects or evidence of -irreversible damage -are - detected i by the monitoring, the: report shall _ provide an analysis of the '

problem and a planned. course of action to alleviate _the' problem.

(c)

The Annual Radiation Environmental Monitoring Reports shall.

-include summarized and tabulated results in the. format of Regulatory Guide 4.8, December 1975 of all: radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with.

the report, the report shall be submitted noting and explaining ;

the reasons for.the ' missing results. The missing data shall;

' be submitted as.scon as ' possible in a: supplementary report.

-t 5

TS.6.7-8 REV (d)

The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 4.10.C.I.

2.

Special Reports (a) When radioactivity levels in samples exceed limits specified in Table 4.10-3, a Special Report shall be submitted within 30 days from the end of the affected calendar quarter.

For certain cases involving long analysis time, determination of quarterly averages may extend beyond the 30 day period.

In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period. to be followed by the Special Report as soon-as practicable.

3.

Other Environmental Reports (non-radiological, non-aquatic)

Written reports for the following items shall be submitted to the appropriate NRC Regional Administrator:

a.

Environmental events that indicate or could result in a sign-ificant environmental impact causally related to plant operation.

~

The following are examples:

excessive bird impaction; onsite plant or animal disease outbreaks; unusual mortality of any.

species protected by the' Endangered Species Act of 1973; or increase in nuisance organisms or conditions. This report shall be submitted within 30 days of the event and shall (a),

describe, analyze, and evaluate the event, including extent-and magnitude of the impact and plant operating characteristics,

-(b); describe ' the probable. cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken ' to preclude repetition of. the..

event' and to ' prevent similar occurrences involving 'similar; compon'ents or systems, and (e) indicate = the agencies notified

. and their preliminary responses.

b.

Proposed changes, test or experiments whichl may result in a-_

significant increase in _any adverse environmental impact which was not previously reviewed or evaluated.in the Final-Environ-mental Statement or supplements thereto.: This report.shall include an evaluation of the environmental impact of the -

proposed activity and.shall be submitted 30 days priorf to implementing the proposed change,-l test or experiment.

1 4

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