ML20064M546

Radiological Effluent Tech Specs Review Document

ML20064M546
Person / Time
Site:	Brunswick
Issue date:	08/31/1982
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20064M537	List: ML20064M534 ML20064M546
References
NUDOCS 8209080007
Download: ML20064M546 (116)

Text

{{#Wiki_filter:O RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS REVIEW DOCUMENT e BRUNSWICK STEAM ELECTRIC PLANT AdGUST 1982 8209080007 820902 PDR ADOCK 05000324 P PDR

NOTE: This document is for information only and is not a formal request for a Technical Specification change. It is intended to be used as a bases for reviews and discussions that will lead to the inclusion of Radiological Effluent. Technical Specifications (RETS) into the Brunswick Steam Electric Plant's Technical" Specifications. This document consists of revised excerpts from those sections of the Tech Specs which will be affected by the RETS. I t { i t

INDEX SECTION PAGE 1.0 DEFINITIONS (Applicable to RETS) ACTI0N.......................................... 1-1 CHANNEL CALIBRATION............................. 1-1 CHANNEL CHECK................................... 1-1 CHANNEL FUNCTIONAL TEST......................... 1-1 DOSE EQUIVALENT I-131........................... 1-2 FREQUENCY N0TATION.............................. 1-2 GASEOUS RADWASTE TREATMENT...................... 1-2 MEMBER (S) OF THE PUBLIC......................... 1-2 0FFSITE DOSE CALCULATION MANUAL (ODCM)........... 1-2 OP ERABLE - OPERABILITY.......................... 1-2 OPERATIONAL CONDITION........................... 1-3 PROCESS CONTROL PROSRAM(PCP).................... 1-3 P URG E - P URG IN G................................. 1-3 RATED THERMAL P0WER............................. 1-3 SITE B0UNDARY................................... 1-3 SOLIDIFICATION.................................. 1-3 SOURCE CHECK.................................... 1-3 THERMAL P0WER................................... 1-3 UNRESTRICTED AREA............................... 1-4 I l VENTILATION EXHAUST TREATMENT SYSTEM............ 1-4 VENTING......................................... 1-4 11 l i r

INDEX SECTION PAGE 1.0 DEFINITIONS (continued) FREQUENCY NOTATION, TABLE 1.1.......................... 1-5 OPERATIONAL CONDITIONS, TABLE 1.2...................... 1-6 2.0 SAFETY LIMITS NOTE: This section of the Technical. Specifications is not affected by RETS and is not included in this review document. 3.0/4.0 LIMITING CONDITIONS FOR_ OPERATION, SURVEILLANCE REQUIREMENTS, AND BASES (Applicable to RETS) 3/4.11 LIQUID EFFLUENTS 3/4 1-1 3/4.11.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION....... 3/4 1-1 3/4.11.2 LIQUID EFFLUENTS CONCENTRATION.................... 3/4 1-8 3/4.11.3 LIQUID EFFLUENTS D0SE............................. 3/4 1-13 3/4.11.4 LIQUID RADWASTE TREATMENT......................... 3/4 1-15 3/4.11.5 LIQUID HOLDUP TANKS............................... 3/4'l-17 3/4.12 GASEOUS EFFLUENTS......'............................ 3/4 1-19 3/4.12.1 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION...... 3/4 1-19 3/4.12.2 GASEOUS EFFLUENTS DOSE RATE....................... 3/4 1-30 3/4.12.3 GASEOUS EFFLUENTS DOSE-NOBLE GASES................ 3/4 1-36 3/4.12.4 GASEOUS EFFLUENTS DOSE-IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE F0RM................. 3/4 1 3/4.12.5 GASEOUS RADWASTE TREATMENT / VENTILATION EXHAUST TREATMENT......................................... 3/4 1-40 3/4.12.6 MAIN CONDENSER AIR EJECTOR RADIOACTIVITY RELEASE RATE.............................................. 3/4 1-42 3/4.12.7 EXPLOSIVE CAS MIXTURE...'.......................... 3/4 1-44 3/4.12.8 DRYWELL PURGES (MARK I CONTAINMENT)............... 3/4 1-45 111 ,L_ .......-n. .-,. - ~. se-s---+--*-~~~~ ' ~ ~ - + - = - * * - ~ - * * * "~.**","#'*"**""'**~

INDEX SECTION PAGE 3.0/4.0 (continued) 3/4.13 SOLID RADIOACTIVE WASTE........................... 3/4 1-46 3/4.14 40 CFR 190........................................ 3/4 1-48 3/4.14.1 TOTAL D0SE...................................... 3/4 1-48 3/4.15 RADIOLOGICAL ENVIRONMENTAL MONITORING............. 3/4 1-51 3/4.15.1 MONITORING PR0 GRAM.............................. 3/4 1-51 3 / 4.15. 2 LAND US E CENS US................................. 3/4 1-61 3/4.15.3 INTERLABORATORY COMPARISON PROGRAM.............. 3/4 1-63 5.0 DESIGN FEATURES (Applicable to RETS) 5.1 SITE i 5.1.1 EXCLUSION AREA..................................... 5-1 5.1.2 LOW POPULATION Z0NE................................ 5-1 5.1.3 SITE B0UNDARY............'.......................... 5-1" i i 1 1 iv 4 l l l L

INDEX SECTION PAGE 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY....................................... 6-1 6.2 ORGANIZATION ,0ffsite.............................................. 6-1 Facility Staff....................................... 6-1 s 6.3 FACILITY STAFF QUALIFICATIONS........................ 6,6 6.4 TRAINING............................................. 6-6 6.5 REVIEW AND AUDIT. 6.5.1 PLANT NUCLEAR SAFETY COMMITTEE Function......................................... 6-6 Composition...................................... 6-6 Alternates....................................... 6-6 Meeting Frequency................................ 6-7 Quorum........................................... 6-7 Responsibilities................................. 6-7 3 Authority........................................ 6-9 Records.......................................... 6-9 6.5.2 CORPORATE NUCLEAR SAFETY AND QUALITY ASSURANCE AUDIT SECTION Responsibility................................... 6-9 6.5.3 CORPORATE NUCLEAR SAFETY UNIT Function......................................... 6-10 Personnel........................................ 6-10 Subj ec ts Requiring Independent Review............ 6-11 Follow-u-) Act1on................................. 6-12 v

INDEX SECTION PAGE 6.5 REVIEW AND AUDIT (continued) 6.5.4 OPERATION AND MAINTENANCE UNIT Function...................,...................... 6-13 Personnel........................................ 6-15 Reports.......................................... 6-15 6.5.5 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM...... 6-15 6.6 REPORTABLE OCCURRENCE ACTI0N......................... 6-16 6.7 SAFETY LIMIT VIOLATION............................... 6-16 6.8 PROCEDURES........................................... 6-16 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES....... 6-18 6.9.2 SP ECIAL REP 0RTS.................................. 6-27 6.10 RECORD RETENTION.................................... 6-27 6.11 RADIATION PROTECTION PR0 GRAM........................ 6-29 6.12 HIGH RADIATION AREA................................. 6-29 6.13 ENVIRONMENTAL QUALIFICATION......................... 6-30 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM).............. 6-31 6.15 PROCESS CONTROL PROGRAM (PCP)....................... 6-32 6.16 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS.......................... 6-33 vi

O SECTION 1.0 DEFINITIONS t k i I l

1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in. capitalized type and are applicable throughout these Technical Specifications. Only those definitions applicable to the Radiological Effluent Technical Spec-ifications are. included. These teras will be incorporated into the de-finitions section of the Technical Specifications as a whole. ACTION ACTIONS are those additional requirements specified as corollary state-ments to each specification and shall be part of the specifications. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment as necessary of the-chan-nel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHAN-NEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNC-TIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a signal into the channel as close to the sensor as practicable to verify OPERABILITY l

including alarm and/or trip functions. t I

b. Bistable channels - the injection of a signal into the sensor to verify OPERABILITY including ~ alarm and/or trip functions.

f The CHANNEL FUNCTIONAL TEST may be performed by any series of sequen-tial, overlapping or total channel steps such that the entire channel is functionally tested. l l 1-1 I l l l

DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131, uC1/ gram, which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The following is defined equivalent to 1 uCi of I-131 as detennined from Table III of TID-14844," Calculation of Test Reactor Sites": I-132, 28 uCi; I-133, 3.7 uCi;. I-134, insignificant; I-135, 12 uC1. FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. CASEOUS RADWASTE TREATMENT SYSTFM A CASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. MEMBER (S) OF THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include no'n-employees of the licensee who are permitted to use portions of the site for recreational, occupa-tional or other purposes not associated with plant functions. This cate-gory shall not include non-employees such as vending. machine service-men or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for the purposes of protection of individuals from exposure to radiation and radioactive materials. OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANdAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate offsite doses resulting from the release of radioactive gaseous and liquid ef-fluents; the methodology to calculate gaseous and liquid effluent moni-toring instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program. OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified func-tion (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electric power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, t ra in, component or device to' perform its function (s) are also capable of per-forming their related support function (s). 1-2

OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2. PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the. current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demon-strated processing of actual or simulated wet solid wastes will be ac-complished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations and other requirements governing the disposal of the radioactive waste. PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concen-tration or other operating condition, in such a manner that replacement air or gas is required to purify the containment. RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of.2436 MWT. SITE BOUNDARY The SITE BOUNDARY shall be that,line beyond which the land is not owned, leased or otherwise controlled by the licensee, as defined by Figure 5.1.3-1. SOLIDIFICATION l l SOLIDIFICATION shall be the conversion of wet wastes into a form that j meets shipping and burial ground requirements. SOURCE CHECK l A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation. THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coolant. 4 1-3

' UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purpose of pro-tection of individuals from exposure to radiation and radioactive mate-rials OR any area within the SITE BOUNDARY used for residential quarters or industrial, commercial, institutional and/or recreational purposes. VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and in-stalled to reduce gaseous radioiodine or radioactive material in parti-culate form in effluents by passing ventilat-ion or vent exhaust gases through charcoal adsorbers and/or HEPA -filters for the purpose of re-moving todines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING VENTING is the coritrolled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas 'i not provided or required during VENTING Vent, used in system names, does not imply a VENTING process. L i 1 1-4

i TABLE 1.1 FREQUENCY NOTATION 4 NOTATION FREQUENCY' S At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. SM At least once per 16 days. j M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. A At least once per 366 days. R At least once per 18 months (550 days). S/U Prior to each reactor startup. N.A. Not applicable. P Prior to each release. -I i t l .i 4 1-5 ..m.

TABLE 1.2 OPERATIONAL CONDITIONS OPERATIONAL MODE SWITCH AVERAGE COOLANT CONDITIONS POSITIONS TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown

> 212*F 7

4. COLD SHUTDOWN Shutdown 6 212* F

5. REFUELING

• Refuel **

f 212'F I

• Reactor vessel head unbolted or removed and fuel in vessel.***
  • See Special Test Exception 3.10.3.
    • See Special Test Exception 3.10.1.

i 1-6

-1 4 -. r i 5 + b SECTION 3/4.11-15 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS l FOR THE BRUNSWICK STEAM ELECTRIC PLANT 4 r I e BRUNSWICK UNIT 1

3/4.11 LIQUID EFFLUENTS 3/4.11.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION LIMITING CONDITI,0N FOR OP,ERATION 3.11.1 The radioactive liquid effluent monitoring instrumentation channels shown in -Table 3.11.1-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2 are not exceeded. The alarm / trip setpoints shall be determined in 'accordance with the ODCM. APPLICABILITY: As shown in Table 3.11.1-1 ACTION: With a radioactive liquid effluent monitoring instrumentation chan-a. nel alarm / trip setpoints less conservative than required by the above specification, without delay suspend the release of radioac-tive liquid effluents monitored by the affected channel, declare the channel inoperative, or change the setpoint so it is acceptably conservative. b. With less than one radioactive liquid effluent monitoring instrumen-tation dhannel in each release pathway OPERABLE, take the ACTION shown in Table 3.11.1-1. Return the instruments to OPERABLE status within 30 days or, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner, c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.8b are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE BRUNSWICK UNIT 1 3/4 1-1

l CHECK, CHANNEL CALIBRATION, AND CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.11.1-1. BASES: I The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid efflu-ents during actual or potential releases ofiliquid effluents. The alarm / trip setpoints for, these instruments shall be calculated and adjusted in accordance with the ODCM to ensure that the alarm / trip' will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of Gereral Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The purpose of tank ' level indicating devices is to assure the detection and control of leaks that, if not controlled, could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.. "Without delay" implies that the' operator, upon determining the LCO is being exceeded, takes the next appropriate action to comply with the specifi-cation. s 1 BRUNSWICK UNIT 1 3/4 1-2

TABLE 3.11.1-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Instrument i Applicability Action a. Liquid Radwaste Radioactivity Effluent Monitor (Providing alarm and automatic termination of release) 110 b. Liquid Radwaste Effluent Flow Measurement Device 111 c. Service Water Effluent Radioactivity Monitor 112 d. Stabilization Pond Effluent Composite Sampler 113 e. Stabilization Pond Effluent Flow Measuring Device 114 f. Outside Tank Level Indicating Devices--Units 1 and 2 CSTs 115 g. Service Wate{ Effluent from A0G I Precooler Radioactivity Monitor 112 l TABLE NOTATION I At all times During releases via this pathway. [This equipment is to be installed. Prior to installation, appropriate action statements 113 or 114 will be imple-mented.] ~ l

• • • At all times once this monitor is installed and af ter the A0G system becomes o,perational; however, if the AOC system becomes operational prior to the monitor being installed, then action statement 112 will be implemented.

(NOTE: This monitor is to be installed). BRUNSWICK UNIT 1 3/4 1-3 W- +1 m

TABLE 3.11.1-1 TABLE NOTATION (CONTINUED) ACTION 110 - With less than one channel 0PERABLE, effluent releases may continue provided that prior to initiating a release: a. At least two independent samples are analyzed in accor-dance with Specification 4.11.2.2, and b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise suspend release of radioactive effluents via this pathway. ACTION 111 With less than one channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump per-formance curves or tank level indicators may be used to esti-mate flow. ACTION 112 - With less than one channel OPERABLE, effluent releases may continue provided that, at least once per 12 hours, grab sam-ples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 i l microcuries per gram. l l ACTION 113 With the stabilization pond effluent composite sampler not OPERABLE, effluent releases may continue provided that, at least once per day, a grab sample is collected and analyzed for principle gamma emitters as per Table 4.11.2-1. O ther-wise, suspend releases via this pathway. l BRUNSWICK UNIT 1 3/4 1-4 l L

TABLE 3.11.1-1 TABLE NOTATION (continued) ACTION 114 - With the stabilization pond effluent flow measuring device not OPERABLE, effluent releases via this pathway may continue provided that flow is estimated at lea ~ t once per day during s ~ actual releases. The V-notch weir may be used. to estimate flow. ACTION 115 - With the tank liquid level device not OPERABLE, liquid addi-tions may continue provided the tank liquid level is estimated once per 8 hours during all liquid additions and deletions to and from the tank. I t l l l l l l l BRUNSWICK UNIT 1 3/4 1-5

TABLE 4.11.1-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Functional Instrument Check Check Calibration Test a. Liquid Radwaste Radioactivity Effluent Monitor D(1) M R(2) g(3) b. Liquid Radwaste Effluent Flow Measurement Device DIII (4) N.A. R Q c. Service Water Effluent Radioactivity Monitor D(1) M R(2) g(3) d. Stabilization Pond Effluent Composite Sampler D(1) N.A. R Q e. Stabilization Pond Effluent Flow Measuring Device D(1) N.A. R Q t I f. Outside Tank Level Indicating Devices-- D(5) N.A. R Q Units 1 and 2 CSTs g. Service Water Effluent From A0G Precooler Radioactivity Monitor D(1) M R(2) Q i BRUNSWICK UNIT 1 3/4 1-6

TABLE 4.11.1-1 TABLE NOTATION 1. During releases via this pathway. 2. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement ~ assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range for subsequent CHAN-NEL CALIBRATION; sources that have been related to the initial calibration shall be used. 3. The CHANNEL FUNCTIONAL TESTS shall also demonstrate that automatic isolation of this pathway,1f applicable, and control room alarm annunciation occurs if any of the following conditions exist: a. Instrument indicates measured levels above the alarm / trip setpoint. b. High-voltage low. I c. Instrument indicates a down scale feature. d. Instrument controls not set in " operate" mode. 4. The CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made. 5. During liquid additions to the tank. BRUNSWICK UNIT 1 3/4 1-7 1 l

3/4.11.2 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.2 The concentration of radioactive material released in liquid efflu-ents to UNRESTRICTED AREA (see Figure 5.1.3-1) after dilution in the discharge canal shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentra-tion shall be limited to 2 x 10-4 microcurie's/ml. APPLICABILITY: At all times ACTION: With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits. SURVEILLANCE REQUIREMENTS 4.11.2.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11.2-1. I 4.11.2.2 The re'sults of radioactivity analyses shall be used in accordance with the methods in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.2. BASES: This specification is provided to ensure that the concentration of radioactive materials released in liquid waste ef fluents to UNRESTRICTED AREAS after dilution in the discharge canal will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in BRUNSWICK UNIT 1 3/4 1-8

bodies of water in UNRESTRICTED AREAS will not result in exposures within (1) the Section II.A design objectives of Appendix 1,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its WC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP), Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the Lower Limits of Detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in WASL Procedures Manuals, HASL-300 (revised annually), Currie, L. A. " Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford l Company Report ARH-SA-215 (June 1975). "Without delay" implies that the operator, upon determining the LCD is being exceeded, takes the next appropriate action to comply with the specification. BRUNSWICK UNIT 1 3/4 1-9

=_ l 4 TABLE 4.11.2-1 BRUNSWICK RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Type.of Lower Limit of Liquid Release Type Sampling Analysis Activity Detection (LLD) Frequency Frequency Analysis-(uCi/ml) a, e b A. Sample Tanks' P P Principal 5 x 10-7,d Detergent Drain Each Batch Each Batch Gamma Emitters 9 Tank, and Salt 1-131 1 x 10-6 Water Release i Tanks (Batch Release)h d P Dissolved and 1 x 10-5 One Batch /M M Entrained Gases P M Gross Alpha 1 x 10-7 Each Batch Composite H-3 1 x 10-5 c 2 P Q Sr-89, Sr-90 5 x 10-8 Each Batch Compositec Fe-55 1 x 10-6 t b B. Stabilization P P Principal 5 x 10-7 Pond Each Release Each Release Gamma Emitters 9 D D During Periods During Periods of Releasef of Releasef l b C. Service Water W W Principal 5 x 10 7 During System During System Gamma Emitters 9 l Operation Operation - l b i D. Circulating P P Principal 5 x 10-7 Water Pit Each Release Each Release Gamma Emitters 9 I l BRUNSWICK UNIT 1 -3/4 1-10

TABLE 4.11.2-1 (continued) TABLE NOTATION a. The detectability limits for activity analysis are based on techni-cal feasibility limits and on the potential significance in the environment of the quantities released. For some nuclides, Lower Detection Limits may be readily achievable; and when nuclides are measured below the stated limits, they should also be reported. b. When operational or other limitations preclude specific gamma radio-nuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radio-active material released in the batch; and a weekly sample com-posited from proportional aliquots from each batch released during the week shall be analyzed for principal gamma-emitting radionu-clides. c. A composite sample is ore in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is repre-sentative of the liquids released. d. For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations. Under these circumstances, it will be more appro-priate to calculate the concentration of such nuclides using measured ratios with those radionuclides which are routinely identi-fied and measured. e. The Lower Limit of Detection (LLD) is determined according to the methodology in the ODCM. BRUNSWICK UNIT 1 3/4 1-11

TABLE 4.11.2-1 (continued) TABLE NOTATION f. The stabilization pond is typically released over a several-day period. The pond is to be sampled and analyzed prior to commencing release. When monitoring instrumentation becomes available and is - OPERABLE, daily sampling of the stabilization pond effluent will not be required during release. g. The principal gamma emitters for which the LLD specifications apply exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported. h. A batch release is the discharge of liquid wastes of a discrete v'ol ume. Prior to sampling for analyses, each batch shall be iso-lated and then thoroughly mixed to assure representative sampling. Once fully operational, the salt water tanks will be in'cluded as indicated in Table 4.11.2-1. BRUNSWICK UNIT 1 3/4 1-12

l 3/4.11.3 LIQUID EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.3 The dose or dose commitment to a MEMBER OF THE PUBLIC from radio-active materials in liquid effluents released to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited: a. During any calendar quarter to 13 mrem to the total body and to 110 mrem to any organ, and b. During any calendar year to 16 mrem to the total body and to 120 mrem to any organ. APPLICABILITY: At all times. ACTION: a. With the calculated doses from the release of radioactive materials in liquid effluents exceeding any of the limits in Specification 3.11.3, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s)'and defines the corrective actions that have been taken to reduce the releases and the proposed corrective action to be taken i 1 to assure that subsequent releases will be in compliance with the l above limits. I b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.3 Dose Calculations - Cumulative dose contributions from liquid efflu-ents shall be determined in accordance with the Off-site Dose Calculation Manual (0DCM) at least once per 31 days. BRUNSWICK UNIT 1 3/4 1-13

BASES This specification 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. The ACTION statements provide the required operating flexihility and at the same time implement the guides set forth in Section IV.A of Appendix I of 10 CFR Part 50 to assure that releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that con-formance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially under-estimated. 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, "Calcu-lation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revisi-on 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. The dose or dose commitment to a MEMBER OF THE PUBLIC is based on the 10 CFR Part 50, Appendix I, guideline of: a. 1.5 mrem to the total body and 5.0 mrem to any organ during any calendar quarter, and b. 3 mrem to the total body and 10 mrem to any organ during any calen-dar year, from radioactive material in liquid effluents from each reactor unit to UNRE-STRICTED AREAS. BRUNSWICK UNIT 1 3/4 1-14

3/4.11.4 LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.4 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site' to UNRESTRICTED AREAS (see Figure 5.1.3-1) would exceed 0.12 mrem to the total body or 0.4 mrem to any organ in a 31-day period. APPLICABILITY: At all times ACTION: a. With radioactive liquid waste being discharged without treatment and in excess of the limits in Specification 4.11.4, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following informa-tion: 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or s'ubsystem, and reason for the inoperability. 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and l 3. Summary of description of action (s) taken to prevent a recurrence. b. The provisions of 3.0) and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.4 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the ODCM. BRUNSWICK UNIT 1 3/4 1-15

BASES The requirement that appropriate portions of this system be used, when spect-fied, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10CFR Part 50.36a, General Design Criteria 60 of Appendix A to 10 'CFR Part 50 and the design object'ives given in Section II.D of A'ppendix 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 dose design ob'jectives set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid effluents. Mechanical filtration as per system design is considered to be an appropriate component of the liquid radwaste treatment system. The requirements of 0.12 mrem total body or 0.4 mrem to any organ in a 31-day period is based on 2 reactor units having a shared liquid radwaste treatment system. BRUNSWICK UNIT 1 3/4 1-16

.= 3/4.11.5 LIQUID HOLDUP TANKS Appropriate alternatives to the ACTIONS and Surveillance requirements below can be accepted if they provide reasonable assurance that in the event of an uncontrolled release of the tanks' content, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. LIMITING CONDITION FOR OPERATION 3.11.5 The quantity of radioactive material suspended in solution in each of the following unprotected outdoor tanks shall be limited to less than or equal to the activity indicated below, excluoing tritium and dissolved or entrained gases. OUTSIDE TANK CURIE LIMIT a. Condensate Storage Tank 1 10 Ci b. Condensate Storage Tank 2 10 Ci c. Outside Temporary Tank 10 Ci APPLICABILITY: At all times ACTION: a. With the quantity of radioactive material in any of the above listed tanks exceeding the limit of Specification 3.11.5, without delay suspend all addition of radioactive material to the tank; and within 48 hours reduce the tank's contents to within the limit. b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable. l l l BRUNSWICK UNIT 1 3/4 1-17

SURVEILLANCE REQUIREMENTS 4.11.5 The quantity of radioactive material contained in each of the tanks listed in Specification 3.11.5 shall be determined to be within the limit of Specification 3.11.5 by analyzing a representative sample of the tank's con-tents at least once per seven days when radioactive materials are being added to the tank. BASES The tanks listed in this specification include all those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system with the exception of the auxiliary surge tank. The auxiliary ' surge tank is excluded from this specification because there could be no uncontrolled release of the tank's contents (even in the event of a tank rupture) due to the storm drain collection system. The storm drains empty to a collection basin; the basin is sampled before being pumped to a stabilization pond, and the pond effluents to the intake canal are sampled and released activity is accounted for. Since the condensate storage tanks have continuous influent and effluent, stratification should not occur. Samples taken from the operating condensate transfer pump (s) vent shall be deemed representative of this system. t i 1 BRUNSWICK UNIT 1 3/4 1-18

3/4.12 GASEOUS EFFLUENTS 3/4.12.1 RADI0ACTI.E GASE0US EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.12.1.1 The radioactive gaseous effluent monitoring instrumentation chan-nels shown in Table 3.12.1-1 shall be OPERABLE with their alarm / trip setpoints set to e'nsure that the limits of Specification 3.12.2 are not exceeded. The setpoints shall be determined in accordance with the methodology as described in the ODCM. 3.12.1.2 The main condenser air ejector monitoring instrumentation channels shown in Table 3.12.1-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.12.6 are not exceeded. The setpoints shall be determined in accordance with the methodology as described in the OCDM. APPLICABILITY: As shown in Table 3.12.1-1 ACTION: a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of Specifications 3.12.1.1 or 3.12.1.2 are met, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative. l b. With less than one radioactive gaseous effluent or main condenser l air ejector monitoring instrumentation channel OPERABLE, take the l ACTION shown in Table 3.12.1-1. Return the instruments to OPERABLE status within 30 days or, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability l was not corrected in a timely manner. l l BRUNSWICK UNIT 1 3/4 1-19 l l

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.8b are not applicable. SURVEILLANCE REQUIREMENTS i 4.12.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST, and at the frequencies shown in Table 4.12.1-1. BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. The main condenser air ejector monitoring instrumentation channels, are pro-vided to monitor and control gross radioactivity removed from the main con-denser. The alarm / trip setpoint for this monitor shall be calculated in accordance with NRC approved methods to provide reasonable assurance that the potential total body accident dose will not exceed a fraction of the limits specified in 10 CFR Part 100. "Without delay" implies that the operator, upon determining the LC0 is being exceeded, takes the next appropriate action to comply with the specification. l BRUNSWICK UNIT 1 3/4 1-20 1

TABLE 3.12.1-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT APPLICABILITY ACTION 1. Main Stack Monitoring System ~ ~ 123 a. Noble Gas Activity Monitor 127 b. Iodine Sampler Cartridge 127 c. Particulate Sampler Filter d. System Effluent Flow Rate 122 Measurement Device e. Sampler Flow Rate 122 Measurement Device 2. Reactor Building Ventilation Monitoring System 123 a. Noble Gas Activity Monitor i 127 b. Iodine Sampler Cartridge 127 c. Particulate Sampler Filter ~ d. System Effluent Flow 122 Rate Measurement Device e. Sampler Flow Rate 122 Measurement Device BRUNSWICK UNIT 1 3/4 1-21

TABLE 3.12.1-1 (continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT APPLICABILITY ACTION 3. Turbine Building Ventilation Monitoring System a. Noble Gas Activity Monitor 123 127 b. Iodine Sampler Cartridge 127 c. Particulate Sampler Filter d. System Effluent Flow Rate Measurement Device 122 e. Sampler Flow Rate Measurement Device 122 4. Main Condenser Air Ejector Radioactivity Monitor (Prior to Input to Treatment System) I 121 a. Noble Gas Activity Monitor -~ Providing Alarm 5. Waste Gas Treatment (Downstream of A0G Treatment System) l a. Noble Gas Activity Monitor - 123 Providing Alarm l BRUNSWICK UNIT 1 3/4 1-22

TABLE 3.12.1-1 (continued) f RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT APPLICABILITY ACTION 4 6. Waste Gas Treatment System Explosive Gas Monitoring System 125 a. Hydrogen Monitor TABLE NOTATION At all times. During main condenser Augmented Off-Gas Treatment System (A0G) operation. During operation of the main condenser air ejector. 1 At all times once the Augmented Off-Gas Treatment System becomes operational. I r k i BRUNSWICK UNIT 1 3/4 1-23

Table 3.12.1-1 (continued) RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATION ACTION 121 - With less than one main condenser air ejector monitoring instr-umentation channel OPERABLE, gases from the main condenser off-gas system may be released to the environment for up to 72 hours provided: a. The augmented off-gas treatment system (once in operation) is not bypassed, and b. The main stack effluent noble gas activity monitor is OPERABLE; otherwise, be in at least HOT STANDBY within 12 hours. ACTION 122 - With less than one channel 0PERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 8 hours. ACTION 123 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided grab samples are taken at least i once per 24 hours and these samples are analyzed for gross noble gas activity within 24 hours. ACTION 125 - With less than one channel OPERABLE in any operating recombiner train, operation of this waste treatment system may continue provided grab samples from the affected train are collected at least once per 24 hours and analyzed within the following 4 hours and proper function of the recombiner is assured by moni?.oring recombiner temperature in accordance with approved procedures. BRUNSWICK UNIT 1 3/4 1-24

Table 3.12.1-1 (continued) TABLE NOTATION ACTION 127 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided samples are continuously col-1ected with auxiliary sampling equipment and analyzed as requi-red in Table 4.12.2-1. 1 1 1 BRUNSWICK UNIT 1 3/4 1-25

TABLE 4.12.1-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Functional Instrument Check Check Calibration Test 1. Main Stack Monitoring System a. Noble Gas Activity Monitor DIII M R(2) g(3) b. Iodine Sampler Cartridge WIII N.A. N.A. N.A. c. Particulate Sampler Filter WIII N.A. N.A. N.A. d. System Effluent Flow Rate Measurement Device Dill N.A. R Q e. Sampler Riow Rate i Measurement Device Dill N.A. R Q 2. Reactor Building Ventilation Monitoring System a. Noble Gas Activity Monitor DIII M R(2) g(3) b. Iodine Sampler Cartridge WIII N.A. N.A. N.A. l c. Particulate Sampler Filter WIII N.A. N.A. N.A. BRUNSWICK UNIT 1 3/4 1-26 ., _ ~., - -

TABLE 4.12.1-1 (continued) RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS P Channel Channel Source Channel Functional Instrument Check Check Calibration Test d. System Effluent Flow Rate Measurement Device D(l'l N.A. R Q e. Sampler Flow Rate Measurement Device D(1) N.A. R Q 3. Turbine Building Ventilation Monitoring System a. Noble Gas Activity Monitor D(1) M R(2) g(3) b. Iodine Sampler Cartridge W(1) N.A. N.A. N.A. I I Particul, ate Sampler c. Fil ter W(1) N.A. N.A. N.A. i d. System Effluent Flow l Rate Measurement Device D(1) N.A. R Q \\ t e. Sampler Flow Rate l Measurement Device D(1) N.A. R Q l i l f BRUNSWICK UNIT 1 3/4 1-27 l L

TABLE 4.12.1-1 (continued) RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 1 SURVEILLANCE RE0UIREMENTS Channel Channel Source Channel Functional Instrument Check Check Calibration Test 4. Main Condenser Air Ejector Radioactivity Monitor (Prior to Input to Treatment System) a. Noble Gas Activity Monitor - Providing Alarm 0(1) M R(2) g(3) e,. Waste Gas Treatment (Downstream of A0G Treatment System (5) a. Noble Gas Activity Monitor - Providing Alarm P P R(2) y 6. Waste. Gas Treatment System Explosive Gas Monitoring System (5) a. Hydrogen Monitor D N.A. Q(4) M h 6RUNSWICK UNIT 1 3/4 1-28

TABLE 4.12.1-1 (ctatinued) TABLENOTATIQ 1. During releases via this pathway. 2. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been o'btained from suppliers ~ that participate -in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. 3. The CHANNEL FUNCTIONAL TESTS shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist: a. Instrument indicates measured levels above the alarm / trip setpoint. b. High-voltage low. c. Instrument indicates a down scale failure. I d. Instrument not set in " operate" mode. 4. The CHANNEL CALIBRATION shall include the use of standard gas sam-ples containing a nominal: a. One volume percent hydrogen, balance nitrogen, and b. Four volume percent hydrogen, balance nitrogen. 5. Instrumentation for this system is only applicable once the Augu-mented Off-Gas Treatment System becomes fully operational at the Brunswick Steam Electric Plant. BRUNSWICK UNIT 1 3/4 1-29

e' ~ ,'/ 1 i 3/4.12.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.12.2 The-bose rate due to radioactive materials released in gaseous efflu-ents from.ttle site to areas at and beyond the SITE BOUNDARY (see Fig-ure 5.1.3-1) shall be, limited to the following: a. For noble gases: Less than or equej!.,to 500 mrems/yr to the total ' body and less than or equal to 3000 mre'ms/yr to the skin, and / b. For Iodi'ne-131, for tritium, and for all radionuclides in particu-a late form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any ' organ. i APPLICABILITY: At all ' times / ,1 i ACTION: With the dose rate (s) exceeding the ab'ove limits, without delay, restore the c release rate to within the above limit (s), and proiid.2 prompt notification to the Commission pursuant to Specification 6.9.1.7. .ti /- SURVEILLANCE REQUIREMENTS / i 4.12.2.1 The, dose rate due to noble gases ~in gaseous effluents shall be determined to be within the above limits in accordance with the methodology as 1 ~ described in the 00CM. ( 1 M / ,.f I f J ? BRUNSWICK UNIT 1 3/4 1-30 .) 'A . ~ - f

4.12.2.2 The dose rate due to Iodine-131, tritium, and all radionuclides in particulate fonn with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology as described in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.12.2-1. BASES This specification is provided to ensure that the dose rate at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose rate limits of 10 CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable assurance that radio-active material discharged in gaseous effluents will not result in the expo-sure of a MEMBER OF THE PUBLIC, either within or outside the SITE BOUNDARY, to annual-average concentrations exceeding the limits specified in Appendix B, Table II, of 10 CFR Part 20 [10 CFR Part 20.106 (b)]. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the total body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all ' times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. This specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are apportioned equally among the units sharing that system. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the Lower Limits of Detection (LL0s). BRUNSWICK UNIT 1 3/4 1-31

Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). "Without delay" implies that the operator, upon determining the LC0 is being exceeded,' takes the next appropriate action to comply with the specification. BRUNSWICK UNIT 1 3/4 1-32

TABLE 4.12.2-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Analysis Type of Detection (LLD)a Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml) A. Drywell Purge P P Principal Gamma 1 x 10-4 Each Purge Each Purge Emittersb Grab Samples B. Environmental Mc, d Mc Principal Gamma 1 x 10-4 Release Points - Grab Sample Emittersb Main Stack, H-3 1 x 10--6 i Unit 1 & Unit 2 f 9 Reactor Building Continuouse We Vents, Unit 1 & Charcoal I-131 1 x 10-12 Unit 2 Turbine Sample Building Vents, W.9 Principle Gamma f Hot Shop Continuouse Particulate Emitterb i x 10-11 Sample (I-131, others) Continuouse M Composite Gross Alpha 1 x 10-11 Particulate Sample Continuouse Q Composite Sr-89, Sr-90 1 x 10-11 Particulate Sample Continuouse Noble Gas Noble Gases Monitor Gross Beta or 1 x 10-6 Ganna i BRUNSWICK UNIT 1 3/4 1-33 .-_,,--,,,_._____-..--__,-,-___~___r

TABLE 4.12.2-1 (continued) RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION a. The Lower Limit of Detection (LLD) is determined in accordance with the methodology as presented in the ODCM. b. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133 m, Xe-135, Xe-135 m, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified -and reported. c. With a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour or following shutdown or start-up, sampling and analyses shall also be performed unless (1) analysis shows that the DOSE EQUIVA-LENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3. d. If during refueling, the tritium concentration in the fuel pool water l exceeds 2 x 10-4 pCi/ml, tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent-fuel pool area whenever spent fuel is in the spent-fuel pool. Fuel pool water will be sampled at least once per 7 days during refueling. e. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calcula-tion made in accordance with Specifications 3.12.2, 3.12.3, and 3.12.4. l BRUNSWICK UNIT 1 3/4 1-34 l

f. Samples cartridges / filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler). g. Sampling shall be performed at least once per 24 hours for at least 7 days following each shutdown, start-up, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in 1 hour, and analyses shall be com-pleted within 48 hours of changing. When samples collected for 24 hours are analyzed, the - corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that efflu-ent activity has not increased more than a factor of 3. l ~ BRUNSWICK UNIT 1 3/4 1-35 l l

3/4.12.3 GASEOUS EFFLUENTS DOSE - N0BLE GASES LIMITING CONDITION FOR OPERATION 3.12.3 The air dose due to noble gases released in gaseous effluents from the site to areas at and beyond the SITE B0UNDARY (see Figure 5.1.3-1) shall be limited to the following: a. During any calendar quarter, to j:_10 mrad for gamma radiation and j,20 mrad for beta radiation; b. During any calendar year, to < 20 mrad for gamma radiation and j,40 mrad for beta radiation. APPLICABILITY: At all times ACTIONS: a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding Limiting Condition For Operation 3.12.3, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases, and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.3 Dose Calculations - Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days. l BRUNSWICK UNIT 1 3/4 1-36

J BASES This specification is provided to implement the requirements of Sections II.B, 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.B of Appendix I. The ACTION statements provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I, to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through the appropriate pathways is unlikely to be substantially underesti-mated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous 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, Appen-dix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and dispersion of Gaseous Effluents in Rou-tine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY will be based upon the historical annual average atmospheric condi-tions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111. i l l l BRUNSWICK UNIT 1 3/4 1-37

3/4.12.4 GASE0US EFFLUENTS DOSE - 10 DINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.12.4 The dose to a MEMBER OF THE PUBLIC from Iodine-131, tritium, and all radionuclides in particulate form with hal f-lives greater than 8 days in gaseous effluents released from the site to areas at and - beyond the SITE BOUADARY (see Figure 5.1.3-1) shall be limited to the following: a. During any calendar quarter, less than.or equal to 15 mress to any organ; and b. During any calendar year, less than or equal to 30 mrems to any organ. APPLICABILITY: At all times ACTION: a. With the calculated dose from the release of Iodine-131, tritium, and radionuclides in particulate fonn with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken - to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. BRUNSWICK UNIT 1 3/4 1-38

SURVEILLANCE REQUIREMENTS 4.12.4 Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year for I-131, tritium, and radionu-clides in particulate form with half-lives greater than 8 days shall be deter-mined in accordance with the ODCM at least once per 31 days.

BgES, This specification is provided to implement the requirements of Section 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 ACTION statements provide 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 materials in gaseous effluents will -be kept "as low as is reasonably achievable." The 00CM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, "Calcu-lating of Annual Doses to Man from Routine Releases of Reactor Eff1'uents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specification for Iodine-131, tritium, and radioactive material in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man j in the areas at and beyond the SITE BOUNDARY. 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 l Vegetation with subsequent consumption by man, (3) deposition onto grassy l 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. i BRUNSWICK UNIT 1 3/4 1-39

3/4.12.5 GASEOUS RADWASTE TREATMENT /YENTILATION EXHAUST TREATMENT LIMITING CONDITION FOR OPERATION 3.12.5 THE GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials 'in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to effluent releases, from the site to areas at and beyond the SITE BOUNDARY ~ (see Figure 5.1.3-1), would exceed 0.4 mrad for gamma radiation and 0.8 mrad for beta radiation over 31 days. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from the site to areas at and beyond the SITE B0UNDARY (see Figure 5.1.3-1), would exceed 0.6 mrem to any organ over 31 days. l APPLICABILITY: At all times ACTION: a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within -30 days, pursuant to-Specification 6.9.2, a Special Report that includes the following inform & tion: l 1. Explanation of why gaseous radwaste was oeing discharged with-out treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability; 2. Action (s) taken to restore the inoperable equipment to OPERABLE status; and 3. Summary description of action (s) taken to prevent a recurrence. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. BRUNSWICK UNIT 1 3/4 1-40

SURVEILLANCE REQUIREMENTS 4.12.5 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the ODCM. BASES This requirement provides reasonable assurance that the releases of radio-active materials in gaseous effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. Until such time as the Augmented Off-Gas System becomes operational at the Brunswick Steam Electric Plant, the GASE0US RADWASTE TREATMENT SYSTEM shall refer to the 30-minute off-gas holdup line including stack filtration; and the only VENTILATION EXHAUST TREATMENT SYSTEMS shall be those installed for the Turbine Building Ventilation and the Standby Gas Treatment System. l l l ~ i i l BRUNSWICK UNIT 1 3/4 1-41 t

3/4.12.6 MAIN CONDENSER AIR EJECTOR RADI0 ACTIVITY RELEASE RATE LIMITING CONDITION FOR OPERATION 3.12.6 The gross radioactivity (beta and/or gamma) rate of noble gases measured at the main condenser air ejector shall be limited to ensure that the total body exposure to an individual in the UNRESTRICTED AREA shall not exceed 2.5 rem in one hour's time. 4 APPLICABILITY: At all times ACTION: With the gross radioactivity (beta and/or gamma) rate of noble gases at the main condenser air ejector exceeding the above limit, restore the gross radio-activity rate to within its limit within 72 hours or be in at least HOT STANDBY within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.12.6.1 The main condenser air ejector radioactivity release rate shall be determined in accordance with the methodology as described in the ODCM. 4.12.6.2 The gross radioactivity (beta and/or gamma) rate of noble gases from the main condenser air ejector shall be determined to be within the above limit at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the main condenser air ejector: a. At least once per 31 days. b. Within 72 hours following an increase, as indicated by the Condenser Air Ejector Noble Gas Activity Monitor, or greater than 50%, after factoring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release from the primary cool-ant. BRUNSWICK UNIT 1 3/4 1-42

BASES Restricting the gross radioactivity rate of noble gases from the-main con-denser provides reasonable assurance that the total body exposure to an indi-vidual in the UNRESTRICTED AREA will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. - i e BRUNSWICK UNIT 1 3/4 1-43

3/4.12.7 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.12.7 The concentration of hydrogen in the waste gas treatment system shall be limited to less than or equal to 4% by volume. APPLICABILITY: At all times (see Bases) ACTION: a. With the concentration of hydrogen in the waste gas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.7 The concentration of hydrogen in the waste gas treatment system shall be determined to be within the above limit by continuously monitoring the waste gases in the waste gas treatment system with the hydrogen monitors required OPERABLE by Table 3.12.1-1 of Specification 3.12.1.1. BASES This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is main-- l tained below the flammability limits of hydrogen. Maintaining the concentra-l tion of hydrogen below the flammability limits provides assurance that the I 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. This specification will become applicable when the Augmented Off-Gas System l becomes fully operational at the Brunswick Steam Electric Plant. There is no requirement for hydrogen monitors on the 30-minute waste gas holdup line which will serve in the interim. BRUNSWICK UNIT 1 3/4 1-44

3/4.12.8 DRYWELL PURGES (MARK I CONTAINMENT) LIMITING CONDITION FOR OPERATION 3.12.8 The drywell shall be purged through the Standby Gas Treatment System or released to the environment at a rate in conformance with Specifica-tion 3.12.2. APPLICABILITY: Whenever the drywell is vented or purged. ACTION: a. With the requirements of the above specification not satisfied, suspend all VENTING or PURGING of the drywell. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.8 A sample analysis, as defined in Table 4.12.2-1, shall be performed prior to each drywell PURGE. BASES This specificatiod provides reasonable assurance that releases from drywell PURGING operations will not exceed the annual dose limits of 10 CFR Part 20 for UNRESTRICTED AREAS. r BRUNSWICK UNIT 1 3/4 1-45

3/4.13 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.13 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements. APPLICABILITY: At all times ACTION: a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site. b. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.8b are not applicable. SURVEILLANCE REQUIREMENTS 4.13 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottomd, and sodium sulfate solutions). l a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-l TION of the batch under test shall be suspended until such time as l additional test specimens can be obtained, alternative SOLIDIFICA-TION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alterna-tive SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM. BRUNSWICK UNIT 1 3/4 1-46

b. If the initial test sp:cimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection of testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATI0h'. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.15, to assure SOLIDIFICATION of subsequent batches of waste. BASES This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constit-uents, mixing, and curing times. I t BRUNSWICK UNIT 1 3/4 1-47

3/4.14 40 CFR PART 190 3/4.14.1 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.14.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrems). APPLICABILITY: At all times ACTION: a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Speci-fications 3.11.3a, 3.11.3b, 3.12.3a, 3.12.3b, 3.12.4a, or 3.12.4b, calculations should be made to determine whether the above limits of Specification 3.14.1 have been exceeded. If such is the case, in lieu of a Licensee Event Report, prepare and submit to the Commis-sion within 30 days pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above l limits and includes the schedule for achieving conformance with the l above limits. This Special Report, as defined in 10 CFR Part l 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for j the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentra-tions of radioactive material involved and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the ~ bove limits; and if the release condition resulting in violation of a 40 CFR Part 190 has not already been corrected, the Special Report I BRUNSWICK UNIT 1 3/4 1-48 l l

f shall include a request for a variance in accordance with the pro-visions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a veelance is granted until staff action on the request is complete. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUREMENTS 4.14.1 Dose Calculations Cumulative dose contributions from liquid and gas-eous effluents shall be determined in accordance with Specifications 4.11.3, 4.12.3, and 4.12.4, and in accordance with the ODCM. Only uranium fuel cycle sources within five miles of the Brunswick Steam Electric Plant will be con-sidered. BASES This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reac-tors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course:of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, 'with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected) in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c is considered to be a timely request and ful-BRUNSWICK UNIT 1 3/4 1-49

= fills the requirements of 40 CFR Part 190 until NRC staff action is com-pleted. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3/4.11 and 3/4.12. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. I 1 i I BRUNSWICK UNIT 1 3/4 1-50 i

3/4.15 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.15.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.15.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.15.11-1. APPLICABILITY: At all times ACTION: a. With the radiological environmental monitoring program not being conducted as specified in Table 3.15.1-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specifi-cation 6.9.1.3, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. b. With the level of radioactivity as the result of plant effluents in an environmental samplimg medium at a specific location exceeding the reporting levels of Table 3.15.1-2 when averaged over any calen-dar quarter, in lieu of a Licensee Event Report, prepare and submit to the ommission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective action to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Specifica-tions 3.11.3, 3.12.2, and 3.12.4. When more than one of the radio-nuclides in Table 3.15.1-2 are detected in the sampling medium, this report shall be submitted if: concentration (1) concentration (2) + + .. 3,1. 0 reporting level (1) reporting level (2) BRUNSWICK UNIT 1 3/4 1-51

When radionuclides other than those in Table 6.9-1 are d tected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.3, 3.12.3, and 3.12.4. This report is not required if the measured level of radioactivity was not the result of plant effluents; how-

ever, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.

c. With milk or fresh leafy vegetables unavailable from one or more of the sample locations required by Table 3.15.1-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program and ODCM. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.9, identify the cause of unavailability of samples; and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report, and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s). d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS I 4.15.1 The radiplogical environmental monitoring samples shall be collected i pursuant to Table 3.15.1-1 from the specific locations given in the table and figure (s) in the ODCM and shall be analyzed pursuant to the requirements of i Table 3.15.1-1 and the detection capabilities required by Table 4.15.1-1. i BRUNSWICK UNIT 1 3/4 1-52

BASES The radiological environmental monitoring program required by this specifica-tion provides nieasurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest poten-tial radiation exposures of MEMBERS OF THE PUBLIC resul ting from station operation. This monitoring program implements Section IV.B.2 of Appendix 1 to 10 CFR Part 50 and thereby supplements the radiological effluents monitoring program _by verifying that the measurable concentrations of radioactive materials are not higher than expected on the basis of effluent measurements and the modeling of the environmental exposure pathways. The required detection capabilities for environmental sample analysis are tabulated in terms of the Lower Limits of Detection (LLDs). The LLDs required by Table 4.15.1-1 are considered optimum for routine environmental measure-ments in industrial laboratories. It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a, posteriori (after the fact) limit for a par-ticular measurement. Detailed discussion of the LLD and other detection limits can be found in HASL Procedure Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determinator Application to Radio-chemsitry" Anal. Chem 40, 586-93 (1968), and Hartwell, L. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). Groundwater is not monitored by this specification because plant liquid efflu-l ents are not tapped as a source for drinking or irrigation purposes. i The following notes provide the specific bases for the Radiological Monitoring Program summarized in Table 3.15.1-1. 1. Direct Radiation - At least 40 routine monitoring stations with two or l more dosimeters or one instrument for measuring and recording dose rate continuously are placed as described in the ODCM. The stations are l BRUNSWICK UNIT 1 3/4 1-53 l l

located to provide an inner ring of stations in each sector located as near the SITE B0UNDARY as is reasonably accessible and practical so as to estimate the highest possible dose to a MEMBERS (s) 0F THE PUBLIC. An l additional outer ring of stations is also included at distances of 8 km or greater, selected in each sector on the basis of accessibility. Sectors extending over open water are omitted. Additional stations are added to represent nearby populated areas such as a religious assembly area, recreational area, and permanent residences. An additional 3 stations are maintained at distances greater than 16 km to serve as control stations. 2. Airborne - 3 samples are taken at stations located as near to the SITE B0UNDARY as is reasonably accessible. In addition, one sampling station f is established in a nearby community - and one is located to provide con-trol data. 3. Waterborne - The surface water at BSEP is brackish, estuary water that is 9 discharged into the Atlantic Ocean. This water is not used as water for any kind of drinking or other human consumption nor is it used to irri-gate consumable foods but only for recreational activities; therefore, one upstream and one downstream sampling location is established to estimate and monitor plant effect via this pathway. 4. Ingestion I a. Milk , Samples from milking animals within a 5-km distance are taken. ' Currently only one such station of a single animal exists. Additions and deletions of stations are accomplished in accordance with Specification 3.15.2. One station of greater than 16 km from the plant is in a low D/Q sector is collected to provide a control data for comparison purposes. b. Fish and Invertebrates - One sample is taken representative of consumable species in the vicinity of the discharge point. One sample of similarly consumable species are located in areas upstream or otherwise not influenced by plant discharges. BRUNSWICK UNIT 1 3/4 1-54

c. Food Products - Samples of broadleaf, edible vegetation grown in two (2) different sectors of historically higher D/Q values at the SITE B0UNDARY are collected as seasonally available as per Table 3.15.1-1. One additional sample of similar edible vegetation grown at a distance of about 10 miles from the plant is also taken to provide control data when it is seasonally available as per Table 3.15.1-1. i BRUNSWICK UNIT 1 3/4 1-55 k - -

TABLE 3.15.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF SAMPLES SAMPLING AND TYPE AND FREQUENCY AND/0R PATHWAY AND COLLECTION FREQUENCY OF ANALYSIS SAMPLE LOCATIONSa 1. Direct Radiation Locations 1-40. At each Q Gamma Dose - Q location with 2 or more dosimeters o'r one or more instruments for continuously measuring and recording dose rate. 2. Airborne - Locations 41-45 Continuous sampler opera-Radiofodine Cannister - W Radiotodine and tion with sample collec-Particulate sam 31er - Ana-Particulate ' tion weekly or as re-lyze for gross beta radio-quired by dust loading, activity > 24 hours fol-whichever is more fre-lowing Tilter change. quent. Perform gamma isotopic analysis on each sample when gross beta activity is > 10 times the yearly mean of control samples. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 days. b 3. Waterborne Locations 46-47 Composite sar.ple col-Gamma Isopic Analysis - M a. Surface lection - M Tritium - Q Analysis SA Gamma Isotopic SA

b. Sediment from Location 48 shoreline

TABLE 3.15.1-1 (continued) NUMBER OF SAMPLES SAMPLING AND TYPE AND FREQUENCY EXPOSURE PATHWAY AND COLLECTION FREQUENCY OF ANALYSIS AND/OR SAMPLE SAMPLE LOCATIONSa 4. Ingestion Locations 49, 50 With animals on pasture-SM Gamma isotopic (animals a. Milkc At other times - M on pasture) SM I-131 analyses (animals on pasture) SM Gamma Isotopic (other times) M I-131 analysis (other times) M

b. Fish and Postions 51, 52 When in season - SA Gammt isotopic on edible Invertebrates portions - SA

c. Food Positions 53-56 When available - M Gamma isotopic - M c

Products I-131 -M

'/ - t, / / s-g s / TABLE 3.15.1-1 (continued) TABLE NOTATION s-f l s. Actual locations,(distance and direction) from the site are provided for a. all sample locations. fn a table and a ' figure (s) *in the ODCM. i b. Compo,ite samples shall be collected with equipment that is capable of coll.ecting an aliquot at time intervals that are short (e.g., once per 6 hodrs) relative to the compositing period (e.g., monthly) 'in order to assure obtaining a represen'tative simple. c. When less than three (3) milking animals are available for testing within an 8-km ' distance and when the dose to a MEMBER OF THE PUBLIC at 3 km in the highest X/Q sector, ' calculated in accordance with the ODCM for a milk pathway is greater than 1 mrem per' year, sa.mpling of food products shall be performed as indicated in Table 3.15.1-1, 4.c, in lieu of milk sam-pling. / ,/ / / 6 f .A J f yi p a 2 l l Y 4 BRUNSWICK UNIT 1 3/4 i-58 ,} r

TABLE 3.15.1-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES 4 Reporting Levels Water Airborne Particulate Fish Milk Food Products 3 Analysis (pCi/1) or Gases (pCi/m ) (pCi/kg, wet) (pC1/1) (pC1/kg, wet) 3 4 Mn-54 1 x 10 3 x 10 2 4 Fe-59 4 x 10 1 x 10 3 4 Co-58 1 x 10 3 x 10 2 4 Co-60 3 x 10 1 x 10 2 4 Zn-65 3 x 10 2 x 10 Zr-Nb-95 4 x 102 2 I-131 2 0.9 3 1 x 10 ~ 3 3 Cs-134 30 10 1 x 10 60 1 x 10 3 3 Cs-137 50 20 2 x 10 70 2 x 10 2 Ba-La-140 2 x 10 3 x 102

TABLE 4.15.1-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD) Water Airborne Particulate Fish Milk Food Products Sediment 3 Analysis (pCi/1) or Gases (pC1/m ) (pCi/kg, wet) (pC1/1) (pC1/kg, wet.) (pCi/kg, dry) gross beta 4 0.01 H 2000 3 130 Mn-54 15 Fe-59 30 260 Co-58, 60 15 130 2n-65 30 260 Zr-Nb-95 15 1-131 1* 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 15 Ba-La-140 15 N::te: This list does not mean that these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported. The LLD is determined according to methodology in the ODCM. LLD for drinking water

3/4.15.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.15.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest resident, and the nearest 2 (500 ft ) producing broadleaf vegetation. (For 2 garden of greater than 50 m elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the location in each of the 16 meteorological sectors of all milk animals and all gardens of greater than 50 m2 producing broadleaf vegetation.) Broadleaf vegetable sampling of at least 3 different kinds of vegetation may be performed at the site boundary in each of 2 different direction sectors with the highest D/Qs in lieu of the garden census. Specifications for broad-leaf vegetation sampling in Table 3.15.1-1(4c) shall be followed, including analysis of control samples. APPLICABILITY: At all times ACTION: a. With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.12.4, in lieu of a Licensee Event Report, identify the new ' location (s) in the next Semiannual Radioactive Effluent Release

Report, pursuant to Specifica-tion 6.9.1.9.

b. With a land use census identi fying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are cur-rently being obtained in accordance with Specification 3.15.1, add the new location (s) to the radiological environmental monitoring program within 30 days. The sampling location (s), excluding the BRUNSWICK UNIT 1 3/4 1-61

central station location, having the low 2st calculated dose or dose commitment (s) (via this same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.9, identify the new location (s) in the next Semiannual Effluent Release Report; and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s). c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.15.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best resul ts, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The result of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. BASES This specification is provided to ensure that changes in the use of area at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of the census. The best information from door-to-door surveys, aerial surveys, or consulting with local agricultural authorities shall be used. l This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine the minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broadleaf vegetation (i.e., similar to lettuce and cabbage; and (2) a vegetation yield of 2 kg/m2, l BRUNSWICK UNIT 1 3/4 1-62 f ~

3/4.15.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.15.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commis-sion. APPLICABILITY: At all times ACTION: a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 8.5.3 A summary of the results, obtained as part of the above, required Interlaboratory Comparison Program and in accordance with the ODCM, shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. BASES The requirement for participation in the Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitor-ing in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. BRUNSWICK UNIT 1 3/4 1-63

a ,.n,--- wa-4"-- O b a SECTION 5.0 DESIGN FEATURES c r 1 f t 1 ~ " " ' '

• wwwr.w.

5.0 DESIGN FEATURES 5.1 SITE EXdLUSIONAREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1. SITE BOUNDARY 5.1.3 The SITE BOUNDARY shall be as shown in Figure 5.1.3-1. For the purpose of effluent release calculations, the boundary for atmospheric releases is the SITE BOUNDARY and the boundary for liquid releases is the SITE BOUNDARY prior to dilution in the Atlantic Ocean. 3 i I 5-1 l

/ ! / i "y P / R 6/i/g ._N s EV i l / ~ Is )[ n A, / 's a j] e x ~ p ts = r lg a f b' c e wg L E AKLA \\ RA N T n / uA C NA on EI C LA / /M y N A C'W E/ KR A T IN / ~ / I E a N c I I L y l N ort O oe is w o t o L T c. y E L I N q OD p I A c D I TO A CR O U R RS TS S SE S T NC E N OC C A CA C LP A N L I AM [ 4 o s ~~~ I A D 2 E E L M mA \\ L O R C A OR o, S N TE C A I T ~ C SN I I E H E VC oP fR A G R FG A l i , y ~' C o i l S l I D h 6 ~~ ~ c'. pS ? > o a2 ?c w

(@ ,,J,~%.3 - d h[to.l u G, i \\ i

O f

. g i, A a ** ~ EJ J h Q r, !:*e 'l ~ ~ I s.' / /

~

bl 'e+ ~ .. f* ',i. n...g f., s a sr \\ ... e s 1, / '%p \\ lJT ~ ./ l ,a- %...a & .sn ~ ^ W f s. ) / ~',- I ~~ ' ; $ f " " *' a. Y y/{ d' s / b i .J e / / / g j'

1. 5

.,,3; s h / k .)'/ 'h!'

• f l ',

&y '.!%h h.~ff'D. ju l u +- .s - . ND j j y "Ni q p%* q',s. f p,h J A (L f 3- ~ s' - 's-, g3 Iy,1 =~ . c#- ) s, p > M-h %.,- h j. c' u A: ~ W p1j -, ~ _ - - l't s N y s~ l L. ~ LOW POPULATION ZONE FIGURE 5.1.2-1 1 .=- = BRUNSWICX 5-3

t i 1 1 i t i i i i e:i i, i.;. 1 1 . \\ r j a a 2 3 3 3 3 3 2 2 s r ,L -T.__t f {' 4 y x v / tN. ,\\i L \\\\, s r' / ' $g 6~ e % my w l. .I t t 7

g. / g -

)\\ i r{. c _, ( e s g i s\\ / e s..,, 3 s 1 l 'c' \\" l- .... - =v g^ ' fpf.siue.h. p M D . -5..I.].._ '/ tg'/, /[.- r\\ 1- -- 4y.. -; A. (;,.,9,y 1. bh 4-- ff '()N. __ I Y / / ',' l ~ j o tg 4.,<,,. v-i %'$..A j,,s N 9- ,) l D ll l [ \\ h_.o Dl\\_.j.dJT% T 3 I / / { f' s I .t j -{4 / ~ g / i 'D f l } e\\-% .li., l .d .. \\ >e. ! N /I '. ),/ . ( i N /.( A( T \\ 2. \\Y /d. f l ,.s %.s ., / / . /- j-,.-- v, .(q! i. \\'"", f. N.sg.- ) \\ \\ .,r p'.. ^ l t / m '1 / [j,eL. _ _ Q }iP l ~ i / SITE BOUNDARY [ k (h i ~' Figure 5.1.3-1 l , sp-j#Q - N. CF&. FROFERTY UNE i a.*, %,, g.] f 1\\ e'I C /'/ n/ m %'

O O e SECTION 6.0 ADMINISTRATIVE CONTROLS i \\ I l f

6.0 ADMINISTRATIVE Cbt iROLS c== ' 6.1. RESPONSIBILITY 6.1.1 The General Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsi-bility during his absence. 6.2 ORGANIZATION OFF' SITE 6.2.1 The offsite organization for facility management and technical ~ ~ support shall be as shown on Figure 6.2.1-1. ~ FACILITY STAFF 6.2.2 The. Faci.lity organization shall be, as shown on Figures 6.2.2-1 and 6.2.2-2 and:. a. Each on duty shift shall be. composed of at least the minimum shift crew composition shown in Table 6.2.2-l'. f~ b. At least one licensed Operator shall be in the centrol room for each reactor containing fuel. c. At least two licensed Operators shall be present in the control room for each reactor in the process of start-up, scheduled reactor, shutdown and during recovery from reactor trips. d. An individual qualified to implement radiation protection procedures shall be on site when fuel is in either reactor, e. All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or. Senior Reactor Operator Limited to Fuel Handling who has no other concurrent respon-sibilities during. this operation. f. A Fire Brigade of at least five members shall be maintained onsite at all times. The Fire Brigade.shall not include the. minimum shift crew shown in Table 6.2.2-1 or any personnel required for other essential functions during a fire emergency. .m b 6-1 6-G W -6M .,g re- -- --

s = ~~ [ Cxilaxu CF TP.E 3:12.3 FF.01D CGlcal EF DC*J*lTC OFFIU2 e ePLAAilcal 3*'3*f D'C }I8' P"'* talet tuars*.ing effleer j 8 l t I 8 rs=-:.1 surru gyg S' WSTDKER 3DYlCD I taecutive fics tresieent 1-8, 4 g 4 4 3I 7M1 Surr'.T 8 3. Sanier Tics r,realeest t e ',3 *e e t' CC.Ilt SATUT l-W*U.11 CPE11T10E5 as str u w sT aut=

u. s.

Vice trusiseert s i e. estnem re..zu umf 4:gtons . w e.t mm u=nua: aux a x. w,, m.w c s r- \\, j s rnus General J w er ri=s n.cs.se usetr e ino. c "Rasponsible for perfomance and r:enitoring of Fire Protection Program. - MANAGEMENT ORGANIZATION CHART FIGURE 6.2-1 .= .~=. ~=-- 6-2 e .99..e. S g g .-w.- r ---n +e

Ilill ) illy I i i i (:111F. Ital. IIAllAttF.R palin:sull:st rIAllt E I I I I luhACf 4 ItAWA013 tianar.8 8 fitfC C4 Cli s AIIDig itAlpitaAnC[ Iffnustat Ane tutlian iss(IV AlsilBl3tAAllt( Amp QUAL. Ao304 ANCE i I I I I 1 I pflagge pfp.. testpf f alat 3F(CIAll38 SP(C18till l Sf(ClatI33 !Niff Ortsallag er(sallen) Atsstal31salll( an a Surtavl304 38thalls ilat fatittl100 Sicutalent Se,rtas Seal Emulp(183 Surlatt104 l e-I i 1 fl Ar.)t l lear s f 88Jt CI str(RIl304 18888 8 ftAuf 4411s s leatid a thelallas* 30rf 4fl30sl cutt lIr IOrlin e plateff3 $st 808tHA4 Silfitfl304 IRAlaine . A13tman[( ' 1 d,3f Co,nl AOL s.f.t C.i.n.t.l 3 8 .. 1 91, 3 g .,.ingos 3 mi <l31 I 7 .....t

e. l.x.

s.,,m .... 0L esistions po I 1 I I i If SflCIAll3I $ffCIAtl3I 3fttintt3t3 f 8488taalf atAlsltal,(( ltalplisanCI AIJ RA hCalist 14838C3 14 Alalat Apaallaar OrisAlp4 4 30flRVl10A 5tsrthfl304 l l_ I I i 11[,'(1 3IIj^l- ,IN!:'fl!" tNu NIli:Na'f ,,,hfi,h, i,, ,y}{} ljs,t ,gt,j, jjig l Ininrea: s- ~ ~

1. ~

• 340 - 31 I lACl04 Ofl8AIPB llCINIC se-88 ttl04 criaA104 LICth3C FACILITY OllGAlli2ATI0tl.

Figure G.2.2-1 O

b PLANT - GENE?.AL 5__ HL4ASER DIRECTOR Kuctcar Szie".1 m quality A:surr.nce s l SPEClAllSTS " .Sy?ERY!SOR SJPERY1502 _ra..ining i F. ire Protection q_-zlity 4 Issu.zoce r ~~' ' FIRE PREYSTION FIEE P10TECT10R CCMH1incE SUPP0ZY G?.0UP i j l gg!GADE FIRE CHIEF' ~ (EMERGENCY COORDINATOR) SHIFT FIRE s s I SRIGADE ~ 1. .e l PLANT FIP.E FF.CTECTIO.N C.e.G/.JilZAT10N ~ ~ _=

• Nu:::ber of Brigade Fire Chiefs varies with shift organization.

- **0ne Engineer is assigned the duties of the plant fire chief. l FIGURE 6.2.2-2 6-4 e__. _.., _ - _, _ _ _ _

T_ABLE 6.2.2-1 ~ HINIMUM SHIFT CREW CCMPOSITIONf ~ Condition of Unit 1 - Unit 2 in CONDITION 1, 2 or 3 i.ICENSE APPLICABLE OPERATIOtAL CONDITIONS CATEGORY ~ 1.2,3 4&5 SQL** 2 l 2* ~ 2 OL** 3 Kon-Licsnsed 4 l 3 i C55fition of Unit 1 - Unit 2 ifi CONDITION 4 cr 5 1 LICENSE APPLICABLE i CATEGORY OPERATIONAL CONDITIONS ~ 1, 2. 3 l4&5 ~-' c-SQL** 2 1* OL** l 2 l 2 Non-Licensed 3 3 l t'- i Condition of Unit 1 - No Fuel in Unit 2 - L y i LICENSE APPLICABLE CATEEDRY OPERATIOMAL CONDITIONS 1 1.2.3 l 4&5 1 1* i l SOL OL l 2 1 i 1 Non-Licensed l 2 h 1 r I Oces not include the licensed Senior React:r Operater. or Senior React:r { Operator Limited to Fuel Handling, supervising CDRE Alic.dTIONS.

• • Assu=es each individual is licensed on both plants.

f Shift crew c mposition, including an individual qualified in radiation r protection procedures, may be less than the minimum requirments l fer a peried of time not to exceed 2 hcurs in order to ace medate unexpected absence of on duty shift crew members provided imediate . acticn is taken to restore.the shift crew cc:pesitien to within the j minimum requirements of Table 6.2.2-1. l 7 ) 6-5.* __._.m._

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each membir of the facility staff shall meet or exceed the mini-mum qualifications of ANSI N18.1-1971 for ccmparable position, except for the Environmental and Radiation Control Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. i-6.4 TRAINING .5.4.1. A.metraining and replacement training program for the facility staff shall be maintained un' ef the'tiirectibn of'the Training Supervisor d and shall meet or exceed the requirements and reccmnendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Fire Chief and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975.

6. 5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY C0K41TTEE (PNSC)

FUNCTION 6.5.1.1 The PNSC shall function to advise the General Manager on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PNSC shall. be tcmposed of the: e, l Chainnan: Plant General Manager

1. ,,;,Vice Chainnan:

Operations Manager, Maintenance Manager, a l < Technical .. Administrative Manager:or l Director-Nuclear Safety and QA Secretary: Administrative Supervisor Member: Maintenance Supervisor (I&C) Member: Maintenance Sugervisor (Mechanical) Member: Engineering Superviser ~ Member: Environmental and Radiation Control, Supervisor i l Member: Quality Assurance Supervisor Member: Shif t Operating Supervisors Member: Training Supervisor ALTERNATES ~ l 5.5.1.3 All alternate members shall be appointed in writing by the PNSC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PNSC activities at any pne mme. 6-6 l

ADMINISTRATIVE Col'TROLS' [= MEETING FRE00ENCY 6.5.1.4 The FHSC shall meet at least once per calendar month and as convened by the FNSC Chainnan or his designated alternate. QUORUM 6.5.1.5 A quorum of the FNSC shall consist of the Chaiman or Vice Chaiman and three members including alternates. RESPONSIBILITIES 6.5.1.6 Tne FNSC shall be responsible for: Review.of 1) all precedures required by Specification 6.8 and a. changes theretb, 2) any other preposed procedures or changes thereto as determined by the General Manager to affect nuclear safety. b. Review of all proposed tests and experiments that affect nuclear safety. c. Review of all proposed changes to Technical Specifications. d. Review of all proposed changes or modifications to plant e systems or equipment that affect nuclear safety. Investigation of all violations of the Technical Specifications e. including the preparation and forwarding of reports covering evaluation and recom.endations to prevent recurrence to the Vice President - Nuclear Operations and to the Manager - Corporate Nuclear Safety and Quality Assurance Audit. f. Review of all events raquiring 24 hour notification to the Cccai ssion. g.. Review of facility operations to detect potential safety. hazards. h. Performance of special reviews, investigations and reports thereon as requested by the Manager - Corporate Nuclear Safety and Quality Assurance Audit. i. Review of the Plant Security Plan and implementing procedures. j. Review of the Emergency Plan and i=plementing precedures. 6-7

~ =- = --

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-Nuclear Operations' and to the Corporate Nuclear Safety Unit.

l. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.

e S 4 e G e 6 m a, e e e O _n. ameneae ~6-8 1l _,

-~ 8 . ADMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 ' The PNSC shall: Recccmend to the General Manager written approval or disapproval a. of items considered under 6.5.1.6(a).through (d) above. b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question. Provide written notification within 24 hours to the Vice c. President - Huclear Operations and the Manager - Corporate Nuclear Safety and Quality Assurance Audit of disagreement between the PHSC and the General Manager; however, the General Manager shall have resp 9nsibility for resolution gf such disagreements pursuant to 6.1.1 above. RECORDS _ 6.5.1.8 The PHSC shall maintain written minutes of each meeting that. at a minimum, document the results of all PHSC activities performed under the responsibility and authority provisions of these technical specificatiens, and copies shall be provided to the Vice President - Huclear Operations and to the Manager - Corporate Nuclear Safety and ~ Quality Assurance Audit. 6.5.2 CORPORATE NUCLEAR SArc6 f AND OUALITY ASSURANCE AUDIT SECTION ~ RESPONSIBILITY i 6.5.2.1 The Manager - Corporate Huclear Safety and-Quality Assurance l Audit, under the Vice President - Nuclear Safety and Research, is charged with the overall responsibility for administering the independent off-site review and quality assurance audit programs as follows: Approves safection of the individuals to conduct off-site a. safety reviews and quality assurance audits.x l Has access to the plant operating'recdrds' and operating personnel b. in performing the independent reviews and quality assurance a'udit: a Prepares and retains written records of, review and audits. ~, c. l d. Assures independent safety reviews are conducted on all items required by Section 6.5.3.3 and quality assurance audits cover all items included in Section 6.5.4.1. e ' Distributes reports, records of PNSC meetings, and other records to the appropriate managers and individuals assigned to conduct __5 the off-site safety reviews and quality assurance audits. 6-9

ADMINISTRATIVE C0f.TROLS = 6.5.3 CORPORATE !!UCLEAR SAFETY UNIT' (CNSU)_ FUNCTION 6.5.3.1 The Corporate Nuclear Safety Unit of the Corporate Nuclear Safety and Quality Assurance Audit Section shall provide independent off-site review of significant plant changes, tests, and procedures; verify that reportable occurrences are pr:mptly investigated and corrected in 4 manner which reduces the probability of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer. g PERSONNEL _ .~ 6.5.3.2 i Personnel assigned responsibtTity for independent reviews a. shall be specified in technical disciplines, and shall collec-tively have the experience and competence required to review problems in the following areas: 1. ' Nuclear power plant operations

2. ' Nuclear engineering 3.

Chemistry and radiochemistry ( 4. Metal'lurgy 5. Instrumentation and control 6. Radiological safety 7. Mechanica1 and electrical engineering 8. Administrative controls 9. Seismic and environmental 10.! Quality assurance practices b. Tlie following minimum experience requirements shall be established for those persons involved in the independent off-site safety review program: 1. Manager'of CNS and QAAS - Bachelor of Science in engineering or related fi. eld and ten (10) years related experience ' including five (5) years involvement with operatioh and/or design of nuclear power plants. .2. Reviewers - Bachelor of Science in enginee. ring or related field or equivalent and five (5) years related experience. including three (3) years involvement with operation and/or design of nuclear power plants. = _. 6-10 ~

=:

2= ' ADMINISTRATIVE C0f(TROLS PERSONNEL (Continued}. An individual may possess competence in more than one specialty c. If sufficient expertise is not available within the area. Corporate Nuclear Safety Unit, ccmpetent individuals from other Carolina Power and Light Ccmpany organizaticns or outside con-sultants shall be utilized in perfonning independent off-site reviews and investigations. d. At least three persons, qualified as discussed in Specification 6.5.2.3.b,. sha11 review each item submitted under the require-ments of Section 6.5.3.3. Independent safety reviews shall be perfermed by perscnn'al not e. directly involved with the activity or responsible for the activi.ty. = SUBJECTS REOUTRING INDEPENDENT REVIEW 5.5.3.3 The'following subjes.ts shall be reviewed by the' Corporate ~ ~ 4 Nuclear Safety Unit: ritten safety evaluations ~ of changes in the facility as a. described in the Safety Analysis Report, changes in procedures as described in the. Safety Analysis Report and tests or experi-ments not described in the Safety Analysis Reoort which are completed without prior NRC' approval under the provisions of 10lCFR 50.59(a)(1). This review is to veriff that such changes, tests, or experiments did not involve a change in the technical specifications or an unreviewed safety question as defined in 10 CFR 50.59(a)(2). b. Proposed changes in procedures, proposed changes in the facility, or proposed tests or experiments, any,o,f which involves a change in the Technical Specifications or an unreviewed safety question pursuant to 10 CFR 50.59(c). satters of this kind shall be referred to the Corporate Nuclear Safety Unit by the c l Plant Nuclear Safety Comnittee following its review, or'by other functional organizational units within Carolina Power & ~ Light Company prior to implenentatien. Proposed changes to the Technical Specifications or this l c. I operating license.

1 --

~ " ' j 6-11 -,,---,,---v--.-

ADMINISTRATIVE CONTROL -- SUBJECTS REOUIRING INDEPENDEhi REVIEW (Continued) =- d. Violations, deviations and reportable events, which require ~ reporting to the HRC within 24 hours, and as defined in the plant technical specifications such as: 1. Violations of applicable codes, regulations, orders, Technical Specifications, license requirements or internal procedures or instructions having safety significance; and. 2 Significant operating abnormalities ~or deviations from ~ normal or expected performance of plant safety-related structures, systens, or components. Review of events covered under this paragraph shall include ~ ~ ~ the results of any investigations made and the recocmendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

e. ' Any other matter involving safe operation of the nudlear power p, ant which the Manager - Corporate Nuclear Safety and Quality l

Assurance Audit Section decs appropriate for consideration, or which is referred.to the Manager - Corporate Nuclear Safety and Quality Assurance Audit Section by the onsite operating organization or by other fuctional organizational units within Carolina Power and Light Company. f. Reports and meeting minutes of the PHSC. FOLLOW-UP ACTION _ 6.5.3.4 Results of Corporate Nuclear Safety (CNS) reviews, including l recocmendations and concerns shall be documented. 7 Copies of the documented review shall be retained in the Cor-a. . potate Nuclear. Safety and Quality Assurance Audit Section files. ~ 3

b.. Recocmendations and concerns shall be subsitted to the Vice President - Nuclear Operations within.14 days of determination.-

l A summation of, Corporate Nuclear Safety recomendations and t c. I concerns shall be submitted to the Chairman / Chief Executive

u f

Officer; Senior Executiv'e Vice President and Chief Operating Officer; Executive Vice President - Power Supply and Custcmar Services; Senior Vice President - Power' Supply; Vice Pr'esident-Nuclear Safety and Research; Plant General Manager; and others, i-as appropriate on at least a bi-monthly frequency. i. 5.5.3.5 The Corporate Nuclear Safety Unit review program shall be

enducted in accordance with written, approved procedures.

6-12 ~. _ m.m.e i I y.. -.,_'r_ _ _., -.,.. _, - -, _,.,, _. _. w ,r.y s.

~ ~* ADMINISTPATIVE CONTROLS 6.5.4 OPERATION AND l'AINTENAtlCE UNIT (OMU) T FUNCTION 6.5.4.1 The Operation and Maintenance Unit of the Corporate Nuclear Safety and Quality Assurance' Audit Section shall perform audits of. plant - activities. These audits shall encompass: a. The confonnance of facility operation to all provisions con-tained within the Tec'hnical Specifications and applicable.- license conditions at least once per 12 months. b. The training and qualifications of the entire facility staff ~ at least once per 12 months. i c. The results of actions taken to correct deficiencies occurrine in facility equipT.ent, structures, systems, or method of operat. ion th'at affect nuclear safety at least once per 6 months. d. The verification of crapliance and implementation of the require.. ants of the Quality Assurance Program.to meet the criteria of Apper. dix "B",10 CFR 50, at least once per 24 months. e.' The Emergency Plan and implementing procedures at least once per 24 months. i f. The Security Plan and implementing procedures at least once l per 24 months. l g. The Facilit'y Fire Protection Program and implenenting procedures at least once per 24 months. h.' Any other area of facility operation considered appropriate by the Corporate Quality Assurance Audit Operation and Maintenance Unit, the Executive Vice President - Power Supply and Customer Service, or the Senior Vice President Power. Supply. e l l 6-13 2 ...... * * * = = = = = = = = =

T-

1. The radiological environmental monitoring program and the results thereof at least once per 12 months.

,_e

j. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

k. The PROCESS CONTROL PROGRAM and implementing procedures at least once per 24 months.

1. The ' performance of activities ~ required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.2, Revision 1, June 1974, and Regulatory Guide 4.1, Revision 1,. April 1975, at least once per 12 months..

e O 4 e em e 1 emusep E f I. e e a se 6-14

ADMINISTPATIVE CO.'fTROLS x-PERSONNEt. ~ 6.5.4.2 Audit personnel shall be independent of the area audited. a. Selection for auditing assignments is based on experience or training which establishes that their qualifications are car.:nensurate with the complexity or special nature of the activities to be audited. In selecting ' auditing personnel, consideration shall be given,to special abilities, specialized technical training, prior pertinent e_xperience, personal characteristics, and education. s-b. Qualified outside consultants or other individuals independent from thosem personnel directly involved in plant operation, but ~ within the Operations Group, shall be used to augment th6 audit 4 . teams when necessary. REPORTS 6.5.4.3 Rekults of audit are approved by the Manager - Corporate Nuclear Safety.and Quality Assurance Audit Section and transmitted directly to the Company President / Chief Executive Officer, the Senior Executive Vice President and Chief Operating Officer, the Executive Vice President - Power Supply and Customer Services, the Senior Vice President - Power Supply, and the Vice President - Nuclear Safety and Research, and ~ others, as appropriate within 30 days after the e pletion of the audit. 6.5.4.4 The Corporate Quality Assurance Audit Program shall be conducted ~ .in accordance with written, approved procedures. 6.5.5 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM 6.5.5.1. An independent fire protection and loss prevention program - inspection and audit shall be perfomed at least once per 12 months l utilizing an outside fire protection fim. 6.5.5.2 ~ An inspection and audit of the fire protection and loss prevention program shall be performed by a q'ualified outside* fire consultant,at ~ least once per 35 months. a e e O G O e g e

==ammes ene 6-J3 H%D . e e e. e e y T l

=_ f A_ DMINSTRATIVE CONTROLS -~ 5.6 REPORTABLE OCCUREU;CE ACTION 6.5.1 The following actions shall be taken for REPORTABLE OCCURP.BiCES: a. The Comission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.'3. b. Each REPORTABLE OCCURRENCE requiring 24 hour notification to the Ccr=:lission shall be reviewed by the PNSC and subnitted to Manager - Corporate Nuclear Safety and Quality Assurance Audit and the Vice President - Nuclear Operations. 6.7 SAFETY LIMIT VIOLATION _ S 7.1 The following actions shall be taken in the event a Safety i.init is violated: , a. The facility shall be placed in at least HOT SP.lTD0kW within two hours. b. The Safety Limit violation shall be reported to the Cen:nission, the Vice President - Nuclear Operations and to the Manager - Corporate Nuclear Safety and Quality Assurance Audit within 24 hours. A Safety Limit Violation Report stia11 be prepared. The report

== c. shall be reviewed by the PNSC. This report shall describe (1) . applicable circumstances preceding the violation, (2) effects of the vi.olation upon facility ccaponents, systens or struc- ~ tures, and (3) corrective action taken to prevent recurrence. d. The Safety Limit Violation Report shall be s::buitted to the Comissie 1, the Manager - Corporate Nuclear Safety and Quality Assurance Audit and the Vice President - Nuclear Operations within 14 days of the violation. S.8 PROCEDURES S.8.1 Written procedures shall be established, implemented and main-tained covering the activities referenced belo' : w a. The applicable procedures rectruended in Appendix "A" of Regulatory Guide 1.33 November,1972. - b. Refueling operations. c. Surveillance and test activities of safety related equipnent. d. Security Plan implementacion. ~- e. Emergency Plan implenentation. f. Fire Protection Pregram implementation. 6-16

e' ..~ ,-e ,, y 1 4.- gl 'the, onsite portion of the radiologica.1. environmental monitoring pro-gram implementation. i I ~

h. OFFSITE DOSE CALCULATION MANUAL implementation.

,I

i. PROCESS CONTROL PROGRAM implementation.

J. Quality Assurance Program for effluent and environ:nental monitoring, using the guidance in Regulatory Guide 1.11, Rerision 1, June 1974, ,'and Regulatory Guide 4 1 Revision 1, April 1975. u e J t i g< r .f / I y _, / l l r' _d,_# Y w ~t <!/ ~ l e + a r y I i ] (! .[ t ,s / l ~/' s',

==*=- 6-17' J G

ADMINISTRATIVE CONTROLS PROCEDURES (Continued) b 6.8.2 Each procedure of-6.8.1. above, and changes thereto, sha'11 be reviewed by the PNSC and approved by the General Manager prior to icple-mentation and reviewed periodically by the PNSC as set fort. in admin-istrative procedures. 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided: a. The intent of the nriginal procedure.is not altered. ~ b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the Brunswick Plant. c. .The change is documented, reviewed by the PNSC and approved by I the General Manager within 14 days of implementation. 6.9 REPORTING' REOUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES s 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to. the Director of the Regional Office of Inspection and Enforcement uniass otherwise noted. q% U STARTUP REPORT 6.9.1.1 A sumary report of plant startup and power esc'alation testin shall be submitted following (1) receipt of an operating license,. (2) g amendment to the license involving a planned increase in power level, l (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thenna1, or hydraulic perfor-mance of the plant. ~ '/ The startup report shall address each of the tests identified in the i' FSAR and shall include a description of. the m.easured val ~ues of the e L operating conditions or chaiacteristics obtained during the test pro-gram and a comparison of these values with design predictions and,- spedifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other co:mit-ments shall be included in this report. s.4,' /. a Startup reports shall be submitted within (1) 90 days follow-in ccwpletion of the startup test program, (2) 90 days following re-sumption or commencement of ccc:mercial power operation, or (3) 9 =onths following initial criticality, whichever is earliest. If the Startup 7-Report does not cover all three events (i.e., initial criticality, 6-18

AXN!STRATIVE C0!TTR0!.5 STARTUP REFORT (C:niinued)'. ~ c:cpietien of startup' test program, and rese=ptien or c:=encuent of c::=ercial pcwer eperation), supplenantary repcrts shall be subitted at least every three cenths until all three aYen:3 hiYe been c:=pletad,' A mjAL rep 0RTS I ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT [v,9,.3 Routine radiolog'ical environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. /s.9./. h These reports shall include:

1. Summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activi-ties for -the report period, including a comparison with pre-operational studies, operational controls (as appropriate), and previous environmental surveillance reports, and an assessment

. of the observed impacts of the plant operation on the environ-ment;

2. The results of land.use

' census required by Specification 3 1S St.

3. The results of analysis of all radiological environmental samples and of all measurements taken during the period pursuant to the Table and Figures in the ODCM,. as well as summarized and tabu-lated.results of these analyses and measurements in the format of Table 6.9-1. In the event that some individual results are not available for inclusion with the report, the report shall b a submitted noting and explaining the reasons for the missing results. The missing data should be submitted as soon as pos-sible in a supplementary report;

4. A summary description of the radiological environmental monitor-ing program; t

2

5. At least two legible maps covering all sample locations keyed to a tabla giving distances and directions from the centerline of one reactor;

6. The results of license participation in the Interlaboratory Comparison Program, required by Specification S.lf,3 :.and

7. Discussion of all analyses in which the LLD required by Table f,15.M was not achievable.

L/ A single submittal may be made for a multiple unit station. 2/ One map shall cover stations near the SITE BOUNDARY; a second shall in-clude the more distant stations. 6-19

KNN1Y0/ERATINGREp0RT Routine reperts of cperating statistics and shutd w expeMenee-

6. 4. / < S' shall be sub=ittad en a c:nthly basis to the Office of Inspection an'd..

Enferc~nt, U.S. Nuclear Regulatory Cc=ission, Washing::n, D.C. 20555, with a copy to.the bgicnal Office, to triive no later than the tenth of each =cnth folicwing the calendir =nd covered by, the rapcc., RE 0RTABLE OCC5JP.RENCo 4. 4. /. (, The REPORTABLE OCCURT54CES of Specific 2tiens (,.f. /. 7. and G.9.i.f below, including corrective acticas and cetsures' to prevent recurrence, shall be reported to the NRC. Supple =entti repcrts =ay.he required to In ctse of c:r e:ted or fully describe final rssolutien of occurrence. supplemental reports, t licensas event report shall be c=pleted'and reference chill be mda to the originti report date. O l 6-20 + - - yy -r

ADMINISTRATIVE CONTROLS .~ ~ PROMPT NOTIFICATI N WITH WRIiitN FOLLO Up ~ G,9 /,y The types of events listed below shall be reported within 24 hours by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate no later than the first working day following the event, with a written followrp report within two weeks. The written followup report shall include, as a minimum, a completed copy'of a licensee event report form. Infomation provided on' the licensee event report form shall be supple-mented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reache's the setpoint specified as the limiting safety system setting in the technical specifications or failure to co=plete the required protective function. b.' Operation of the unit or affected systems when, any parameter or operation. subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications., c. Abnomal degradation discovered in fuel cla'dding, reactor ' coolant pressure boundary, or primary co.ntainment. - d. Reactivity anomal'ies-involving' disagreement.with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% 6k/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifica-l tions; short-term reactivity increases that correspond to a j reactor period of less than 5 seconds or, if suberitical, an unplanned reactivity insertion of more than 0.5% ak/k; or occurrence of any unplanned criticality. Failure or malfunckion of one or more co=ponents which p'r'e9ents e. or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed . in the SAR. f. Pe'rsonnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents ant.lyzed in the SAR. 6-21

PROMPT NOTIFICATION.WITH WRITTEN FOLLO'J-UP (Continued) g. Conditions arising from natur.al or man-made events that, as a direct result of the event require plant shutdown, operation ^ of safet systems, or other protective measures required by technical specifications. h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifica- _ tions that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses. i. Performancet of struct:re.s systebs, or components that requires cemed4/ act4= or cweedse measures to prevent operation in -

n. mansee len conexMiss than. assumed in the accident analyses in the safety analysis report or technical specifica-tions bases; or discovery during plant life of conditions not specifically considered in the safety analysis report er technical specifications that require remedial action or cor-rective measures to prevent the existence or development of an unsafe condition.

s

j. Offsite releases of radioactive materials in liquid and gaseous effluents that exceed the Ifmits of Specification 3, //, 2 or 3r, /2, 2.

k. Exceeding the limits in Specification. 3, //,6 for the storage of radioactive liquids in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the, contents to within the specified limits.

l 6-22

TRTRTY~ DAY WRITTEN' REPORTS The types of e' ents listed below shall be the subject of 6,4,/ 7 v written reports to the Director of the Regional,0ffice within thirty l days of occur,rence 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 form shall be supplemented, as needed, by additional narrative mater ~ial to provide complete explanation of the circumstances surrounding the event. Reactor protection system or engineered safe ~ty feature instru-I a. ment settings which are found_to be less conservative 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 permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.\\ Observed inadeduacies in the implement'ation of administrative c. or procedural controls which threaten to cause reduction of ~ degree of redundancy provided in reactor protection systems or ' engineered safety feature systems. - d. Abnorrgl degradation of systems other than those specified in G.4.1,9 e above designed to contain radioactive material resulting from the fission process.

e. An unplanned offsite release of 1) more than 1 curie of radioactive mate-rial in liquid effluents,,2) more than 150 curies of noble gases in gas-eous effluents, or 3) more than 0.05 curie of radiciodine in gaseous ef-I fluents. The report of an ' unplanned offsite release of radioactive mate-rial shall include the following information:

1. A description of the event and equipment involved; i

2. Cause(s) for unplanned release;

3. Actions taken to prevent recurrence; and

4. Consequences of the unplanned release.

s l l 6-23 i .a----..,- -- - --,

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1 6.9.1.9 Routine radioactive effluent release reports covering the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. 6.9.1.10 These shall include the following:

1. A summary of the quantities of radioactive liquid'and gaseous effluents and solid waste released from the unit as out-lined in Regulatory Guide l'.21, " Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof;

2. A quarterly summary of hourly meteorological data collected during the reporting period in the form of joint frequency-distributigns of wind speed, wind direction and atmospheric stability;

3. An assessment of the radiation doses to individuals due to radioactive liquid and gaseous effluents released from the station during each calendar quarter;

4. The following information for each class of solid waste (as defined by 10 CFP. Part 61) shipped offsite during the report period:

a. Container volume I

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

c. Principal radionuclides (specify whether determined by measurement or estimate)

d. Source of waste and processing employed (e.g., de-watered spent resin, compacted dry waste, evaporator bottoms)

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

f. Solidification agent or absorbent (e.g., c ement,

urea formaldehyde)

5. A list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period; and, 6-24

6. Any changes made during the reporting period to the PRO-CESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CAL-CULATION MANUAL (ODCM), as well as a listing of new loca-tions for dose calculations and/or environmental moni-toring identified by the land use census pursuant to Specification 3.15.2.

1/ A single submittal may be made for a multiple unit section. The submittal should combine those sections that are common to all units at the station. 2/ In lieu of submission with the radioactive effluent release report, the licensee has the option of retaining this summary of required meteorological data in a file that shall be provided to the NRC upon request. l I i l l 6-25

l i . ~ .uwtJ.! ' 6,7-/.. .'llji' .. Awvi%wrp2. poicl tc#c noaineW 0- ecO6fAM SOM,N[,' l h. i ILGt esatuy. .%ket4 ~ ..kat:kf rahny 9 a-s%ss 9 %h,suw) l

1. 1 i m 0,,4 h<chas o< Pauway 1(a Oc*t'5 "I'h AIh'5E.^"*8 *"" Co"f"'I lod o

~ dioto r ~ it+$,,3efd,;,p ~ *i 9 b G

• d _et anew"M N "7 4,

tocg b r N As, Cow - Mow GP n?A.7]h a_.

a. m c0 **je oinam.xod.

Tsanse Rmene. ,m a m p z ,e j i ? i ~. 9.,4.i L.w y t.it 4 oct.a4 (u ni n dep.a A a I b* NY b W d (Afby d<tcCl4Mt lyi O % venttu h onlf. ffghe'n of k}<Cidble yh(MLAvtsws sli ab $ Cth gj !Old{ eon? a t'ad;yit.J & gudwi,c.

~ E~ SPECIAL REPORTS Special reports may be required covering inspections, test, and mainte-nance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications. 6.9.2 Special reports shall be submitted to the Director of the NRC Re-. gional Office listed in Appendix D, 10 CFR Part 20, with a copy to' the Director,0ffice of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, within the time period specified for each report. These reports shall include the fol-loving information: i

1. The cause(s) for exceeding the limit (s);

'2. The action (s) taken to restore the releas'e of radioactive effluents to be within the limit (s); and

3. A summary description of action (s) taken to prevent a similiar recurrence in the future 6.10 RECORD RETENTION Facility records shall be retained in accordance with A!!SI-N45.2.g'-1974.'

6.10.1 The following records shall be retained for at least five years: Records and logs of facility operation covering' time interval [ a. at each power level. b. Records and logs of principal maintenance activities, inspec-tions, repair and replacement of principal ite.ms of equipment related to nuclear safety. c. All REPORTABLE ~ 0CCURRENCE submitted to the Cccmission. d. Records of surveillance ' activities,. inspections and calibra-tions required by these Technical Specifications. i e. Records of changes made to Operating Procedures. f. Records of radioactive shipments. g. Records of sealed source and fission detectors leak tests and ( results. h. Records of annual physical inventory of all sealed source material of record. X.__ 6-27 l ll. . ~.

P.ECORD RETEhTION (Continued) ~ 6.10.2 The.following records shall be retained for the duration of the Facility Operating License: l a. Records and drawing changes reflecting facility design modif.i-cations made to syste.~s and equipment described in the Final Safety Analysis Report. b. Records of new and irradiat=d fuel inventory, fuel tFarisfers .and assembly burnup histories. c.. Records of facility radiation and contamination. surveys., d. Records or radiation exposure for all individuals entering radiation. control areas. e. Records of gaseous and liquid radioactive material released to the environs. f. Records of transient or operational cycles for those facility components identified in Table 5.7.1-1 I g. Records of reactor" tests and experiments. h. Records of training and qualification for current members of the plant staff. i. Reccrds of in-service inspections performed pursuant to these . Technical Specifications. j. Records of Quality Assurance activities required by the QA Panual. k. Records of reviews perfonned for changes made to procedures or equipment or reviews Df tests and experiments pursuant to 10 CFR 50.59. 1. Records of meetings of the PNSC and of the previous off-site review organization, the Company Nuclear Safety,Co:rer.ittee '(CNSC) Records for Environmental Qualification which are cov m. the provisions of paragraph 6.13.

n. Records of analyses required by the radiological environmental moni-toring program.

6-28

8.11 RADIATIDN PROT ~tCTION PROGRAM Procedures for personnel radiation protection shall be prepared consistea with the requirements of 10 CFP, Part 20 and shall be approved, maintaineef =. E= - and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA i 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each High Radiation Area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following: a. A radiation monitoring device which continuously indicates the. radiation dose rate in the area. b. A radiation' monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made know-ledgeable of them.. c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monito' ring device. \\. This' individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiatiion Work Permit.- 6.12.2 The requirements of 6.12.1, above, shall also apply to each hi radiation area in which the intensity of radiation is'-greater than 10' gh007..:C *it. mrem /hr. In addition, locked doors shall be provided to prevent unauthorized ' entry into such areas and the' keys shall be maintained under the administrative control of the Shift Foreman on duty and/or the Plant Health Physicist. l

• Health Physics personnel shall'be exempt from the RWP issuance require-ment during the performance of their assigned radiation protection duties, provided they ccmply with approved radiation protection pro-cedures for entry into high radiation areas.

l l l', (- 6-29 l ~ - ,-,,,-,,.3. y-. ,,,,,,-,,w ,.y,. -p ,,.m --,.m ,.r-

6.13 ENVIRONMDITAL 00ALIFICAT10N A. By no later than June '30.19EI all safety-related el ectrical equipent in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors " Guidelines for Evaluating Envircr;nental. Qualification of class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUP.EG-0553 " Interim Staff Position on Envirer. mental Qualification of Safety-Related Electrical Equipment". Decenber 1979. Copies of these documents are attached to Order for F,odification of License DPR 71 dated.0ctober 24; 1980. B. By.no later than. December 1,1980, complete and auditible records cust be availabit and r.aintained at a central location which describe the enviremental qualification method.csed for all.ufety-related electrical ecuipment in sufficient detail to, document the degree of compliance with the DDR Guidelines or NUP.EC--05SS. Thereafter, such reccrds should be updated and maintained current as equipent'is replaced, further tested, or other. se further qualified. 4 8 l l l 6-30 I.-- -

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Co= mission prior to implemen-tation. 6.14.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radio-active Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

a. Sufficiently' detailed information to totally support ratio-nale without benefit of hdditional or supplemental infor-mation. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval.and date box, to-gether with appropriate analyses or evaluations justifying the change (s);

b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations;

and,

c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSC.

2. Shall become effective.upon review and acceptance by the PNSC.

emer 8 I l

• =

i ~~ - 6-31 e-- ..-.w. ,,,.,,-,,-,,r,,,,--. m, a-, -, - - - ~, _,, _,,..

~~ 6.15 PROCESS CONTROL PROGRAM (PGP) @C ~_ 6.15.1 The PCP shall be approved by the Commission ' rior to implementa-p tion. 4 6.15.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the Semiannual Radio-active Effluent Release Report for the per'iod in which the change (s) was made. This submittal shall contain:

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

b. A determin,ation that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and,

c. Docum' ntation of the fact that the change has been reviewed e

and found acceptable by the PNSC.

2. Shall become effective upon review and acceptance by the PNSC.

O ? e b[ f f I E ~~~ 6-32 i l

7._~ _ _ 6.16 MAJOR CHANGES TO LIQUID, CASEOUS AND SOLID WASTE TREATMENT SYSTEMSI 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

1. Shall be reported to the Commission in the Semiannual Radio-active Effluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

a. A summary of.the evaluation that led to the determination that the change could be made in accordance with 10 CFR Tart 50.59.

b. Sufficient detailed Laformation to totally support the reason for the change without benefit of additional or sup-plemental dnformation;

c. A de. tailed description of the equipment, components and processes involved and the ' interfaces with other plant systems;

d. An evaluation of, the change that shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those pre-viously predicted in the license. application and admend-ments thereto;

e. An evaluation of the change that shows the expected maxi-mum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated'in the license application and admendments thereto;

f. A comparison of the predicted releases of radioactive mate-rials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when'the changes are.to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that the. change was reviewed and found acceptable to the PNSC.

2. Shall become effective upon review and acceptance by the PNSC.

1/ Licensees may chose to submit the information called for in this Specification as part of the annual FSAR update. 4 c J= -'-~ 6-33 -- --}}