ML19263E197
| ML19263E197 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/01/1979 |
| From: | Wachter L NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML19263E196 | List: |
| References | |
| NUDOCS 7906050292 | |
| Download: ML19263E197 (62) | |
Text
e s UNITED STATES NUCLEAR REGUIATORY COMMISSION
.N O RIMEP'. STATES POWER COMPANY PRAIRI" ' IAND NUCLEAR GENERATING PIANT Docke t No. 50- 282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSE NO. DPR-42 & DPR-60 (License Amendment Request Dated June 1,1979)
Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specifications as shown on the att ehments labeled Exhibit A snd Exhibit B. Exhibit A describes the pr i . sed changes along with reasons for the change. Exhibit B is a set of Technical Specification pages incorporating the proposed changes.
This request contains no restricted or other defense information.
NORTIERN STATES POWER COMPANY By 6 /
/' L J Wachter Vice President, Power Production &
System Operation On this 1st day of Junc , 1979 , before me a notary public in and for said County, personally appeared L J Wachter, Vice President, Power Production & System Operation, and first being duly sworn acknowledged that he is authorized to execute this document in behalf of Northern States Power Company, that he knows the contents thereof and that to the best of his knowledge, information and belief, the statements made in it are true and that it is not interposed for delay.
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!! #% DENISE E. HALVORSON !!
- i NOTARY PUBUC . MfNNESOTA ll
- I HENNEPIN COUNTY
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e 1 EXHIBIT A PRAIRIE ISLAND NUCLEAR GENERATING PLAh"r LICENSE AMENDMENT REQUEST DATED JUNE 1,1979 PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS APPENDIX A 0F OPERATING LICENSES DPR-42 & DPR-60 Pursuant to 10 CFR 50.59, the holders of Operating License DPR-42 and DPR-60 propose the following changes to Appendix A, Technical Specifications:
PROPOSED CHANGES Revise the Technical Specifications as shown in Exhibit B.
REASON FOR CHANCES These changes are being proposed at the request of the NRC staff. They are needed to impicment the requirements of 10 CFR Part 50, Appendix I, and 40 CFR Part 190.
Guidance for the preparation of these proposed changes was included in a letter, dated November 15, 1978, from Mr. Brian Grimes, Assistat.t Director for Engineer-ing and Projects, Division of Operating Reactors, USNRC. Included with this letter were model radiological effluent Technical Specifications, " Draft Radiological Effluent Technical Specifications for PRR's", NUREG-0472, October, 1978.
Additional guidance was provided at a Regional meeting in Chicago on November 28, 1978.
The proposed changes included in Exhibit B conform to the guidance received from the NRC staff except where it was not practical or possible to do so due to site specific features or technical difficulties. Where it was necessary to depart from the guidance contained in NUREG-0472, alternate requirements intended to provide an equivalent high degree of protection to the health of the public have been proposed.
The changes contained in Exhibit B differ from the guidance provided by the NRC staff in the following areas:
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EXHIBIT A Area of Departure from NRC Guidance Reason
- 1. IRELEG-0472, Specification 1, Definitions The definition of solidification The definition of solidification in has been generalized. The defini- the model Technical Specifications is tion of Limiting Condition for too restrictive. It implies all Operation (LCO) was expanded to note solidified waste is blended into a that the requirements listed in the homogeneous mass. The definition of an Radioactive Effluent as LCO's are LCO has been expanded to permit the term not nuclear safety related, to continue to represent conditions essential to safety while at the same time allowing the new Radioactive Effluent requirements to be listed using the existing Technical Specification format.
Refer to Exhibit B, pages TS.1-4 and TS.1-7.
- 2. NUREG-0472, Specifications 4.3. 3. 8.1, 4. 3. 3. 8. 3, 4.3.3.9.1. and 4.3.3.9.3 These specifications have been These specifications refer to documentation deleted. of effluent montioring surveillance and setpoint determinations. They are unnecessary since the Administrative Controls Section of the Technical Specifications addresses the requirements for written procedures and documentation of results.
- 3. NUREG-0472, Tables 3.3-12 and 4.3-12 Operability and surveillance Flow measuring devices are not installed requirements for flow measuring in exhaust ducts and design flow rates devices in ventilation exhausts are used. There are no surveillance tests have been deleted. Surveillance applicable to iodire and particulate requirements for iodine and samplers. The analyses reqvfred by Table particulate samplers have been 4.11-2 of NUREG-0472 are the only actions deleted, applicable to these devices.
- 4. NUREG-0472, Specification 4.11.1.1.4 Post-release calculations for Post-release calculations are unnecessary, liquid batch releases to verify Pre-release calculations will provide assurance conformance to 10 CFR 20 have been that 10 CFR 20 limits are not exceeded since deleted. conservative assumptions are made (e.g.
maximum possible effluent release rate and minimum dilution flow). Refer to page TS.4.17-1 of Exhibit B.
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EX111 BIT A Arca of Departure frem NRC Guidance Reason
- 5. NUREG-0472, Specifications 4.11.1.2.2, 4.11.2.1.5, 4.11.2.2.2, 4.11.2.3.2, 4.11.2.5.2, and 4.11.3.1.3 These specifications have been These sections refer to reporting require-deleted. ments which are contained in the Administrative Controls Section of the Technical Specifications. It is not necessary to repeat these requirements in the Surveillance Section.
- 6. NUREG-0472, Table 4.11-1 Analysis requirements for P-32 and Analyses for P-32 and Fe-55 are not Fu-55 have been deleted. practical at the present time. Calculated values for P-32 and Fe-55 dose vill be included in the ODCM to account for these isotopes. Refer to page TS.4.17-2 of Exhibit n and Section 2.3.3 of the Prairie Island ODCM.
- 7. NUREG-0472, Specifications 3.11.1.3 and 3.11.2.4 The projected doses at which radwaste As noted in Supplement No. 1 of the treatment system components must be Prairie Island Appendix I Analysis, dated operated have been increased. One- July 21, 1976, calculated Prairie Island fourth of the Appendix I annual offsite doses with IMR GALE Code source design objectives have been proposed terms were below the Appendix I design on a quarterly basis, objectives with existing radwaste treatment equipment. The margins between the com-puted doses and the design objectives were small in some cases. Using existing equip-ment, the radwaste system operational re-quirements contained in Model Technical Specifications 3.11.1.3 and 3.11.2.4 cannot be met within the cost / benefit criteria established under regulatory guidance.
Further, it has been shown that it is not cost-beneficial to install additional equip-ment. Refer to pages TS.3.9-2 and TS.3.9-4 of Exhibit 3.
- 8. IMREG-0472, Specifications 4.11.1.3.2, 4.11.2.4.2, and 4.11.3.1.1 2} }gg The requirement to demonstrate These requirements would result in unnecessary operability of radwaste treatment occupational radiation exposure. We have and solidification equipment at included proposed wording which requires liquid 1 cast once every 92 days when and gaseous radwaste treatment equipment to operation of the equipment is not be maintained and used to assure the Appendix necessary has been deleted. I design objectives are satisfied. Proposed wording also requires solidification of waste prior to shipment or the shipment will
EXHIBIT A Area of Departure from NRC Guidance Reason not be made. If equipment needed to reduce releases to Appendix I design objective levels or to solidify waste prior to shipping is not available for use when required, a report will be made to the Commission.
- 9. tWREG-0472, Table 4.11J A note has been added to permit Grab samples of ventilation exhaust air basing noble gas ratios on PRR are generally below detectable icvels GA1E Code calculations when grab and quantification of individual noble gas samples are below detectable 1cvels. isotopes is not possible. Basing noble gas ratios on PRR GALE Code calculations when noble gases in grab samples are below detectable icvels is a conservative and reasonable action. Refer to Table 4.17-4, note (c), of Exhibit B.
- 10. NUREG-0472, Specification 4.11.2.7 The requirement to determine gas Tank contents will not exceed a small storage tank contents daily during fraction of the tank activity limit. Tanks filling has been revised to specify in the low level waste system are not filled, a monthly determination for each but " float" on the low level loop. The tank in use. proposed monthly activity determination is adequate. Refer to page TS.4.17-3 of Exhibit B.
- 11. NUREG-0472, Specification 3.12.1.b The special reporting pcriod was The 30-day requirement for special reporting extended. was extended to 45 days to allow sufficient time for all radioanalyses to be completed and reported by the laboratory.
- 12. IRTREG-0472, Specification 3.12.1.c The special reporting period was The 30-day requirement for special reporting extended. Was extended to 45 days to allow sufficient time to obtain replacement sample locations.
It may be necessary to resurvey the area in order to find the best replacement.
- 13. NUREG-0472, Table 3.12-1 Short time composite sampics Weekly surface water-borne grab samples cannot be furnished. are canposited monthly and are used instead of two hour composite samples. The climatic conditions in this area create severe problems in compositing samples over short intervals.
Water level fluctuations, heavy ice cover,
EXHIBIT A Area of Departure from NRC Cuidance Reason and extreme spring break-up conditions prevent any system of composite collection to operate reliably. Normal levels of radioactivity in the river can be readily determined by weekly grab samples composited for monthly analysis. Weekly drinking water grab samples taken from a public water supply system and composited for monthly analysis have been substituted for the two-hour composite sampling procedure.
- 14. NUREG-0472, Table 3.12-2 The census period has been extended The census period, June 1 - October 1, was and the formula for LLD deleted. extended to May 1 - October 31 to provide sufficient time during the growing season to complete the census The formula for determining the LLD for a particular measure-ment system is unnecessarily stringent. The table is sufficient and our laboratory does meet these levels.
- 15. NUREG-0472, Specification 6.9.1.8 The Semi-Annual Effluent Report We do not have computerized data collection will not contain annual dose equipment of the quality needed to gather and summaries based on current process meteorological data on a continuing metcarology. basis. The cost and manpower required to do this cannot be justified on a cost-benefit basis. We will submit summaries of calcula-tions performed in accordance with the ODCM using historical meteorology.
- 16. NUREG-0472, Specification 6.9.1.10 Changes to the ODCM will not be We have agreed to submit these changes as a submitted with the Monthly separate report within 90 days of their Operating Reports. effective date. Refer to page TS.6.5-4 of Exhibit B. Submitting them with the Monthly Operating Reports is not appropriate.
- 17. IRRIEG-0472, Specifications 6.14.2 and 6.15.2 Commission initiated changes We believe Coanission initiated changes have been deleted. to the PCP and ODCM can be handled in the usual way through NRC transmittal of suggested changes followed by licensee review and implementation where appropriate.
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EXHIBIT A Area of Departure from NRC Guidance Reason
- 18. NUREG-0472, Specification 6.15.2 Changes to the ODCM will become Review by the Safety Audit Caanittee (CNRAG) effective following Operations prior to impicmenting ODCM changes is not Committee (URG) review and practical since meetings are generally held app: oval. quarterly. We will schedule review of ODCM changes for the next regular Safety Audit Committee meeting following the date of the change. This is consistent with the current Technical Specification requirements for Safety Audit Committee review of 10 CFR 50, Section 50.59(a) nuclear safety related design changes. Refer to page TS.6.5-4 of Exhibit B.
SAFETY EVALUATION The proposed changes implement the requirenents of 10 CFR Part 50, Appendix I, and 40 CFR Part 190. They are being submitted at the request of the NRC Staff and generally conform to the model technical specifications contained in NUREG-0472. When impicmented, they will enhance the protection provided to the public by requiring adherence to the design objectives of Appendix I to be periodically verified with calculations using models approved by the NRC Staff. One new numerical limit, waste gas storage tank activity, has been calculated and included in these proposed changes (Exhibit B, page TS.3.9-5). This calculation conforms to the methodology contained in NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October, 1978. 2292 302
EXHIBIT B License Amendment Request dated June 1, 1979 Exhibit B consists of revised pages of Appendix A Technical Specifications as listed below: Pages TS-1 TS-iii thru TS-v TS.1-4 TS.1-7 TS.3.9-1 thru TS.3.9-11 Figure TS.3.9-1 (new figure) Table TS.3.9-1 (2 pages) Table TS.3.9-2 (2 pages) Table TS.4.1-1 (Page 3 of 4) TS.4.10-1 thru TS.4.10-2 TS.4.10-3 (new page) Table TS.4.10-1 (4 pages) Tables TS.4.10-2 (2 pages) Table TS.4.10-3 TS.4.17-1 thru TS.4.17-5 (new pages) Table TS.4.17-1 Table TS.4.17-2 Table TS.4.17-3 (2 pages) Table TS.4.17-4 (2 pages) TS.6.2-3 TS.6.2-6 TS.6.5-1 TS.6.5-3 thru TS.6.5-4 TS.6.7-2 thru TS.6.7-3 TS.6.7-6 thru TS.6.7-8 Figures TS.4.10-1 and TS.4.10-2 are deleted by these proposed changes 2292 503
TS-1 REV APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 10 Definitions TS.1-1 2.0 Safety Limits and Limiting Safety System TS.2.1-1 Settings 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 31 Reactor Coolant System TS.3.1-1 3.2 Chemical and 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 end Power Distribution Limits TS.3.10-1 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Shock Suppressors (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 4.0 Surveillance Requirements TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Primary System Surveillance TS.4.2-1 4.3 Reactor Coolant System Integrity Testing TS.4.3-1 4.4 Containment System Tests TS.4.4-1 4.5 Engineered Safety Features TS.4.5-1 46 Periodic Testing of Emergency Power System TS.4.6-1 4.7 Main Steam Stop Valves TS.4.7-1 4.8 Auxiliary Feedwater System 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 Shock Suppressors (snubbers) TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Fire Detectics and Protection Systems TS.4.16-1 4.17 Radioactive Effluent Surveillance TS.4.17-1 2292 304
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 Cotditions 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 Instrunent Operating Corditions for Ventilation Systems 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.12-1 Safety Related Shock Suppressors (Snubbers) 3.14-1 Safety Related Fire Detection Instruments 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 4.2-1 Reactor Coolant System In-Service Inspection Schedule Section 1.0 - Reactor Vessel Section 2.0 - Pressurizer Section 3.0 - Steam Generators and Class A Heat Exchangers Section 4.0 - Piping Systems Section 5.0 - Reactor Coolant Pumps Section 6.0 - Valves 4.2-2 System Boundaries for Piping Requiring Volumetric Inspection Under Examination Category IS-251 J-1 4.2-3 System Boundaries Extending Beyond Those of Table TS.4.2-2 for Piping Requiring Surface Inspection Under Examination Category IS-251 J-1 4.2-4 System Boundaries Extending Beyond Those of Tables TS.4.2-2 and -3 for Piping Excluded f rom Examination under IS-251 but Requiring Visual Inspection (Which need not Require Removal of Insulation) of all Welds during System Hydrostatic Test 4.4-1 Unit 1 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 RCMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instru entation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 2292 ;05
TS-iv REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES (Continued) TS TABLE TITLE 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition 6.7-1 Special Reports 2292 306
TS-v 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 1 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 3.9-1 Prairie Island Nuclear Generating Plant Restricted Area Boundary 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 q 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 Rod Bow Penalty (RBP) Fraction Versus Region Average Burnup 3.10-8 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 Functional Organization for On-site Operating Group 2292 307
TS.1-4 REV G. Limiting Safety System Settings Limiting safety system settings are settings on protective instrunentation that initiate automatic protective action at a level such that safety limits will not be exceeded. H. Limiting Conditions for Operation Limiting conditions for operation are those restrictions on unit operation resulting from equipnent performance capability that must be met in order to assure safe operation of the unit. Section 3.9 of the Technical Specifications contains the regulatory limits related to radioactive effluents. Limits contained in this section are not based on reactor safety considerations and are not limiting conditions for operation for purposes of taking actions directed by these Technical Specifications. I. Operable A system or component is operable when it is capable of performing its intended function in the required manner. The operability of a system or component shall be considered to be established when: (1) it satisfies the Limiting Conditions for Operation in Specification 3.0, and (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requirements. Follcwing removal f rom service, a system or component is considered inoperable until its operability has been reestablished. J. Power Operation Power operation of a unit is cny operating condition that results when the reactor of that unit is critical, and the neutron flux power range instrumentation indicates greater than 2% of rated power. K. Protection Instrumentation and Logic
- 1. Protection S/ stem The protection system consists of both the reactor trip system and the engineered safety feature system. The protection system encompasses all electrical and mechanical devices and circuitry (from sensors through the actuating devices) which are required to operate in order to produce the required protective function. Tests of protection systems will be considered acceptable when overlapped if run in parts.
- 2. Protection System Channel A protection system channel is an arrangement of components and modules as required to generate a single protective action signal when required by a unit condition. The channel loses its ider'ity where single action signals are combined.
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TS.1-7 REV W. Process Control Program (PCP) The PCP is the manual describing the program of sampling, analysis, and evaluation within which Solidification of radioactive wastes from liquid systems is assured. Requirements of the PCP are contained in Section 6.5.E of the Technical Specifications. X. Solidification Solidification means the conversion of radioactive wastes from liquid systems to an immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). 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 and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm and/or trip eetpoints. Requirements of the ODCM are contained in Section 6.5.F of the Technical Specifications. 2292 ?09
TS.3 9-1 REV 39 RADIOACTIVE EFFLUENTS , Applicability Applies at all times to the liquid and gaseous radioactive effluents from the plant and the solidification, packaging, and shipping of solid radioactive waste. Obiective To implement the requirements of 10CFR50 Section 50 36a, Appendix I to 10CFR50 and 40CFR190. Specificctions A. Liquid Effluents
- 1. Concentration
- a. The concentration of liquid radioactive material released at any time from the site to the Mississippi River 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 entrgined noble gases, the concentration shall be limited to 2 x 10 uci/ml total activity.
- b. The concentration of radioactive material in liquid effluent released from the site shall be continuously monitored in accordance with Table 3.9-1.
- c. The liquid effluent continuous monitors having provisions for the automatie termination of liquid releases, as listed in Table 3.9-1, shall be used to limit the concentration of radioactive material released at any time from the site to the values given in 3.9.A.1.a.
- d. When the concentration of radioactive material in liquid released from the site exceeds the limits in (a) above, immediately restcre concentration within acceptable limits and provide prompt notification with written followup to the Commission.
- 2. Dose
- a. The dose or dose commitment to an individual f rom radioactive materials in liquid effluents released to the unrestricted area from each unit shall be limited:
- 1. During any calendar quarter to <1.5 mrem to the total body and to <5 mrem to any organ, and
- 2. During any calendar year to <3 mrem to the total body and to <10 mrem to any organ.
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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 corec-tive actions to be taken to reduce the releases of radioactive materials in liquid effluents.
- 3. Liquid Radwaste System
- a. Liquid radwaste system components shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose due to liquid effluent released to the unrestricted area from either unit when averaged over the current calendar quarter would exceed 0.75 mrem to the total body or 2.5 mrem to any organ.
- b. With radioactive liquid 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 information:
- 1. Identification of equipment or sub-systems not functional and the reason.
- 2. Action (s) to be taken to restore equipment to functional status.
- 3. Summary description of action (s) taken to prevent a recurrence.
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TS.3 9-3 REV B. Gaseous Effluents
- 1. Dose Rate
- a. The dose rate at any time in the unrestricted area (Figure 3 9-1) due to radioactive materials released in gaseous effluents f rom the site 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 <3000 mrem / year to the skin, and
- 2. The dose rate limit for all radioiodines, radioactive parti-culates, and radionuclides other than noble gases with half-lives greater than eight days shall be <1500 mrem / year to any organ ,
- b. The concentration of radioactive material in gaseous effluent released from the site shall be continuously monitored in accordance with Table 3 9-2.
- c. The continuous noble gas effluent monitors having provisions for the automatic termination of gaseous releases, as listed in Table 3.9-2, shall be used to limit offsite dose rates within the values established in Specification 3.9.B.I.a.
- d. 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 Cases
- a. The air dose in unrestricted areas due to noble gases released in gaseous effluents from each unit shall be limited to the following values:
- 1. During any calendar quarter, to 15 mrad for gamma radiation and 110 mrad for beta radiation, and
- 2. During any calendar year, to 110 mrad for gamma radiation and 120 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) to be taken to reduce the releases of radioactive noble gases in gaseous effluents.
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TS.3 9-4 REV
- 3. Dose f rom Radioiodines, Radioactive Particulates, and Radionuclides Other Than Noble Cases With Half-Lives Greater Than Eight Days
- a. The dose to any organ of an individual in the unrestricted area due to radioiodines, radioactive particulates, and radionuclides other than nobic gases released in gaseous effluents from each unit shall be limited to the following:
- 1. During any calendar quarter to <7.5 mrem, and
- 2. During any calendar year to <15 mrem
- b. With the calculated dose from the release of radioiodines, radioactive particulates, or radionuclides other than noble gases with half-lives greater than eight days in gaseous >
effluents exceeding the limit (s) in (a) above, within 30 days l submit to the Commission a special report which identifies the cause(s) for exceeding the limit (s) and defines the corrective ; actions to be taken to reduce the releases of radioiodines, ' radioactive particulates, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents.
- 4. Gaseous Radwaste Treatment System
- a. Vaste gas treatment system components shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected air dose due to gaseous effluents released to the unrestricted area from either unit when averaged over the current calendar quarter would exceed 2.5 mrad for gamma radiation or 5 mrad for beta radiation.
- b. With gaseous waste being discharged for more than 31 days 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 information:
- 1. Identification of equipment or subsystems not functional and the reason.
- 2. Action (s) taken to restore equipment to functional status.
- 3. Summary description of action (s) taken to prevent a recurrence.
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'"S . 3. 9-5 iEV
- c. Except as provided for in (d) below, the concentration of oxygen at the outlet of each operating recombiner shall be limited to 12% by volume.
- d. With the concentration of oxygen measured at the outlee of operating recombiner(s) >2% by volume, restore the concentration of oxygen to 12% by volume within 48 hours.
- e. With the concentration of oxygen at the outlet of operating recombiner(s) >4% by volume, immediately suspend all additions of oxygen.
- f. The quantity of radioactivity contained in each gas storage tank shall be limited to $78,800 curies of noble gases (con-sidered as dose equivalent Xe-133).
- g. The radioactive gas contained in the waste gas holdup system i shall not be deliberately discharged to the environment ,1 during unfavorabic wind conditions. 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.
The containment shall be purged during power operation through charcoal and particulate filters of the in-service purge system.
- b. Prior to and purging containment during power operation, the sampling and analysis specified in Table 4.17-4 shall be completed.
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TS.3 9-6 REV C. Solid Radioactive Waste
- 1. Solid Radwaste System
- a. The solid radwaste system shall be maintained and used for Solidification.
- b. The Solidification of wet solid wastes shall be conducted in accordance with the Process Control Program (PCP).
- c. The requirements of the 10 CFR Part 20 and 10 CFR Part 71 shall be satisfied prior to shipment of solid radioactive wastes from the site.
- d. When any of the requirements of the PCP are not complied with, suspend shipments of the affected wastes until conformance is achieved.
- 2. Verification of Solidification
- a. The PCP shall be used to verify Solidification.
- b. If any test specimen fails to verify Solidification, pro-cessing of wet solid wastes shall be suspended until such time as corrective action is taken and subsequent testing of additional specimens confirms Solidification.
2292 315
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 it liquid or gaseous effluents 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, prepare and submit within 90 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 exposur i are less than the 40 CFR Part 190 standard, either apply to ti.e 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.
2292 316
TS.3.9-8 REV Bases: A. Liquid Effluents Specification 3.9.A.1 is provided to ensure that the concentration of radio-active materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in the Mississippi River 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(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. Specification 3 9.A.2.a is provided to implement the requirements of Sections 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 ef fluents will be kept "as low as is reasonably achievable." Consider-ing that the nearest drinking water supply using the receiving 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 radwaste treatment system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the 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. It 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. These allocations will be added to the releases which can be specifically attributed to each unit to obtain the total release per unit. Restrictions on the quantity of radioactive liquid material contained in tanks are not required. No radioactive, or potentially radioactive, liquids are stored in outdoor tanks. 2292 517
TS.3 9-9 REV B. Caseous Effluents Specification 3 9.B.1.a is provided to ensure that the dose rate at anytime at the restricted area boundary from gaseous ef fluents 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. These limits provide reasonable assurance that radioactive ! material discharged in gaseous ef fluents will not result in the exposure of an individual in an unrestricted area, 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 restricted area boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion f actor above that for the restricted area boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the restricted area boundary to .1500 mrem / year to the total body or to 13000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to .11500 mrem / year. 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 assure that the releases of radioactive material in gaseous ef fluents will be kept "as low as is reasonably achievable." Specification 3.9.B.3.a is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. Specification 3.9.3.4.a provides assurance that the waste treatment system will be available for use whenever gaseous wastes are released to the environment. The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID 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. 2292 318
TS.3.9-10 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 belcw 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 system with two potential sources of oxygen: 1) oxygen added for recombiner oiperation, 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 Ne isolated to prevent oxygen from entering the system from this source. fanks 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 ie 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.S.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 f rom the gaseous radwaste system could be at or near ground level near the cooling towers and remain long encugh 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 gaseous vaste treatment sy stem, containment purge release vent, and spent fuel pool vent are shared by both units. Experience has shown that contributions 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 1. 2292 3l9
TS.3 9-11 REV C. Solid Radioactive Waste Specification 3.9.C.1 provides assurance that the solid radwaste system will be used 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 Criteria 60 of Appendix A to 10 CFR Part 50. D. Dose from All Uranium Fuel Cycle Sources Specification 3.9.D is provided to satisfy the reporting requirements of 40 CFR Part 190. 2292 520
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TABLE TS.3 9-1 (Pg 1 of 2) REV TABLE 3 9-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT OPERABLE APPLICABILITi ACTION
- 1. Gross Radioactivity Monitors Providing Automatic Termination of Release
- a. Liquid Radwaste Ef fluent Line 1 During taleases 1
- b. Steam Generator Blowdown 1 During releases 2 Effluent Line (Unit No. I and Unit No. 2)
- 2. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line 1 During releases 4 requiring throt-tling of flow
- b. Steam Generator Blowdown 1 During releases 4 Effluent Line #
- 3. Continuous Composite Samplers and Sampler Flow Monitor
- a. Steam Generator Blowdown 1 During releases 2 Effluent Line i
- b. Turbine Building Sumps Effluent 1 During releases 3 Line #
(Unit No. I and Unit No. 2)
- 4. Component Cooling Water System 1 At all times 5 Monitor
- 5. Discharge Canal Monitor 1 At all times 5 2292 J22
# Plant modifications are required to provide this instrumentation. These modifi-cations will be completed prior to July 1, 1980. In the interim, alternate methods may be used for flow measurement and weekly grab samples may be sub-stituted for composite samples.
TABLE TS.3 9-1 (Pg 2 of 2) REV TABLE 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, at least two independent samples are analyzed in accordance with Specification 4.17.A.1.d and at least two qualified members of the plant staf f independently verify the release rate calculations and discharge valving. 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 14 days provided grab samples are analyzed for gross radjoactivity(betaorgamma)atalimitofdetectionofatleast 10 uCi/ gram;
- 1. At least once per 8 hours when the srecific activity of the secondary coolant is > 0.01 uCi/ gram dose eqtuvalent I-131.
- 2. At least once per 24 hours when the specific activity of the secondary coolant is < 0.01 uC1/ gram dose equivalent I-131.
A_fl0N 3 With the number cf channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, at least once every 24 hours grab samples shall be collected. These samples shall be composited and analyzed for gross radioactivity (beta or gamma) once each week at a lower Ibnit of detection of at least 10-7 uCi/ml. ACTION 4 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 alternate methods are used for flow determination. ACTION 5 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, at least once every 24 hours grab samples shall be collected and analyzed for gross radioactivity (beta or gamma) at a lower Ibnit of detection of at least 10-7 uCi/ml. 2: 2 9 2 a 2 3
TABLE TS.3 9-2 (Pg 1 of 2) REV TABLE 3 9-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION
- 1. Waste Gas Holdup System 2 During system 3 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 2 Monitor *
- b. Iodine Sampler 1 During re' .ases 4 Cartridge
- c. Particulate Sampler 1 During releases 4 Filter
- d. Sampler Flow 1 During releases 1 Measuring Device
- 3. Air Ejector Noble Gas 1 During power 5 Monitors (Each Unit) operation
- Noble gas activity monitors provide automatic termination of releases. They are provided at all release points except the Radwaste Building.
22"92 324
TABLE TS.3.9-2 (Pg 2 of 2) REV TABLE 3 9-2 (Continued) TABLE NOTATION ACTION 1 With the number of channels operable less than required by the Minimum Channels operable requirement, within 24 hours provide alternate methods for determining flow. ACTION 2 With the number of channels operable less than required by the Minimum Channels operable requirement, within 24 hours install auxiliary sampling equipment providing automatic isolation capability (if applicable). ACTION 3 With the number of channels operabla less than required by the Minimum Channels operable requirement, imediately isolate the oxygen supply to the recombiners. ' ACTION 4 With the numbers of channels operable less than required by the Minimum Channels operable requirement, within 24 hours install auxiliary sampling equipment. ACTION 5 With the number of channels operable less than required by the Minimum Channels operable requirement, within 24 hours install auxiliary sampling equipment.
TABLE TS.4.1-1 (Page 3 of 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. Accumulator Level S R R NA and Pressure
- 23. Steam Generator Pressure S R M NA
- 24. Turbine First Stage S R M NA Pressure
- 25. Emergen'cy Plan *M R M NA Includes those named in the emergency Radiation 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. hOspagngegtShell M NA NA NA Includes those used per Spec. 3.6 D.
- 29. Containment Air Temperature M NA NA NA Includes those used per Spec. 3.6 C.
E
- 30. Environmental Monitors M NA NA NA Include.: ti.ase used per Spec. 4.10 E e
- 31. Seismic Monitors R p) R NA NA Includes those reported in Item 4 of ."
- 32. Coolant Flow - RTD S PA) R M NA Table TS.6.7-1 .'*e Bypass Flowmeter 'I) M r\)
- 33. CRDM Cooling Shroud S NA R NA FSAR page 3.2-56 Q Exhaust Air Temperature g
- 34. Reactor Gap Exhaust S' PN) NA R NA FSAR page 5.4-2
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TS.4.10-1 REV 4.10 RADIATION ENVIRONMENTAL MONITORING PROGRAM Applicability Applies to the periodic monitoring and recording of radioactive effluents found in the plant environs. Obiective ; To provide for measurement of radiation levels and radioactivity in the site l environs on a continuing basis. Specification A. Sample Collection and Analysis The Radiation Environmental Monitoring Program described in Table 4.10-1 shall be conducted. Radioanalysis shall be conducted meeting the requirements of Table TS.4.10-1. A map and a table identifying the locations of the sampling points are found in the Offsite Dose Calculation Manual (0DCM). The Table provides further clarification of these locations by giving a descriptive name to each location and the distance and direction from the plant.
- 1. 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.
- 2. 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 effort shall be made to complete corrective action prior to the end of the next sampling period.
- 3. Whenever the level of radioactivity in an environmental sample medium at one or more of the locations specified in Table 4.10-1 exceeds the limits of Table 4.10-3, when averaged over any calendar quarter, a Special Report shall be submitted within 45 days from the end of the af fected calendar quarter. The report shall include an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table TS.4.10-3 to be exceeded. The report is not required if the elevated level of radioactivity was not the result of plant effluents. However, if such an event occurs, it shall be reported and described in the Annual Radiation Environmental Monitoring Report.
- 4. Although deviations from the required sampling sche.dule are permitted under Paragraph 2 above, whenever milk or leafy green vegetables can no longer be obtained f rom the designated sample locations required by Table 4.10-1, a Special Report shall be submitted within 45 days of the scheduled sample date. The report will explain why the samples can no longer be obtained and will identify the new locations which will be added to the monitoring program as soon as practicable.
2292 ;27
TS.4.10-2 REV 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 thy 1 and October 31 by door to door survey, aerial survey, or by consulting local agricultural authorities or associations. If the land use census identifies a location that yields a calculated dose or dose commitment greater than at a location currently being sampled, the new location shall be added to the program as soon as practicable. The sampling location having the lowest calculated dose or dose commitment may be deleted from the program af ter October 31 of the year in which this land use census was conducted. This sampling location change shall be reported in the Annual Radiation Environmental Monitoring Report.
C. Interlaboratory Comparison Program
- 1. Analyses shall be performed on radioactive materials supplied as part of an interlaboratory comparison program.
- 2. The results of analyses performed as a part of the above required program shall be included in the Annual Radiation Environmental Monitoring Report.
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 raal
- ion are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. Af ter a specific program has been in effect for at least three years of opctation, program changes may be initiated based on this experience.
The detection capabilities required b,; Table 4.10-2 are state-of-the-art for routine environmental measurements 'e induatrial laboratories, and the LLD's for drinking water meet the requirement 6f AJCFR141. 22"92 ;28
TS.4.10-3 REV Reporting of Gamma Scan Analyses will be limited to four isotopes, based on the fact that the concentration of each isotope is about 10% of the total curies of activity of the fission products ten days af ter reactor shutdown, but are found in fallout from thermo-nuclear testing approximately ten days af ter detonation, in extremely small percentages of total activity. The choice of isotopes are Cesium-137, Corium-144, Zirconium-95 and Potassium-40. Another reason for their choice is that their half-lives are long enough not to introduce significant errors in computatin by delays in analyses. Potassium-40 was chosen as a calibration monitor and should not be considered a radio-logical inpact indicator. B. Land Use Census This specification is provided to assure that changes in the use of unrestricted areas are identified and that modifications to the program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10CFR50. C. Interlaboratory Comparison Prog am 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 menitoring in order to demonstrate that the results are reasonably valid. 2292 329
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_ TABLE TS.4.10-1 (Page 2 of 4) Type of Sample Type of Analysis Collection Site Collection Frequency Aquatic Vegetation GS 1 Sample upstream of plant Semi-annually 1 Sample downstream of plant (when a.silable) Fish (1 sample each GS 2 Samples upstream of plant Semi-annually of two game specie) 2 Samples downstream of plant (when available water & ice conditions permitting) 1' 1 Cs,* 1 Sample at the offsite dairy Monthly farm having the highest D/Q 89' 90Sr* 3 Samples from dairy farmsi ggle-ulated to have doses from I> 1 mrem./yr 1 Sample from 10-20 mile Topsoil GS, Sr From the 4 air sampling locations, Once every 3 years and from 5 fields in the vicinity of the plant Natural Vegetation GS, 131 7 1 Sample from field having highest Semi-annually , D/Q (same as for milk) Qh 1 Sample from a field downwind p of the plant (within 2 miles) g 1 Sample from 10-20 mile ,m (Same as for milk) f'
- Performed only on D/Q and Control Samples o g,
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TABLE 4.10-1 (Page 3 of 4) Type of Sample Type of Analysis Collection Site Collection Frequency Cultivated Crops Leafy Green I 1 Sample from nearest garden Annually (at harvest, Vegetables 1 Sample from 10-20 miles location if available) Co r-' GS 1 Sample from highest D/Q farm Annually (at harvest, 1 Sample from 10-20 miles if available) Air CB, CS (M) 2 locations in different sectors Weekly (Particulates) having the highest calculated ground level concentrations I location near residence having highest X/Q value 1 location near closest community 2 locations within 9-20 miles Air 131 1 location near residence having Weekly 7 (Radioiodine) highest X/Q value g 1 location near closest community g I location within 10-20 miles N d b N
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TABL 'S.4.10-1 (Page 4 of 4) Type of Sample Type of Analysis Collection Site Collection Frequency Air Gamma dose (TLD) 2 dosimeters at each air particulate Quarterly sampling location Coding System: GB - Gross beta GS - Gamma scan M - Monthly Q - Quarterly
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TABLE TS.4.10-3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Airborne Particulate Water or Ggs F'ish Milk Vege tables Analysis (pCi/1) (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) 11 - 3 3x 10 0 3 0 Mn-54 1 x 10 3x 10 Fe-59 4x 10 2 1x 10 Co-58 1 x 10 3x 10' Co-60 3x 10 1 x 10' Zn-65 3x 10 2x 10 2 Zr-Nb-95 4 x 10 I-131 2 0.9 2 3 1 x 10 Cs-134 30 10 1x 10 3 60 1x 10 3 SJ ! y Cs-l'7 50 20 2x 10 3 70 2 x 10 3 g 4 N Ba-La-140 2 x 10 2 9 h 3 x 10] - y o O u b Ch
TS.4 17-1 REV 4.17 RADIOACTIVE EFM.UENT SURVEILLANCE Applicability: Applies to the periodic monitoring and recording of liquid and gaseous radioactive effluents, or verification of solidification, and verification of equipment operability. Objective: To imolement the requirments of 10CFR50 Secton 50.36a, Appendix I to 10CFR50, 40CFR190. Specification: A. Liquid Effluents
- 1. Dose Rate Monitoring and Calculations
- a. Surveillance of liquid effluent monitoring instrumentation shall be performed as required by Table 4.17-1.
- b. The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.17-3. The results of pre-release analyses shall be used with the calculational methods in the Offsite Dose Calculation Manual (ODCM) to assure that the concentration at the point of release to the Mississippi River is limited to the values given in 3 9.A.1.a.
- c. The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.17-3. The results of the analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release to the Mississippi River are limited to the values given in 3.9.A.1.a.
- 2. Dose Calculations
- a. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM monthly.
- b. Because effluent sampling of the following isotopes is not practical, predicted values specified in the ODCM may be used in the dose calculations required in (a) above:
P-32 Fe-55 2292 337
TS.4.17-2
- 3. Liquid Radwaste System
- a. Doses due to liquid releases to unrestricted areas shall be projected monthly. Releases considered in the projection should include all plant effluents from all liquid radioactive waste management and disposal system components that are planned to be operated at the projected capacity and ef ficiency of each. A projected dose in excess of the limits specified in 3 9.A.3.b indicates that additional components or subsystems of the liquid radwaste treatment system must be placed in service to reduce radioactive materials in liquid effluents.
B. Gaseous Effluents
- 1. Dose Rate Monitoring and Calculations
- a. Surveillance of gaseous effluent monitoring instruments shall be performed as required by Table 4.17-2.
- b. The release rate of radioiodines, radioactive particulates, and radionuclides other than noble gases with half-lives greater than eight days shall be determined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.17-4. Following each analysis, the dose rate due to radioiodines, radioactive particulates, and radionuclides other than noble gases with half-lives greater than eight days shall be determined to be less than the linit in specification 3 9.B.1.a.2 in accordcace with the ODCM.
- 2. Dose Calculations for Noble Gases
- a. Cumulative dose contributions from noble gases in gaseous effluents shall be determined in accordance with the ODCM monthly.
- 3. Dose Calculations for Radioiodines, Radioactive Particulates, and Radionuclides Other Than Noble Gases With Half-Lives Greater Than Eight Days
- a. Cumulative dose contributions from radioiodines, radioactive particulates, and radionuclides other than noble gases with half-lives greater than eight days in gaseous ef fluents shall be determined in accordance with the ODCM monthly.
- b. Because effluent sampling of the following radioisotopes is not practical, predicted values specified in the ODCM may be used in the dose calculations required in (a) above:
C-14 2292 38
TS.4.17-3 REV
- 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 vaste gas holdup system shall be continuously monitored with the explosive gas monitors required 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.
- 5. Atmospheric Steam Dump Monitoring
- a. The 1-131 activity in the steam and water on the secondary side of each steam generator shall be determined as required in Specification Table TS.4.1.2B, Item 8.
- 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 steam and water.
- 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 from 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 1.5 pCi/1.
C. Solid Radioactive Waste
- 1. Verification of Solidification
- a. Specimens to verify Solidification shall be processed 3 2 accordance with the PCP.
2292 339
TS.4.17-4 REV D. Dose from All Uranium Fuel Cycle Sources
- a. Cumulative dose contributions from all plant effluents shall be determined in accordance with Specifications 4.17.A.2.a. 4.17.B.2.a and 4.17.B.3.a.
Basis Radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments will be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20. The operability requirements for instrumentation are consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. Radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments 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 will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1 113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I, Revision 1," April 1977. NUREG-0133, October, 1978, provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113 2292 340
TS.4.17-5 REV 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 ef fluents will be consistent with the methodology provided in Regulatory Guide 1 109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Extimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at the restricted area 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. 2292 341
TABLE TS 4.17-1 REV TABLE 4.17.1 - RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Source Channel Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency Liquid Radwaste Effluent Daily during Monthly Quarterly At least once every Line Gross Radioactivity releases 18 months
- Monitor Liquid Radwaste Effluent Daily during Monthly Quarterly At least once every Line Flow Instruments releases 18 months Steam Generator Blowdown Daily during Monthly Qua rterly At least once every Gross Radioactivity releases 18 months
- Monitors Steam Generator Blowdown -- --
Quarterly At least once every and Turbine Building Sump - 18 months Continuous Composite Samplers Component Cooling Daily during Monthly Quarterly At least once every Water System Monitor releases 18 months Discharge Canal Daily during Monthly Quarterly At least once every Monitor releases 18 months
- An initial calibration shall be performed using standerds certified by the National Bureau of Standards (NBS) or using standards obtained f rom participants in measure-ment assurance activities with the NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. Subsequent calibrations may use sources that have been related to the initial calibration sources.
2292 342
TABLE TS.4 17-2 REV TABLE 4.17.2 - RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Source Channel Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency Waste Gas Holdup System Daily During -- Quarterly At least once every Explosive Gas System 18 months (OxyLan) Manitors 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 At least once every Monitor # Releases 18 months
- Sampler Flow Rate Weekly --
Quarterly At least once every Measuring Device 18 months Air Ejector Noble Gas Daily During Monthly Quarterly At least once every Monitors (Each Unit) Releases 18 months
- An initial calibration shall be performed using standards certified by the National Bureau of Standards (NBS) or using standards obtained from participants in measure-ment assurance activities with the NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. Subsequent calibrations may use sources that have been related to the initial calibration sources.
# Not installed in Radwaste Building vent.
2292 343
TABLE TS.4.17-3 (Pg 1 of 2) REV TABLE 4.17.3 - RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM I
! Sampling Minimum Type of Activity '
Lower Limit of l Liquid Release Type Frequency Analysis Analysis Detection Frequency i (uCi/ml)*'fLLD) Batch Tank Release Each Batch Each Batch a 5 x 10
-7b Principakdga Emitters I-131 1 x 10 -6 One Batch One Batch Dissolved and 1 x 10
-5 Each Month Each Month Entrained Cases Each Batch Monthly H-3 1 x 10~
Composite"
~
Gross alpha 1 x 10 Each Batch Quarterly Sr-89, Sr-90 5 x 10~ Composite" Continuous Releases Continuous 8 Weekly Principal Gamma 5 x 10~ Composite Emitters (d) I-131 1 x 10
-6 Grab Sample Grab Sample Dissolved and 1 x 10" Each Month Each Month Entrained Gases Continuous 8 Monthly H-3 1 x 10-Composite 8 Gross alpha 1 x 10~
Continuous 8 -8 Quarterly Sr-89, Sr-90 5 x 10 Composite 8 2292 ;44
TABLE TS.4.17-3 (Pg 2 of 2) REV TABLE TS.4.17-3 TABLE NOTATION Notes:
- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of a f alsely conclud-ing that a blank observation represents a "real" signal.
- b. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD May be increas9d inversely proportionally to the magnitude of the gamma yield (i.e.,
5 x 10 /I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10CFR20, Appendix B, Table II, Column 2.
- c. 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.
- d. 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.
- e. 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 Effluent Release Report.
- f. 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.
- g. To be representative of the quantities and concentrations of radioactive materials in lliquid 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.
2292 ;45
TABLE TS.4.17-4 (Pg 1 of 2) REV TABLE TS.4.17-4 - RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Type of Sampling Mininum Activity Lower Limit of Gaseous Release Frequency Analysis Analysis Detection,,(LLD Type Frequency (uCi/ml) Each Tank Each Tank Principal Gamma 1 x 10' Waste Gas Release Release Emitters (f) Storage Tank Grab Sample Grab Sample -6
-3 1 x 10 Containment Purge Each Purge Each Purge Principal Gamma 1 x 10 Grab Emitters (f)
Sample -6
-3 1 x 10 Effluents Release Monthly' Monthly Principal Gamma Points (Unit No. 1 Grab Emitters (f) 1 x 10-4b Sample Reactor Bldg, Unit No. 1 Aux Bldg, H-3 1 x 10-6 Unit No. 2 Reactor d
Bldg, Unit No. 2 Continuous Weekly I-131 1 x 10-10 Aux Bldg, Spent Charcoal Fuel Pool, Sample I-133 1 x 10-10 Radwaste Bldg) Continuous Weekly d Principal Gamma 1 x 10
-11 Particulate Emitters (I-131, Sample Others) h Gross alpha -11 Continuous Monthly 1x 10 Particulate Sample Continuous Quarterly
- Sr-89, Sr-90 1 x 10 -11 Composite Particulate Sattple Air Ejector Monthly' Monthly Principal Gamma 1 x 10 Vents Grab Emitters
-0 H-3 1 x 10 2292 ;46
TABLE TS.4.17-4 (Pg 2 of 2) 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 f alsely concluding that a blank observation represents a "real" signal.
- b. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inverself proportionally to the magnitude of the gamma yield (i.e., 1 x 10 /I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10CFR20, Appendix B, Table II, Column 1.
- c. 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 GALE Code noble gas isotopic ratios may be assumed.
- d. Daily sampling shall be required during those period (e.g. refueling) when experience has shown that the release rate of iodines and particu-lates may exceed ten times the long-term average for that release point.
- e. 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.
- f. 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-1134, Cs-137, Ce-141, and Cc-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.
- g. 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 resul t in LLD's higher than reported, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report,
- h. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period sampled.
2292 ;47
TS.6.2-3 REV
- f. Investigation of all events which are required by regulation or technica*. specifications (Appendix A) to be reported to NRC in writing within 24 hours.
- 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 or 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 Offsite Dose Calculation Manual (ODCM).
- 6. Audit - The operation of the nuclear power p.lant 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 N28.7-1972, 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. Periodic review of the audit program sheuld be performed by the SAC at least twice a year to assure its adequacy.
- c. Written reports of the audits shall be reviewed by the Vice President - Power Production & System Operation, by the SAC at a scheduled meeting, and by members of management having responsi-bility in the areas audited.
- 7. Authority The SAC shall be advisory to the Vice President - Power Production &
System Operation.
- 8. Records Minutes shall be prepared and retained for all scheduled meetings of the Safety Audit Committtee. The minutes shall be distributed to the Vice President - Power Production & System Operation, the General Superintendent of Nuclear Power Plant Operation, each member of the SAC and others designated by the Chairman or Vice Chairman within one month of the meeting. There shall be a formal approval of the minutes.
2292 ;48
TS.6.2-6 REV
- t. All events which are required by regulations or Technical Specifications to be reported to the NRC in writing within 24 hours.
- g. Drills on emergency procedures (including plant evacuation) and adequa:y of communication with offsite 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. Anthority 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 Superin-tendent of Nuclear Power Plant Operation 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 miutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Superintendent of Nuclear Power Plant Operation and othars designated by the OC Chairman or Vice Chairman.
- 7. Procedures A written charter for the OC shall be prepared that contains:
- a. Responsibility and authority of the group
- b. Content and cethod of submission of presentations to the Operations Committee
- c. Mechanism for scheduling meetings
- d. Provision for meeting agenda 2292 349
TS.6.5-1 REV 6.5 PLANT OPERATING PROCEDURES Detailed written procedures, including the applicable checkoff 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 Offiste Dose Calculation Manual.
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 tc, 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. 2292 -50
TS.6.5-3 REV C. Maintenance and Test The following maintenance and test procedures will be developed to satisfy routine inspection, preventive maintenance programs, and operating license requirements.
- 1. Routine testing of Enginected 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-ment and systems that could have an effect on nuclear safety.
- 4. Calibration and preventive maintenance of instrumentation that could affect the nuclear safety of the plant.
- 5. Special testing of equipment for proposed changes to operational procedures or proposed system design changes.
D. Temporary Changes to Procedures Temporary changes to procedures described in A, B, and C 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. E. Process Control Program A Process Control Program (PCP) shall be developed to outline the equipment operating procedures, process parameters, and the program of sampling, analysis, and evaluation within which solidification of radioactive wastes from liquid systems is assured. Detailed operating and laboratory are not included in the PCP. The PCP shall be submitted to the Commission prior to initial implementa-tion for their review. Changes to the PCP shall satisfy the following requirements:
- 1. Changes shall be submitted to the Commission with the Ef fluent and Waste Disposal Semiannual Report for the period in which the changes were made. Information submitted to the Commission shall include a) sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informa-tion; 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 f act that the change has been reviewed and found acceptable by the Operations Committee.
2292 .51
TS.6.5-4 REV
- 2. Changes become effective upon review and acceptance by the Operations Committee unless otherwise acted upon by the Commission through written notification.
F. Offsite Dose Calculation Manual (00CM) An Of fsite Dose Calculation Manual (ODCM) shall be developed to outline the methodology and parameters to be used in the calculation of of fsite doses due to radioactive liquids and gases released to the unrestricted area and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm and trip setpoints consistent with the applicable Limiting Conditions for Operation. Methodologies and calculational procedures will be based on the guidance contained in NUREG-0133, October, 1978. The ODCM shall be submitted to the Commission prior to initial imple-mentation for their review. Changes to the ODCM shall satisfy the follow-ing requirements:
- 1. Changes shall be submitted to the Commission within 90 days of the date the changes were made effective. Information submitted to the Commission shall include a) sufficiently detailed information to totally support the rationale of 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 dated, together with appropriate analyses or evaluations justifying the changes b) a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c) a statement that the changes have been reviewed and found acceptable by the Operations Committee and Safety Audit Committee.
- 2. Changes become effective upon review and acceptance by the Operations Committee unless otherwise acted upon by the Safety Audit Committee during their review or by the Commission through written notification.
2292 352
TS.6 7-2 REV
- 2. Occupational Exposure Report. An annual report of 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-ren exposure according to work and job functions, e.g., reactor opertions and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste 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 f rom 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 Nuclear 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 e and nercent of wall-thickness penetration for each indication of aa imperfection, and (3) identification of tubes plugged.
- 5. Effluent and Waste Disposal Semiannual Report. Routine radioactive effluent release reports covering the oparcuion 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 ef fluent releabe reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste shipped from the site using Appendix B of Regulatory Guide 1.21, Revision 1, June, 1974 as gu'Jance. The report shall include an acsessment of the radiation doses from radioactive effluents to individuals in the unrestricted area during each quarter. In addition, the unrestricted area boundary maximum noble gas gamma air and beta air does shall be evaluated. The assessment of radiation doses shall be performed using the methods out-lined in the Offsite Dose Calculation Manual (ODCM). 1/ This report supplements the requirements of 10CFR20, section 20.407. If 10CFR20, Section 20.407 is revised to include such information, this Specification is unnecessary. 2292 353
a' . TS.6.7-3 REV Changes to the Process Control Program (PCP) made during the reporting period shall be described in the report. B. Reportable Occurrences 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. 2292 354
TS.6.7-6 REV
. (j ) Release 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 thgn dissolved or entrained noble gases or in excess of 2 x 10 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 (1) Interruption of fire suppression water supply to any safety-related structure, system or component.
- 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.
(a) Reactor protection system or engineered safety feature 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 requirements of affected systems. (b) Conditions leading to operation in a degraded mode per-mitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation. Note: Routine surveillance testing, instrument calibra-tion, or preventative maintenance which require system configurations as described in items B.2(a) and B.2(b) need not be 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 in reactor protection systems or engineered saf ety feature systems. 22"92 355
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: Scaled sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth b. technical specifications need not be reported under this item. (e) An unplanned release to the unrestricted area 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 radioiodine in gaseous effluents. The report of an unplanned release 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
- 1. Annual Radiation Environmental Monitoring Report (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 report shall include summaries, interpretations, and statistical evaluation of the results of the monitor-ing activities for the report period, including a comparison with preoperational studies, control levels (as appropriate), and previous environmental monitoring reports and an assessment of the observed impact of the plant operation on the environment. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem. 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 soon as possible in a supplementary report. (c) The report shall also include the results of the land use census. (d) The results of analyses performed as part of the Interlaboratory Comparison Program shall be included in the report. 2272 :56
TS.6.7-8 REV
- 2. Special Reports (a) When radioactivity levels in samples exceed limits Specified in Table TS.4.10-3, a Special Report shall be submitted within 45 days from the end of the affected calendar quarter.
(b) Whenever milk or leafy green vegetables can no longer be obtained from the designated sample locations stated in Table TS.4.10-1, a Special Report shall be submitted within 45 days of the scheduled sample date. 2292 357}}