Text

Proposed Tech Specs,Consisting of Change NPF-10/15-293. Change Allows for Implementation of Programmatic Controls for Radiological Effluent Tech Specs & Radiological Environ Monitoring Program
	ML20247C862
Person / Time
Site:	San Onofre
Issue date:	09/11/1989
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
Shared Package
ML13303B126	List: ML20247C862; ML20247C880;
References
	NUDOCS 8909140128
	Download: ML20247C862 (154)
	v • d • e

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7 '- s 51 NPT-10/15-293 J.

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ATTACHMENT A UNIT 2 EXISTING & PROPOSED TECHNICAL SPECIFICATION 8909140178 890211 PDR ADOCK 05000361 PDC p

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INDEX~ l 1

DEFINITIONS' -

.I

-. SECTION PAGE .{

1.0 DEFINITIONS

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1.1 ACTI0N...................................................... 1-1 i

12. LAXIAL SHAPE INDEX........................................... 1 1 1.3- AZIMUTHAL POWER TILT......................................... 1-1 1.4 ' CHANNEL-CALIBRATION......................................... 1-1 1.5- CHANNEL CHECK.... ..............'................ . .. ....... 1-1

. 1. 6 - CHANNEL FUNCTIONAL TEST...................................:... 1-2; '

1.7 CONTAINMENT INTEGRITY....................................... 1-2 1.8 CONTROLLED LEAKAGE.......................................... 2 1.9. CORE ALTERATION.............................................. 1-2 1.10 DOSE EQUIVALENT I-131....................................... 1-3 1:11 ,E-AVERAGE-DISINTEGRATION ENERGY............................. 1 11.12- ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.13. FREQUENCY N0TATION........................................... 1-3 1.14- GASEOUS RADWASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.15 IDENTIFIED LEAKAGE.......................................... 1-3 1.15 0FFSITE DOSE CALCULATION MANUAL (0DCM)...................... 1-4 1.17 OPERABLE ' 0PERABILITY...................................... 1-4 1.'18 OPERATIONAL MODE-M0DE....................................... 1-4

j. 1.19 PHYSICS TESTS............................ .................. 1-4 F 1.20 PLANAR RADIAL PEAKING FACTOR - Fxy.......................... 1-4 i-

[< 1.21 PRESSURE BOUNDARY' LEAKAGE........:.......................... 1-4 (PCP)....................

. 1.22 PROCESS CONTROL PROGRAM . ........ 1-4 f[ 1.23 P U R G E- P U RG I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 l' 1.24 RATED THERMAL P0WER................................... ..... 1-5

.L >

.1.25 REACTOR. TRIP SYSTEM RESPONSE TIME........................... 1-5 1.25' REPORTABLE'0CCURRENCE....................................... 1-5

~

1.27 SHUTDOWN MARGIN............................................. 1-5 X 1.2E. SOFTWARE....................... .... ..................... 1 - 5 ,4-3 1. 29150LIDIFICATION. . . . ... ........ ........ . .. ... .. . 1-5 )

. CHECK................................................ 1- 5 (E.1.501500RCE 1.31 STAGGERED TEST BASIC 1-5 7'

1.32 THERMAL P0WER...................:........................... 1-6 1 1.33 UN IDENTI FI ED . LEAKAG E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 L 1. 34 - VENTI LATION EXHAUST TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.35 VENTING..................................................... 1-6 7 C)E L.S TE D MAY 16 lofi

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[0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS- ,,

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7SECTION PAGE 4 -3/4.21 POWER DISTRIBUTION LIMITS' 3/4.2.1 .....;................. 3/.4 2-l' LINEARLHEAT RATE.................

m. 3/4.2.2 PLA'NAR PLANAR RADIAL PEAKING FACTORS.............

C ...... 3/4 2-2 3/4.2.3 AZIMUTHAL POWER TILT................-.................... 3/4'2 3/4. 2.~4 l DNBR MARGIN......................... . .................... 3/4.2-5.

3/4.2.5. -RCS FLOW RATE........................................... 3/4 2-9 "3/4.2.6 REALTOR: COOLANT COLD' LEG TEMPERATURE.................... ~3/4 2-10 3/4.2.7 AXIAL SHAPE INDEX......'.. '.. ............................ 3/4 2-11

~3/4.2.8- PRESSURIZER PRESSURE.................................... 3/4 2-12 3/4.3 ~ INSTRUMENTATION:

. 3/4.3.1 ; REACTOR PROTECTIVE; INSTRUMENTATION...................... 3/4 3-1 .

3/4.3.2- ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION........... .............'.............. 3/4.3-13

3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING ALARM ~ INSTRUMENTATION........... 3/4 3-34 INCORE DETECTORS..................................... 3/4 3-41 SEISMIC INSTRUMENTATION.............................. 3/4 3-42 METEOROLOGICAL INSTRUMENTATION....................... 3/4 3-45 REMOTE SHUTDOWN INSTRUMENTATION...... ................ 3/4 3-48 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-51' FIRE DETECTION INSTRUMENTATION....................... 3/4 3-56 A RADIOACTIVE LIQUID EFFLUENT MONITORING 1 INSTRUMENTATION.................................... 3/4 3-63 )

gget.oglut ( RADI0ACTIVEiGATEOUS EFFLUENT ONITORING i'

INSTRUMENTATION...........~......................... 3/4 3-68 LOOSE PART DETECTION INSTRUMENTATION................. 3/4 3-74 3/4.3.4' TURBINE OVERSPEED PROTECTION............................. 3/4 3-75 t

3/4.4 REACTOR COOLANT SYSTEM 3/4.c.1 REACTOR CD')LANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION... ... .... ... . . ....... 3/4 4-1 HOT STANDBV............................ ............ . . 3/4 4-2 SAN ONOFRE-UNIT 2 IV

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INDEX'

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' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ,

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- PAGE SECTION 3/4.8.3 '0NSITE POWER DISTRIBUTION SYSTEMS OPERATING............................................ 3/4 8-13 1 SHUTD0WN.............................................

3/4 8-15L 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.................................- 3/4 8-16 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION -

BYPASS............................................. 3/4 8-31 e

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION..................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION......................................... 3/4 9-2 3/4.9.3 DECAY TIME..............................................

3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS....................... 3/4 9-4 s

3/4.9.5 COMMUNICATIONS.......................................... 3/4 9-5 3/4.9.6 REFUELING MACHINE.......................................

3/4 9-6 3/4;9.7 FUEL HANDLING MACHINE - SPENT FUEL STORAGE POOL BUILDING 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION 3/4 9-8 HIGH WATER LEVEL..................... ...............

LOW WATER LEVEL......................................

3/4 9-9 CONTAINMENT PURGE ISOLATION SYSTEM......................

3/4 9-10 3/4.9.9 3/4 9-11 l 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................

1 3/4 9-12

!~ 3/4. 9.11 . WATER LEVEL -- STORAGE P00L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4.9.1.2 FUEL HANDLING BUILDING POST-ACCIDENT CLEANUP FILTER

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3/4 9-13 l SYSTEM...............................................

3/4.10 SPECIAL TEST EXCEPTIONS 3/4 10-1 3/4.10.1 SHUTDOWN MARGIN.........................................

l 3/4.10.2' GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS............................

3/4 10-2 3/4 10-3 3/4.10.3 REACTOR COOLANT L00PS.............. ....................

3/4.10.4 CENTER CEA MISALIGNMENT AND REGULATING CEA INSERTION 3/4 10-4 LIMITS................................................ b 3/4 10-5 )

3/4.10.5(RADIATIONMONITORING/ SAMPLING...........................

DELETED 01198 i

' VIII MENDMENT NO. 3 SAN ONOFRE-UNIT 2

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-g m INDEX f LIMITING CCNDITIONS FO's CPERATION AND; SURVEILLANCE REQUIREMENTS

'PAGE SECTION' s ,

w .

L .: RADI0 ACTIVE EFFLUENTS p .3/4.11 B

[3/4.11.I'LIQUIDEFFLUENTS f b _

CONCENTRATION......................................... 3/4 .11-lh 00SE.'.:.....,........................................... 3/4 11-5:

.(LIQUIDWASTETREATMENT............................... 3/411-6j E LIQUID HOLDUP TANKS...................................

3/4 11-7 e

3/4.11.2 GASEOUS EFFLUENTS .g.

DOSE RATE............................................. 3/4 11-8 DO S E- NOB LE GAS E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4-11-12

< DOSE-RADIOIODINES; RADI,0 ACTIVE MATERIALS IN P' ARTICULATE FORM AND TRITIUM........................

3/4 11-13 GASE005'RADWASTE TREATMENT............................ . 3/4 11-14 EXPLOSIVE GAS-MIXTURE.................................

. 3/4 11-15 GAS STORAGE TANKS.....................................

3/4 11 E WASTE............................... 3/411-17)

[3/4.11.3 SOLID.RADI0 ACTIVE 1

' 3/411-19.J DOSE............................................

{3/4.11.4 T0TAL ,

3/4.12 RADIOLOGICAL ENVIP^NMENTAL MONITORING

) (

3/4 12-1 3/4.12.1 MONITORING PR0 GRAM....................................

3/4 12-11 l 3/4.12.2 LAND USE CENSUS...................................-...

3/4 12-12 3/4.12.3 INTERLABORATORYCOMPARISCNPR6 GRAM....................

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. . _ _ g IX AMENCMENT NO.if SAN ONOTRE-UNIT 2 _

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INDEX 1 BASES PAGE SECTION-3/4.9.6 REFUELING MACHINE.....................................

B 3/4 9-2 3/4.9.7 FUEL HANDLING MACHINE - SPENT FUEL STORAGE BUILOING...

B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION.............. B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM.............. B 3/4 9-3 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL ......................................... B 3/4 9-3 3/4.9.12' FUEL HANDLING BUILDING POST-ACCIDENT CLEANUP FILTER SYSTEM.............................................. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS........................... B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS.................................

B 3/4 10-1 3/4.10.4 CENTER CEA MISALIGNMENT AND REGULATING CEA INSERTION LIMITS....................................

B 3/4 10-1 jg,,

(3/4.10.5 RADIATION MONITORING / SAMPLING.........................B3/410-1) l l

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g 01 bN XIV AMENDHENT NO.32 SAN ONOFRE-UNIT 2 L-____-_______-___-

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'INDEX u.' BASES" I

N SECTION- P AG E, ,

. 3/4.11- RADI0 ACTIVE EFFLUENTS 3/4. H;1 LIQUID EFFLUENTS................... ................'.. B 3/4 U-1 3/4. H~.2' GASEOUS EFFLUENTS..................................... B 3/4 n-2

'3/4. n.3 SOLID RADIOACTIVE' WASTE............................... B 3/4' n-S 3/4.n.4~T0TdL' DOSE............................................. B 3/4' n-5 3/4.12 RADI0 ACTIVE ~ ENVIRONMENTAL MONITORING'

.3/4.12.1 MONITORING PR0 GRAM.................................... B'3/4-12-1 3/4.12.2 ' LAND USE CENSUS.................................v..... 'B 3/4 12-1 3/4.12.3. INTERLABORATORY COMPARISON PRDGRAM.................... 'B 3/4 12-2 t

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SAN ONOFRE-UNIT 2 XV

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R0 .PAGE-j5ECTION:

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. . . .AUTHO R I TY . . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .

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. 5 REC 0RDS..................................................-

ACTION'.=................,.............:.. 6-13

~6.6 REPORTABLE OCCURRENCE 6-13

6. 7 J S AF ETY LIMIT - VIO LATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6-13'

-6.8: PROCEDURES'AND PROGRAM 5......................................

6. 9 REPORTING' REQUIREMENTS 6-15 6.9.1l ROUTINE AND' REPORTABLE OCCURRENCES.......................

5TARTUP-REP 0RT......,................................. .6 6-16 ANNUAL REPORTS.......................................

'6-17 ANNUAL RADIOLOGICAL ENVIRONMENTAL' OPERATING' REPORT...

.6-17 SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT.......

5-19.

MONTHLY OPERATING REPORT.............................

OCCURRENCES................................ 6-19' REPORTABLE 6-19 PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP. . . . . . . . . . . .

7 6-21 THIRTY DAY WRITTEN REP 0RTS........................... 6-21 HAZARDOUS CARGO TRAFFIC REP 0RT. . . . . . . . . . . . . . . . . . . . . . .

6-21 6.9.2' SPECIAL REPORT 5,........................................

6-21

-6.10 RECORD RETENTION. ...........................................

6-23 6.11- RADIATION PROTECTION PR0 GRAM..... .......................... -

l 6-23 6'.12 HIGH iADIATION AREA.........................................

6-24 6.13' PROCESS CONTROL PROGRAM (PCP)............. .................

6-25 6.14- 0FFSITE DOSE CALCULATION MANUAL............................. }

6-25 )

(6.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TRE ATMENT p ~. , _ -

AMENDMENT NO.

SAN OND:RE-UNIT 2 XVIII

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' INDEX l-LIST OF TABLES TABLE PAGE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.....................-.. 3/4 3 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-54

~ 3.3-11 FIRE DETECTION INSTRUMENTS................................

33 3/4 3-57 3

3.3-1J2 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-64

-d 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l$_ SURVEILLANCE REQUIREMENTS................................

3/4 3-66 3

}

9 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION., 3/4 3-69 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................. 3/4 3-71 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION..................................... 3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................... 3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......... 3/4 4-19 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY......................... 3/4 4-21 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS..,.......................................... 3/4 4-22 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-25 4.4-5 REACTOR VESSEL MATERIAL SURVETU ANCE PROGRAM-WITHDRAWAL SCHEDULE................................................. 3/4 4-28 4.6-1 TENDON SURVEILLANCE...................................... 3/4 6-12 4.6-2 TENDON LIFT-OFF F0RCE.................................... 3/4 6-12a 3.6-1 CONTAINMENT ISOLATION VALVES............................. 3/4 6-20 3.7-1 STEAM LINE SAFETY VALVES PER L00P........................ 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS............................... 3/4 7-3 SAN ONOFRE-UNIT 2 XX hENDMENTNO.

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iin v INDEX LIST OF TABLES i

jf g . TABLE PAGE

-4.7-1 cSECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY

_ SAMPLE AND ANALYSIS. PROGRAM............................... 3/4 7-8 3.7-5 SAFETY-RELATED SPRAY AND/OR SPRINKLER SYSTEMS............ 3/4 7-31 3.7-6 FIRE HOSE STATION',.......................................

3/4 7-33 4.8-1. DIESEL GENERATOR TEST SCHEDULE........................... 3/4 6-7 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS......................... 3/4 8-11 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES....................................... 3/4 8-18 3.8 MOTOR ODERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES........................................... 3/4 8-32 3.10-1 RADIATION MONITORING / SAMPLING EXCEPTIONS................. 3/4 10-6

-4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND AhALYSIS PROGRAM... 3/4 11-2 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM.................................................. 3/4 11-9 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PR0GP.AM............ 3/4 12-3 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES.................................... 3/4 12-7

- 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECION (LLD).... 3/4 12-8 j B 3/4.4-1 REACTOR VESSEL T0VGHNESS................................. B 3/4 4-8 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................... 5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION.............. ............ 6-4 SEP 2 41555 4

l SAN ONOFRE-UNIT 2 XXI AMENDMENT NO. 3

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l DEFINITIONS:

10FF5ITE DOSE CALCULATION MANUAL (ODCM) 1.1EJ The 0FFSITE DOSE CALCULATION. MANUAL shall contain the metnocology and

' ( parameters used in.the calculation of offsite doses due to radioactive caseous j and liquid effluents and in the calculation of gaseous and liquid effluent Qonitoring alarm / trip setpoints. i 4

OPERABLE - OPERABILITY. .Ue I E P L/\C E ' %/ITN IN % AT l 1.17 A system, subsystem, train, component cr device shall be OPERABLE'or l

-have OPERABILITY vhen it is capable-of performing its specified function (s),-

and when all necessary attendant-instrumentation, controls, electrical power, cooling or seal water,. lubrication or other auxiliary equipment that are required for the system,' subsystem, train, component or device to perform its i

' function (s) are also capable of performing their related support' function (s). i OPERATIONAL MODE - MODE l

1.18 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1. .

PHYSICS TESTS 1.19_ PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of.10 CFR 50.59, or 3) otherwise approved by the. Commission.

PLANAR RADIAL PEAKING FACTOR - F y

1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane .

average power density of the individual. fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in r Reactor Coolant System component body, pit.e wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP)

. 1.22fThe PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and i formulation determination by which SOLIDIFICATION of radioactive wastes from (liquidsystemsisassured.

t KE PLAcc WITH 10 segr 2.

SAN ONOFRE-UNIT 2 1-4 i

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The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology l

and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The 00CM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Pro- i grams required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.f.

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

_ _ _ _ _ _ _ - _ _ _ _ _ _ . . _ - - - - - - - - - _ - - - - - - . - - - - - - - - j

_ = ~ -

)' DEFINITIONS PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of' discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3390 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

REPORTABLE OCCURRENCE 1.26 A REPORTABLE OCCURRENCE shall be any of those conditions specified in -

Specifications 6.9.1.12 and 6.9.1.13.

SHUTDOWN MARGIN l

1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming (1) all full length control element assemblies (shutdown and regulating) are which reactivity worth fully inserted except is' assumed forfully to be the withJrawn, single assembly and (of highest 2) n6 change in part length control element assembly position.

SOFTWARE 1.28 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation and procedures. N (SOLIDIFICATION ;

SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct J

{(outlineonallsides(free-standing). /

SOURCE CHECK b W L ET~E D 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response when the channe1~ sensor is exposed to a radioactive source.

SAN ONOFRE-UNIT 2 1-5 i i

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'INSTRUMENTATI'ON '

L3/4.3.3' MONITORING INSTRUMENTATION-b RADIATION MONITORING INSTRUMENTATION 1

4

-LIMITING' CONDITION'FOR OPERATION ~

~

'3.3.3.1.-The radiation tbnitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified limits.

APPLICABILITY: As shown in ble 3.3-6.

ACTION:

a. With a radiation acnitoring channel alarm / trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit.

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b .' - With one'or more radiation monitoring channels inoperable,'take the ACTION shown in Table 3.3-6.

c.. The orovisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

~

4. 3. 3.1 : Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in

', Table 4.3-3.

"See Special Test Exception 3.10.5.

a v SAN ONOFRE-UNIT 2 3/4 3-34 QWDENT NO. 31

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, INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM).

APPLICABILITY: At all times.*

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation '

channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioacthe liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in. Table 3.3-12. Exert best efforts to return the instrument to OPERABLE status within 30 days and, additionally, if the inoperable instrument (s) remain inoperable for greater than 30 days, 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 6.9.1.13b are not applicable.

l SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CK!CK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.

4.3.3.8.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , all pumps required to be providing dilu-tion to meet the site radioactive effluent concentration limits of Specifica-tion 3.11.1.1 shall be deterrined to be operating and providing dilution to the discharge structure.

{5ee Special Test Exception 3.10.5. j SAN ONOFRE-UNIT 2 3/4 3 5 - 6 g hENDMENTNO.

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, TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 28 - With the numbe'c of channels OPERABLE less than required by the Minimum Channel's OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and

b. At least two technically qualified members of the Fecility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for ;

grossradioa9tivity(betaorgamma)atalimitofdetectionof i at least 10 microcuries/ gram:

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/

gram DOSE EQUIVALENT I-131; j l

b. At leadt once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/ gram DOSE EQUIVALENT I-131; or

c. Lock closed valve HV-3773 and divert flow to T-064 for pro-cessing as liquid radwaste.

ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , rab samples are collected and analyzed for gross radio- ;

act{vity beta or gamma) at a limit of detection of at least l 10 microcuries/mi or lock closed valve 522U19-MUO77 cr !

S22U19-MUO78 and divert flow to the radwaste sump for processing as liquid,radwaste.

ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

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e TABLE NOTATION l(1)- The. CHANNEL FUNCTIONAL. TEST shall also demonstrate that automatic

~ isolation _of this pathway and control room) alarm annu'nciation. occurs if ~

any of.the' following conditions exists: "-

1.-

Instrument . indicates measured levels' above the[ alarm / trip setpoint;

2. Circuit ~ f aflure.

3. Instrument indicates' a downscale failure.

L L L(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 ever its intended range. of energy and -

measurement' range. -For subsequent CHANNEL CALIBRATION,' sources that have been related to the initial calibration shall' be used.

E -(3) CHANNEL CHECK shall consist of verifying. indication of flow during ;'

peri ods .. of ; rel eas e. CHANNEL CHECK shall be made at least once per_

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous,' periodic, or batch releases are 7 ,

made. .

li,

";f tne instrument controis are not in the cperate mode, procedures shall require ,

that the channel be declared inoperable.

W *,

3/4 2-57 AMEN ~ENT N:.If .)

k5;4:'.:::.E-UNIT 2

i i

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INSTRUMENTATION .

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1 LIMITING CONDITION FOR OPERATION ;

exples We - j p

3.3.3.9-- ThegadioactivDgan6us of fluenDmonitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with the s set to -

ensure that the limits of Specification 3.ll.2. 17are gtalarm/ trip setpoi not exceeded. he F 1 ralarm/ trip setpoints of these channels shall bequetermined in accordance with] i w

QheODCM.f. g 3.tl . 2.T

, APPLICABILITY: As shown in Table 3.3-1 F ;

ACTION:

a. Wittyg radioactive)gadous effluentJmonitoring instrumentation ,

channel aiarme ip setpoint less conservative than required by the i above.Scecifie .cion. Immediately suspend the release of_ radioactive 1 J (paseous effl snts monitored by the affected channel orideclar_e the_ l channel inoperabl %" ' p- & A PW:: Tau h Aw= M k5.5-

^ g_=y:q ,

b. With less than the minimum number gui ractoactivergaseous effluenti !

monitoring instrumentation channels OPERABLE, take the ACTION sJcwn' g

~

.-g_i___ ic Table 3.3-13 4xert best ef forts to return the instri==nt> t r _

~" U lRABLE status within 30 days and,/ Additionally, if the inoperable lm~'* k .

instrument (s) remain inoperable for greater than 30 days, explai

% w4=keDn the next Semiannual Radioactive Effluent Release Reportf iy the !

iiioperability was not corrected in a ti manner.

F

c. The provisions of Specification,s 3.0. . 0.4Dnd 6. 9.1.13_]p are not i applicable.

\

. SURVEILLANCE REQUIREMENTS QE?W@ % !

4.3.3.9 Eacheadioactivalgasttous effluent monitoring instrumentation channel !

shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE :

CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

's un30ete.

m ma :su a geekt byeA- 4.

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TABLE 3.3-13 (Continued) i l

TABLE NOTATION

(* At all times

- Durirg waste gas, holdup system operation (treatment for primary system I offgases). p ]

- - MODES 1-4 with any main steam isolation valve and/or any main steam iso-lating bypass valve not fully closed.

(1) Provided 2RT-7865-1 is equipped to automatically terminate containment purge release.

(2) Prior to completion of DCP$3N, Containment Airborne Radiation Monitor 2L.'-7804-1 performs the functions of 2RT-7828. 2RT-7804-1 is not equipped to monitor purge flow.

(3) Prior to completion of DCP53N, 2RT-7865-1 may perform this function for minipurge only. Otherwise comply with ACTION 36 if another means of continuously monitoring purge flow is not available.

(4) 2RT-7818 is not equipped to monitor process flow. If another means of con-tinously monitoring process flow is not available, then comply with ACTION 36.

(5) 2/3 RT-7808 is not equipped to monitor plant vent stack flow. If another means of continuously monitoring plant vent stack flow is not available, then comply with ACTION 36.

ACTIOW 35 - With the number of :hannels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiat-ing the release.

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

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

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provic'ed the flow rate is estiniated at least once per 8 ht.urs. System design characteristics may be used to estimate flow.

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab sathpies are taken at least once per12 hours and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t . _

JAN 11885 3/4 3-70 AMENDENT NO. 31 SAN ONOFRE-UNIT 2 e s

V .,

n .

'~

i- TABLE 3.3-13'(Continued)

TABLE' NOTATION f.

ACTION 38'- With the number of channels OPERABLE.less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway- .i OR Prior to campletion of DCP53N, and with Plant Vent Stack Monitor 2RT-7865-1 not capable of terminating containment purge. release, PURGING m6y continue' using 2RT-7865-1 provided that:

1)' Plant' Vent Stack' Monitor 2RT-7865-1 is aligned to the purge

^ stack for the duration of the purge; and,

2) Plant Vent Stack Monitor 2/3 RT-7808 or 3RT-7865-1 is OPERABLE and aligned to the plant vent stack; and,

3) When PURGING is complete, 2RT-7865-1 is realigned to the plant vent stack; and,

4) In the' event of a high activity alare during the PURGE from any of 2RT-7865-1, 3RT-7865-1 or 2/3 RT-7808, an operator immediately suspends containment PURGING and realigns 2RT-7865-1.to the Plant Vent Stack. }

t ACTION 39 - With the number of channnels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this

!- system may continue provided that the remaining OPERABLE channel '

' is aligned to the waste gas surge tank. With two channels .

inoperab_le, operation of this' system may continue provided that grab samples are taken at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 40 - With the number of channels OPERABLE less than required by the

Minimum Channels OPERABLE requirement, effluent releases via' the affected pathway may continue provided samples are con--

tinuously collected with auxiliary sampling equipment as required in Table 4.11-2. ;

JAN 11816]

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% - 'l m

TABLE 4.3-9 (Continued) ,

_ TABLE NOTATION 31--

6 At all times)

- During waste gas holdup system operation (treatment for primary system offgases). f '

- - Modes 1-4 with any main steam isolation valve and/or any main steam isola-ting valve bypass valve not fully closed.

(1) The CHANNEL FUNCTIONAL TEST shi.11 also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditians exists:#

1. Instrument indicates ineasured levels above the alarm / trip setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

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

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers tnat 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. y

[ @ (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

g (5) The CHANNEL CALIBRATION shall include the use of standard gas samples containirg a nominal:

1. One volume percent oxygen, balance nitrogen, and ;

1

2. Four volume percent oxygen, balance nitrogen.

(6) Priortoeachrbeaseandatleastoncepermor,th.

(7) Prior to completion of DCP53N, these

surveillance requirements are tc be performed en the instrumentation indicated by Table 3.3-13.

If the instrument controls are not set in the operate mode, procedures shall,j g 1 call for declaring the channel inoperable.

} AEND2;T NO. 31 SAN ONOFRE-UNIT 2 3/4 3-74

L , ,

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MECIALTESTEECEPTIONS M- ' '

'3/4.10.5 RADIATION MONITORING / SAMPLING 1

LIMITING CONDITION FOR OPERATION ,

3.10.5 TheOPERABILITYrequirementsof. Specification',3/4.3[2,3/4.3.3.1,-

'3/4.3.3.6, 3/4.3.3.8, and 3/4.3.3.9 for the: radiation monitoring and sampling-instrumentation listed in Table 3.10-1 may be modified per Table 3.10-l'.

provided the. requirements listed in Table 3.10-1 are met.

.APPLIC53ILITY: 'As shown in Table 3.10-1.

ACTION:

~

With.the THERMAL POWER or criticality condi+ ion exceeding'the limit Lfor~

monitoring / sampling instrumentation as- shown.in Table 3.10-1, immediately trip the reactor.

SURVEILLANCE REQUIREMENTS 4.10.5' The monitoring / sampling instrumentation-. listed in Table 3.10-1 shall be demonstrated OPERABLE accordance with Specification 4.3.2,.4.3.3.1, 4.3.3.6, (4.3.3.8 or 4.3.3.9, as applicable, except as modified by Table 3.10-1.

SAN ONOFRE-UNIT 2 3/4 10-5 p...', .

_____._____-__________o____

7 y ,,

V 1

3 E

TABLE 3.10-J RADIAT_ ION MONITORING / SAMPLING EXCEPTIONS

1. Testing performed pursuant to FSAR Section 11.5.2.1.5.2 in startup program shall satisfy the initial CHANNEL CALIBRATION for the following monitors prior to first exceeding 5% RATED THERMAL POWER:

a. Control Room Airborne Monitors 2RT-7824 2RT-7825 1

L. Containment Airborne Monitors. 2RT-7804-1 2RT-7807-2 !

c. Containment Purge Area Monitors 2RT-7856-1 2RT-7857-2

d. Containment Area Radiation - 2RT-7820-1 High Range Moniters 2RT-7820-2

e. . Plant Vent Stack Airborne Monitor 2/3RT-7808

f. Radwaste Discharge Line Monitor 2/3RT-7813

g. ' Blowdown Neutralization Sump Monitor 2RT-7817

h. Turbine Building Sump Monitor 2RT-7821

2. The following monitors and samplers shall be OPERABLE prior to first j exceeding 5% RATED THERMAL POWER:

. a. Main Steam Line Area Monitors 2RT-7874A1 2RT-7874B1 2RT-7875A1 2RT-7875B1

(

'b. Condenser Evacuation System - 2RT-7870-1 Wide Range Monitor

c. Purge / Vent Stack Monitors - 2RT-7865-1 Wide Range 3RT-786!,-1

d. Plant Vent Stack Flow Rate Monitor

e. Containment Purge Flow Rate Monitor

f. Condenser Evacuation fyrtem Iodine Sampler Particulate Sampler Flow Rate Monitor

3. T'he Steam Jet Air Ejector Monitor (2RT-7818) shall be OPERABLE prior to initial criticality, j l

(_ SAN ONOFRE-UNIl 2 3/4 10-6 Amendment No. 4 L

s: A r, ', '

r 3 TABLE 3.10-1 (Continued)

4. Testing performed pursuant to FSAR Sectino 14.2.12 in.startup program is acceptable for the initial CHANNEL FUNCTIONAL TEST for a period up to 30 days following initial criticality for the following liquid effluent monitors:

a. Radwaste Discharge Line Moniter 2/3 RT-7813

b. Blowdown Neutralization Sump Monitor 2RT-7817

c. Turbine Building Sump Monitor ,

2RT-7821

5. Continuous monitoring and sampling of the containment purge exhaust directly from the purge . stack shall be provided for the low and high volume (8-inch and 42-inch) containment purge prior to startup following the first refueling outage. Containment airborne monitor 2RT-7804-1 or 2RT-7807-2 and associated sampling media shall perform these functions ^

prior to initial criticality. From initial criticality to the startup following the first refueling outage containment airborne monitor 2RT-7804-1 and associated sampling media shall perform the above required functions. I C )

i e

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j 3/4 10-7 Amendment No. 4 (SANONOFRE-UNIT 2

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g' l3/4.11.1 LIQUID EFFLUEN_TS, !

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, C_0NCENTRATION' LIMITING CONDITION FOR OPERATION-3.11.1 1.. The concentration of radioactive material released from the site L(see Figure 5.'l-4) shall be-limited to the concentrations specified.in 10 CFR Part 20, Appendix.B Table 1II, Column 2 for radionuclides other than dissolved

, : -or entrained noble gases. For dissolved or entrained noble gases, the !

- concentration shall be limited to 2 x 10

microcuries/a1 total activity. , "

N APPLICABILITY: -At'all times s

ACTION:

With the concentration of radioactive material released from the site Lexceeding the above:1imits, immediately restore the concentration to within the above limits. .

SURVEILLANCE REQUIREMENTS 4.11.1.1.1- The radioactivity content of eac *atch of radioactive liquid-waste shal_1 be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.

4.11.1.1.2 . Post-release analyses of samples composited from batch releases

. shall be performed in accordance with Table 4.11-1. The results of the previous post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.11.1.1.

4.11.1.1.3 The radioactivity concentration of liquids discharged from continuous release p'oints shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods in the ODCM to assure that the concen-trations at the point of release are maintained within the limits of (Specification 3.11.1.1.

SAN ONOFRE-UNIT 2 3/4 11-1 2

9

m ... .

. y TABLE 4.11-1 \

RADIOACTIVE' LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum. of Detection Liquid Release Sampling- Analysis Type of Activity (LLD)

LType Frequency Frequency ,.

Analysis (pci/ml)"'

A. Batch Waste --

P , P

,,7 Release Each Batch Each Batch Principa} Gamma 5x10 d

Tanks Emitters

' 1. ' Primary Plant I-131 1x10

-6

. .. Makaup Storage Tanb -5

, P M Dissolved and 1x10

2. Radwaste Primary One Batch /M Entrained Gases Tanks , (Gamma emittars)-.

3. Radweste- P M

. Secondary Tanks Each Batch CompositeD H-3 1x10

-5 4.' Miscellaneous-Wasta Condensate _7 Gross Alpha 1x10 Monitor Tanks

5. Neutralization Sr-89, Sr-90 -8 Sump P Q h 5x10 Each Batch Composite

, Fe-55 1x10

-6

~7 D W Principa} Gamma 5x10 8.* Continuous Releases *'# Grab Sample Composite

Emitters Steam-Generator -6 Blowdown I-131 1x10

2. Turbine Building M M Dissolved and 1x10

-5 Sump Grab Sample Entrained Gases

3. Miscellaneous (Gamma Emitters)

Wast'e- Evaporator 0 M

, Condensate

Grab Sample Composite c

H-3 1x10

-5

~7

4. Salt Water Gross Alpha 1x10 Discharge From Component.

Cooling Heat 0 Sr-89, Sr-90 5x10

-8 Q c

. Exchanger Grab Sample Composite

5. Steam Generator 810wdown Bypass ##**

Fe-55 1x10-6 I [W5] V 't lewd (SANONOFRE-UNIT 2 3/4 11-2 AMENDMENT NO.19 )

-y 'r. l TABLE 4.11-1 (Continued) -

TABLE NOTATION

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

For a carticular measurement systen (which may include radiochemical separation):

4.66 sg bb0

E V 2.22 x 10* Y exp (-Aat)

Where:

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

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

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

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

2.22 x 10s is the number of transformations.per minute,per microcurie, Y. is the fractional radiochemical yield (when applicable),

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

The value of s., used in the calculation of the LLD for a particular measurement system shall bV based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.

In calculating the LLD for a radionuclides determined by gamma ray spectrometry, the background should include the typical contributions of other radio-nuclides normally present in the samples. Typcial values of E, V, Y and At should be used in the calculation.

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

"For a more complete discussion of the LLD, and other detection limits, see the following:

(1) HASL Procedures Manual, HASL-300 (revised annually).

(2) Currie, L. A. , " Limits for Qualitative Detection and Quar titative Determination - Application to Radiochemistry" Antl. Chem. 40, SL6-93 (1968).

(3) Hartwell, J. K. , " Detection Limits for Radioisotopic Countiig Techniques,"

Atlantic Richfield Hanford Company Report ARH-2537 (June 22, 1972)..

u _- ,

A

[SANONOFRE-UNIT 2 3/4 11-3

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c

y F

f a

'< . : TABLE 4.11-1 (Continued)

TABLE NOTATION' ,,

b. A: composite sample is one:in which the quantity of liquid sampled.is-proportional to the quantity of liquid waste discharged and in which' the method of sampling employed results in a' specimen which is re'prese'ntative of the liquids released.- ,

c. .To be representative of the quantities and' concentrations of ,

radioactive materials in liquid effluents, samples shall be -

collected continuously in proportion to the' rate of flow of the . .

' effluent stream.,.. Prior.to analyses,'all samp1hs taken'for:the-composite shallcbe thoroughly mixed in et der. for the composite- .

sample tolbe representative of the effluent release.-

d.. 'A batch releasefis:the discharge.of liquid wastes of a_ discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the DDCM, to assure representative sampling.

e. - A continuous release is the disenarge of liquid wastes of.a.

nondiscrete volume; e.g., from a. volume of system that has"an input flow during the continuous release.-

f. The principal gamma emitters for.which the LLD specification. applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, co-60, Zn-65, Mo-99, Cs-134 Cs-137, Ce-141, and Ca-144. This list' does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

Sampling of this flow 'is not required if, at least once per 31 days, condensate monitor tank bypass valve, SA 1415-21"-200, s is verified' locked shut.

Administrative controls shall provide for composite sampling of the continuous releases per nota b vice note e until January 1,1983.

Continuous proportional sampling shall be .in accordance with note c

~L from January 1, 1983 and all times subsequent as required by l Table 4.11-1.

. N Administrative controls shall provit'e for composite sampling of the

'. continuous releases per nota b vice note c until January 1, 1984.

Continuous proportional sampling shall be in accordance with note c j from January 1, 1984 and all times subsequent as required by Table 4.11-1.

Sampling of this flow is not required if at least once per 31 days blowdown bypass isolation valve (S21301MU618 for Steam Generator E088 and $21301MU619 for Steam Generator E089) is verified locked shut. )

i

'fW 0 41983 SA 3/4 11-4 L_N'ONOFRE-UNIT 2_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _AMENDMENT

- - - - - - - - _ _ _ - _ _NO.

n , . , - ,, . .

4

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' RADI0 ACTIVE EFFLUENTS DOLE-LIMITING' CONDITION'FOR OPERATION' 3.11.1^.2 The' dose or dose commitment to an individual'from radioactive' -

materials in-liquid effluents released, from each reactor unit, from the site (see Figure 5.1-4)_ shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the

. total body and to less than or equal to 5 mrem to any organ, ar.d

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

APPLICABILITY: :At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials .

in liquid' effluents exceeding any of the above limits, in . lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification-6.0.2, a-Special Report which identifies the cause(Q for exceeding the limit (s) and defines the corrective actions'taken tc reduce the releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with Specification 3.11.1.2

b. The provisions of specifications 3.0.3, 3.0.4 and 6.9.1.13b are'not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose calculations. Cumulative dose contributions from liquid effluents shall be tietermined in accordance with the ODCM at least once per 31 days.

k )

Y r m-(SANONOFRE-UNIT 2 3/4 11-5 3)

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RAD 10ACTD/E EFFLUENTS LIOUID WASTE TREATMENT l

LIMITING CONDITION FOR OPERATION '

3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appropriate portiens of the system shall be used te reduce the radioactive materials .in liquid wastes prior to their discharge when the projected doses

.due to the liquid effluent'from the site (see Figure 5.1-4) when averaged over

31. days,-would exceed 0.06 arem to the total body or 0.2 mrem to any organ.*

APPLICABILITY,: At all times.

ACTION:

e.. With the liquid radwaste treatment system' inoperable'for more than

'31 days or with radioactive liquid waste being, discharged without treatment and in excess of the above limits, in lieu'of any other-report required by Specification 6.9.1,' prepare and submit to the-Commission within 30 days pursuant to Specification 6.9.2 a Special.

Report which includes the.following information:

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

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

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days, in accordance with the ODCM.

4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at i least lb minutes at least once per 92 days unlass the liquid radwaste system has been utilized to process radioactive liquid efficents during the previous 92 days. '

x

( Per reactor unit N

f FEF s c- )

SAN ONOFRE-UNIT 2 3/4 11-6 )

j 3M. n. 2.1 - 3M //. 2. V L2eleic/

(RADIDACTIVEEFFLUENTS~

L 3/4.11.2 GA5EOUS EFFLUENTS DOSE' RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:

a. For noble gases: - Less than or equal to 500 arem/yr to the total body and less than or equal to 3000 meem/yr to the stin, and

b. For all radiciodines, tritium and for all radioactive materials in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding the above limits, iniediately decrease the release rate to within the above limit (s),

gjRVEILLANCEREQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.2 Yhe dose rate due to radiciodines, tritium and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and

'hperforminganalysesinacccrdancewiththesamplingandanalysisprogram pecified in Table 4.11-2, j e

I

N TABLE 4.11-2 )

f RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower.

Limit'of. ,

Minimum Detection. I Sampling Analysis Type of (LLD) l Gaseous Release Type Frecuency Frecuency Activity Analysis (uCi/ml)"' j P P A. Wasta Gas Storage Each Tank Each Tank Principal Gamma. Emitters 8 1x10 4 Tank Grab Sample

8. Containment Purge P P Principal Gamma Emitters 8 1x10

b b 42 inch Each Purge ** Each Purge H-3 1x10

b gb 8 inch M Principal Gam 6a Emmitters 8 1x10

Grab Sample -

H-3 1x10

C. 1. Condenser M D

M D

Principal Gamma Emitters 8 '1x10

Evacuation Grab System Sample H-3 1x10 *

2. Plant Vent Stack l** l D. All Release Types. Continuous # / I-131 1x10 18 as listed in 8 and Sampler Charcoal C above. Sample I-133 1x10 10 Continuous # l Principal Gamma Emitters 8 1x10 11 1~ Sampler Particulate (I-131,Others)

Sample Continuous # M Gross Alpha 1x10 11 Sampler Composite Particulate l

Sample _

Continuous # q Sr-89, Sr-90 1x10 11 Sampler Composite.

Particulate Sample Continuous # Noble Gas Noble Gases 1x10.s Monitor Monitor Gross Seta or Gamma D I I Principal Gamma Emitters 8 5x10 7 E. Incinerated 011 Each batch Each batch Grab Sample k b o

(SANONOFRE-UNIT 2 3/4 11-9 AMENDMENTNO.52) 1

3 f'

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' E TABLE-4.11-2 (Continued)'

~

TABLE' NOTATION 9 The LLD is the smallest concentration of radioactive material in a sample

~

a .-

that will to detected with 95% probability with 5% probability of...falselyc concluding that a blank observation represents a "real" signal.

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

4.66 s .

E V 2.22 x 10' Y + exp (-Aat) ,

4-

.Whare:

c , LLD.is the "a priori" lower liuit of detection as defined above (as microcurie per unit mass or. volume),

s is the standard deviation of the background counting rate or of tNecountingrateofablanksampleasappropriate(ascountsper.

minute),

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

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

-2.22 x 108 is the number of transformations.per minute per microcurie, Y'is the fractional' radiochemic'al yield (when applicable), .

A is the radioactive decay coristaat.for the particular radionuclides, and at'is the elapsed time between midpoint of sample collection and time of counting.(for plant effluents, not environmental' samples).

The value of ss used in the calculation of the.LLD for a particular measurement syEtem shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samplesL(as appropriate) rather than on an unverified theoretically predicted variance.

In calculating the LLD for a redionuclide determine by gamma ray spectrometry, the background should include the typical contributions of other radio-nuclides normally present in the samples. Typical values of E, V, Y and l At should be used in the calculation.

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

I l "For a more complett discussion of the LLD, and other detectiori limits, set the following:

(1) HASL Procedures Mant.al, HASL-300 (revised annually).

(2) Currie, L. A., " Limits for Qu ditative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).

(3) Hartwell, J. K. , " Detection Limits for Radioisotopic Counting Techniques,"

Atlantic Richfield Hanfore Company Report ARH-2537 (June 22,1972). j rSAN ONOFRE-UNIT 2 3/4 11-10 rg

( '

4 . .

g

~~ ~ ~~ ~ --

%g V 4

v '

c ,

' l h>^' TABLE 4.11-2 (ContinuedF 5 TABLE NOTATION IK

b. ~ Analyses shall also be performed following shutdown,<startup, or_a THERMAL POWER change exceeding 15 percent of.the RATED THERMAL POWER within a 1-hour' period.

c. Tritium grab samples shall be taken~at:1 east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when-the' refueling canal is flooded.

K

- 'd. Samples shall be changed at leasti once per 7 days and analyses shall_-

be completed,within'48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall _also be performed at least once per-24 hours for at least 7 days following each shutdown, startup..or-THERMAL POWER change exceeding 15 percent of' RATED THERMAL; POWER in

-one' hour and analyses shall be completed within 48-hours of' changing.' When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor'of 10.

e. . Tritium grab samples shall be taken at least once per 7; days.from the ventilation exhaust from the spent fuel poc1 arca, whenever.

spent fuel is.in-the spent fuel pool,

f. The ratio. of the sample flow rate to: the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

g. The principal gamma. emitters'for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, -

Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59,-

Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for-particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.

h. Incinerated oil may be discharged at points other than the plant vent stack. Release shall be accounted for based on pre-release grab sample data.

i. Samples for incinerated oil releases shall be collected from repre-sentative samples of filtered oil in liquid form.

3/4 11-11 AMENDMENT NO. 5 (SAN ONOFRE-UNIT 2

~

$ '1

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p 4 ].

' RADIOACTIVE' EFFLUENTS

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. DOSEf- NOBLE GASES-JLIMITING CONDITION FOR OPERATION p' 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site (see Figure 5.1-3) shall'be limited to the

following:

a. During any calendar. quarter: Less than or equal to 5 mrad for gamma-radiation and less than or equal to 10 mrad for beta radiation and, b.-. ' During any calendar. year: Less than or equal to-10 mrad for gamma radiation and..lsss than or equal to 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION-

a. With the. calculated air dose from radioactive noble gases in gaseous affluents exceeding any of the above' limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the .

Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines.the corrective actions taken to reduce releases and the proposed corrective actions to be taken'to assure that subsequent releases will be in compliance with Specification 3.11.2.2.

b. The provisions of Specifications 3.0.3 and' 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS-4.11.2.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

0 P

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AN ONOFRE-UNIT 2 3/4 11-12 L

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p' RADI0 ACTIVE EFFLUENTS ,]

p _ , ~ DOSE - RADICIODINES. RADI0 ACTIVE MATERIAi.S IN PARTICULATE FORM AND TRITIUM s.

L -

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H . LIMITING CONDITION FOR OPERATION ._

3.11.2.3 The dose to an individual from tritium, radiciodines and radioactive materials in particulate form with half-lives greater than 8 days in gaseous ~

affluents released, from each. reactor unit, from the site (see Figure 5.1-3)'

shall'be limited to the following:

l. a. . During any calendar quarter: Less.than'or equal'to 7.5' ares to any organ and,

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

c. .Less than 0.1% of the limits of 3.11.2.3(a) and (b) as a result-of burning contaminated oil.

i APPLICA8ILITY: 'At all times.

ACTION:

a.- With the calculated dose from the release of tritium, radioio' dines.

and radioactive materials in particulate' form, with' half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by-Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification.6.9.2,-a Special Report which-identifies the.cause(s) for exceeding the limit and-defines the corrective actions'taken'to reduce releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with Specification 3.11.2.3.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.

P 3/4 11-13 AMENDMENT NO. 52)

(SANONOFRE-UNIT 2

Y 1 I

f._ _

RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT ,

LIMITING C0KOITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the profected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-3), when

. averaged over 31 days, would exceed 0.2 mrad'for gamma radiation and 0.4 mrad for beta radiation. The appropriate pcrtions of 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 (see Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ."

APPLICABILITY: At all times.

ACTION:

a. With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above liniits, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report'which includes the following information:

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

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

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the ODCM.

4.11.2.4.2 The GA!!OUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM thall be demonstrated OPERABLE by operating the GASEOUS RADWASTE TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

These doses are per reactor unit.

SAN ONOFRE-UNIT 2 3/4 11-14 b'~*- j

\; ~?

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y ll. 5 - 3kl.//.'f T)Clc0 J uw~

i FRAD10 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION .

3.11.3 The solid radwaste system shall be OPERABLE and used, as applicable in accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATION and packaging of radioactive wastes to ensure mecting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times." .

ACTION:

a. With the packaging requirements of-10 CFR Part 20 and/or 10 CFR Part 71 not satisfied, suspend shipments, of defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specifica-tion 6.9.2 a Spacial Report which includes the following information:

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

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, 4

3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and

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

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS ,,

4.11.3.1 The solid radwsste system shall be demonstrated OPERABLE at least once per 92 days by:

a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or a v

b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accoreance with a PROCESS CONTROL PROGRAM.

( "See Specification 6.13.1. ,

p.p..m SAN ONDFRE-UNIT 2 3/4 11-17 6 g g... ,

b c' i ' r u " -

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s. ,

. RADIOACTIVE EFFLUENTS-I' y ' SURVEILLANCE REQUIREMENTS (Continued)

s. '

i; '4.11.3.2 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 wasta (e.g. , filter sludges; .

spent resins, other than dewatered brd type, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

~

a. If'any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be; suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICA-TION.' SOLIDIFICATION of the batch may then be resce,ed using the alternative SOLIDIFICATION parawters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify 1 SOLIDIFICATION, the PROCESS' CONTROL PROGRAM shall provide for.the

collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at.least 3 consecutive initial test specimens demonstrate SOLIDIFICATION.

The PROCESS CONTROL PROGRAM shall be modified'as required, as-provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

L aa 9

rt: '~

3/4 11-18 CSANONOFRE-UNIT 2 3..

E '

y >

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4 RADI0 ACTIVE EFFLUENTS

'33 11.4 TOTAL DOSE .

a.

~t

- LIMITINGCONDITIONFOROPERATION

3'.11.4 The dose or dose. commitment.to any member.of the public, due'to

' releases of radioactivity and radiatjon,.from uranium fuel cycle sources shall be limited to less than or' equal to 25 mrem to the total' body or any organ-(except the thyroid, which shall:be limited to less than or equal to 75 mrem) over 12 consecutive months.. -

' APPLICABILITY: At all times.

= ACTION:

o

a. With the calculated docas from the release of radioactive' materials in liquid or gaseous affluents exceeding twice the limits of Specification 3.11.1.2.a. 3.11.1.2.b. 3.11.2.2.a, 3.11.2.2.b.

3.11.2.3.a. or 3.11.2.3.b, in lieu of any other report required by.

Specification 6.9.1, prepare and submit a Special Report to the .

Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, within 30 days, which defines the corrective action to be taken to redur.e subsequent releases to prevent recurrence of exceeding the. limits of Specification.3.11.4.

This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel' cycle sources (including all effluent pathways and direct

. radiation) for a'12 consecutive month period that includes the release (s) covered by ttis report. If the estimated dose (s) exceeds the limits of Sper,1fication 3.11.4, and if the release condition resulting in violation of 40 CFR 19; has not already been corrected, '

the Special Report shall include a request for a' variance in accordance with the provisions of 40 CFR 190 and including the specified information of $ 190.11(b). Submittal of the report.is considered a timely request, and a variance is granted until staff action on the request is cumplete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the require-ments for dose limitation of 10 CFR Part 20, as addressed in other cections of this technical specification.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and

. gaseous effluents shall be determined in accordance with Specifications I 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

N ONOFRE-UNIT 2 3/4 11-19 'D ' ' R / - j

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r 3/4.12 RADIOLOGICAL ~ ENVIRONMENTAL MONITORING 1

3/4.12.1 ' MONITORING PROGRAM LIMITING' CONDITION FOR OPERATION-1 12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1. .

4

. APPLICABILITY: At all times.

ACTION:

. .a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1,.in lieu of any other report required by Specification 4.9.1, prepare and submit to the Com-mission, in the Annual Radiological Operating Report, a description of the reasons ".,r not conducting the program as required and the s plans for preventing a recurrence,

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting . levels of Table 3.12-2 when averaged over

, any calendar quarter, in lieu of any other report required by

Specification 6.9.1, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to Specification 6.9.1.13. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) + * * *> 1* 0 l' limit level (1) limit level (2) -

}

When radionuclides other than those in Table 3.12-2 are detected and are the . result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 l

and 3.11.2.3. This report is not required if the measured level'ef radioactivity was not the result of plant' effluents; howtver, in such.an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report,

c. With fresh leafy vegetable samples or fleshy vegetable samples unavail-able from one or more of the sample locations required by Table 3.12-1, in lieu of any other report required by Specification. 6.9.1, prepare -

and submit to the Commission within 30 days, pursuant to Specific-ation 6.9.2, a Special Report which identifies the cause of the unavail-( ability of samples and identifies locations for obtaining replacement

( samples. The locations from which samples were unavailable may then be deleted from those reouired by Table 3.12-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

l: d. The provisions of Specifications 3.0.3 cnd 3.0.4 are not applicable.

SAN ONOFRE-UNIT 2 3/4 12-1 IE! ' . ;

. ;r, , ..

RADIOLOGICAL ENVIRONMENTAL _ MONITORING u

, SURVEILLANCE REQUIREMENTS.

'4.12,1 The' radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in

'the ODCM and shall be analyzed pursuant to.the'~ requirements of Tables 3.12-1 i

and 4.12-1.

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TABLE 4.12-1 (Continued)

TABLE NOTATION

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

For a particular measurement' system (which'mayI include radiochemical separation): "

4.66 s U ~ II -

V -

2.22 -

Y -

exp(-Aat)

Where:

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

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

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

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

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

A is the radioactive decay constant for the particular radionuclides, and atisthe'elapsedtimebetweensamplecollectiob(orendofthe sample collection period) and time-of counting (for environmental I

samples, not plant effluent samples).

The value of sbused in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather t.han on an unverified theoretically predicted variance. In calculating the LLD for a radionuclides determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides

I normally present in the samples (e.g. , potassium-40 in milk samples).

Typical values of E, V, Y and at shall be used in the calculations.

L In calculating the LLD for a radionuclides determined by gamma-ray spectro-metry, the background should include the typical contributions of other radionuclides normally present in the samples (t..g. , potassium-40 in milk

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

L 3/4 12-9 T[. .y (SANONOFRE-UNIT 2 w l

l

-S g- _ l

,TABLE 4.12-1 (Continued) l TABLE NOTATION j It should he recognized that the LLD is defined as an a priori (before the fact)'-

limit representing the capability of a measurement' system anc not as'a posteriori ~

(after the fact) limit for a particular measurement."

b. LLD for' drinking water. .

I

c. Other peaks which are measurable'and identifiable, together with the '

radionuclides in Table 4.12-1, shall be identified and reported. .

4

For a more complete discussion of the LLD, and other detection limits, see the following:

(1) -HASL Procedures Manu~al, HASL-300 (revised annually).

(2) Currie. L. A. , " Limits for Qualitative Detection and Quantitative Determin-ation - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).

(3) Hartwell, J. K., " Detection Limits for Radioisotopic Counting ~ Techniques,"

Atlantic Richfield Hanford Company Report ARH-2537 (June 22, 1972). .

L J e

9 e

SAN ONOFRE-UNIT 2 3/4 12-10 r:q.._.. .

_._.-.__-um_-.____a__m__ _. ___-.._________.___2- _ . _ _ _ _ _ _ - _ -.s___.__..-______________ _ _ _ _ . - _ _ _ _ _ _ _ . _ _ _ ___m._-__._ ______ __ _.__.__ .______._________ __ _____ ________________m_______._m__.m__---

a M y c .

3 H RADIOLOGICAL' ENVIRON", ENTAL M NITORING

1 3/4.12.2 LAND ~USE CENSUS LIMITING CONDITION FOR DPERATION 3.12.2 A land use- census shall be conducted and shall identify the location of the nearest milx animal, the nearest residence and the nearest garden" of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. For. ele.ated-releases as' defined in Regulatory Guide 1.'111, Revision 1, July 1977, the land use census shall also identify the locations of all milk animals and all gardens of greater than 500. square feet producing fresh leafy vegeta'bles in each of the 16 meteorological sectors within a distance of three miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in ' lieu c' any other

- report required by Specification 6.9.1. , prepare and submit to the Commission within 30 days,. pursuant to.5 specification 6.9.2, a Special Report wnich identifies the new location (s).

b. With a land use t:ensus identifying a ' location (s) which yields a calculated dose or dose commitment via the same exposure pathway 20 percent greater than at a location from which samples are currently being obtained.in accordance with Specification 3.12.1, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specifica- '

tion 6.9.2, a Special' Report which identifies the new location. The new location shall be added to the radiological envt onmental monitoring program within 30 days. The sampling location, excludir.g

the control station location, having'the lowest calculated dose or dose commitment via the same exposure pathway may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE, REQUIREMENTS .

4.12.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 using that information wnich will -

provide the best results, such as by a coor-to-door survey, aerial survey, or i

by consulting local agriculture authorities.

" Broad leaf vegetation sampling may be performed at the site councary in the cirecticn sector with the hignest 0/Q in lieu of the garden census, gv i 61,9M

.. 3/' 12-11 MEGE*, *g , H QANcN0?RE-UNIT 2

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RADIOLOGICAL' ENVIRONMENTAL'~ MONITORING J '3/4!12?3 'INTERL L EARATORY COMPARISC:/ FROGRAM ,

la :/ ) '

,a '

I

. LIMITING CONDITION FOR~ OPERATION

..w i; . . . - . ,

... 3.12.31 Analyses shall be performed on- radioactive. materials supplied.as jart ,

Lof an.: Interlaboratory Comparison Program wnich hasLbeen approved by the 9 > -Commission. ,

n:

APPLICA5ILITY: ' Atiall times..

'"i.f.

f L' . ACTION:

o :a, With. analyses'not being performed as required above, report the

, . correcti ve actions'taken to prevent a recurrence to the Commission-in.the Annual Radiological Environmental Operating Report.

.b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE' REQUIREMENTS' J- .

4.12.3 - ~ A summary of. the' results obtained as part of. the above required

-Interlaboratory Comparison Program and in'accordance with.the 0DCM:shall be

.i included in the Annual Radiological Envirorsental Operating Report.

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- P MM i 519 3 SAN CNOFRE-UNIT 2 3/4 12-12 A!'EC'.ESTNO.16}

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_ _ _ _ _ _ _ _ _ _ __ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ m._._m_ -m______.___________.___m_-m_ _ _ . - . _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ -

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JINSTRUMENTATIONr

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3/4.3.3.7 FIRE DETECTION INSTRUMENTATION-20PERASILITY of the fire' detection 1 instrumentation. ensures that adequate' This-

.Twarning' capability.is available for theLprompt detection of fires.

f ferpability..is required in order to detect and locate' fires-in.their early '

istiges. Prompt detection of fires'will;raduce the potential for damage to

, ' safe shutdown and/or, safety-related equipment and.is an integral- element in' the-cvarallt facility fire protect'on program.

~

-In the event- that less than SC% of the fire detection instrumentation is

~

inoperable in any fire:aret/ zone, the' establishment of frequent fire' patrols in

> the affected. areas is required to provide detection capability until the inoperable instrumentation is: restored to OPERABLE'.

Since the fira detectors are non-seismic, a' plant visual inspecticn for

' fires is. required within'two hours following an earthquake (> 0.05g).- Since safe shutdown systems are protected by seismic Category I baFriers, any fire after an earthquake'should be detected by this inspection before safe shutdown systems would be affected. Additionally, to verify the continued OPERABILITY'

'of fire detection systems after an earthquake, an engineering evaluation of the

fire detection instrumentation in the required zones is required to be per-formed.within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months folkwing an earthquake. p 3/4.3.3 8 fRADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 3

ided to monitor

,-The radioactive liquid effluent instrumentation is mr erials in'11guid 9 41814 fand control, as_ applicable, the releases of radioactive'prov effluents during actual or, potential releases of liquid eriluents. The alarm / y trip setpoints'for these instruments shall be calculated in accordance with the procedures:in the DDCM to ensure that the alarm / trip will occur prior-to excceding the 11mits of-10 CFR Part 20. The OPERABILITY and use of this I instrumentation is consistent with the requirements of General Design ~ J

< Criteria 60.-63 and 64 of Ar==ndix A tcL10 CFR Part 50.

3/4.3.3.9 manIcacuv 7GA WOUS EFFlUI INSTRUMENTATION D- -

V

. (rzys.s.MyF3WirC@

l- Irie]eaaroacpfs gaannus w. affluenD instrumentation is providedito monitor ru ssames or rauipactive materials in gaseous -

gna consrei, se syyncanie, The effluents during actual or potential releases of gaseous effluents. >

alarm / trip hetpoints for these' instruments shall be calculated in accordance with.the procedures in the 00CM to ensure that Thisthe alarm / trip will instr"=ntation alsooccur prior j includeg/

to exceedina the .limito of 10 CFR Part 20.

orovisions/for monitoring and contro1 Ling the concentrat'ons of potentially The OPERABILITY and explosive gas' mixtures in the waste gas holdup system.use of this instrum

'D2 sign Criteria'60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.10 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose-part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in theThe pri-eary system and avoid or mitigate damage to primary system components.

S 8 3/4 3-4 GRENDMENT NO. 6J

. SAN ONOFRE - UNIT 2

4

3/4.10 SPECIAL' TEST EXCEPTIONS ,

1 .

BASES-3/4.10.1 SHUTDOWN MARGIN This'special test exception provides that a minimum amount of CEA worth is immediately available for reactivity control when CEA worth measurement tests are performed. This special test exception is required to permit the periodic verification of the actual'versus predicted core' reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

Although CEA worth testing is conducted in MODE 2, during the performance

, of these tests sufficient negative reactivity is inserted to result in temocrary entry into MODE 3. Because the intent is to immediately return to MODE 2 to continue CEA worth measurements, the special test exception allows limited operation in MODE 3 without having to borate to meet the SHUT 00WN MARGIN -

requirements of Technical Specification 3.1.1.1.

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual CEAs to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure CEA worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 CENTERCEAMISALIGNMENTANDREGULATIN{JEAINSERTIONLIMITS This special test exception permits the center CEA to be misaligned or Regulating Group 6 inserted beyond the Transient Insertion Limit during PHYSICS TESTS required to determine the isothermal temperature coefficient, moderator temperature coefficient and power coefficient. gy, f 3/4.10.5 RADIATION MONITORING / SAMPLING This special test exception permits fuel loading and reactor operation with radiation monitoring / sampling instrumentation calibration and quality assurance conforming to either FSAR procedures or Regulatory Guide 4.15 Rev 1, February 1979. This test exception is required to allow for a phased implementa-tion of Regulatory Guide 4.15 Rev. 1, February 1979. Equivalent instrumentation, quality assurance and/or calibration is provided until full implementation of Regulatory Guide 4.15 Rev. 1, February 1979.

2 A b

01 885 SAN ONOFRE-UNIT 2 B 3/4 10-1 AEiDEn NO. 3' l

L - - - - - _ __

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u 3/4.10 SPECIAL TEST EXCEPTIONS BASE 5 .

The containment airborne ~ monitors and associated sampling' media test )

4 exception:is required to allow for operation prior to and during installation of upgraded monitors / media. Adequate monitoring is provided until and subsequent to the completion of the upgraded installation. Extensive containment air mixing during high volume purge (MODES 5 and 6) occurs as a result of containment HVAC and fans resulting in representative air monitoring via either 2RT-7804-1 or 2RT-7807-2. During low volume purge operations (MODES 1, 2, 3 and 4) 2RT-7804-1 provides representative indication of purged air due to its location in the imraediate vicinity of.- the low volume purge c.xhaust.

e 4

3 (w~ 1e w E 3/4 10-2 d!'E'CE'i~ IC . '. E 5 A!i O'C?EE-U!i~T 2

_ _ _ _ _ _ _ _ _ _ _ . _____________-_-____w

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-3/4.11 RADI0 ACTIVE EFFLUENTS BASES

3/4.11.1 LIQUID EFFLUENTS 1 3/4.11.1.1(CONCENTRATION <

a j

( . This specification is provided to ensure that the concentration of '

radioactive materials released in liquid waste effluents from the site 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 bodies of water outside the site will j result in exposures within (1) the Section II.A design objectives of Appen- )

dix I,10 CFR 50, to an individual, and (2) the limits of 10 CFR 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 MPC in air (submersion) was converted to an equivalent concentra-tion in water using the methods described in International Commission on j QdiologicalProtection(ICRP) Publication 2.f g 3/4.11.1.2> k [ ,

F This specification is provided to implement the requirements of l 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 flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents 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 conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data, such that the actual exposure of an individual througn appropriate pathways is unlikely to be substantially j underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials iri liquid effluents are consistent with the methodology provided in Regulatory G91de 1.109, *

" Calculation of Annual Doses to Man from Routine Releases of Reactor. Effluents I

for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,"

Revision 1, October 1977 and R6gulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the l

Purpose of Implementing 1.ppendix I," April 1977.

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

, (sharingthatsystem. y 1

SAN ONOFRE-UNIT 2 B 3/4 11-1

(.4 .

g I

I RADIOACTIVE EFFLUENTS BASES 3Ckt \

3/4.11.1.3 /fl0VID WASTE TREATMENT 2 JV r The OPERABILITY of the liquid radwaste treatment system ensures that this 3 i system will be available for use whenever liquid effluents require treatment prior to release to the env M nment. The requirement that the appropriate portions of this system be ussi when specified provides assurance that the releases of radioactive materia's in liquid effluents will be kept "as low as '

is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50.and the design objective given in Section II.D of Appendix 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 objectives set forth in Section II.A of Appendix I, 10 CFR Qe.rt 50, for liquid effluents. _j 3/4.11.1.4 OUTSIDE TEMPORARY TANKS Restricting the quantity of radioactive material centained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, 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.

3/4.11.2 GASEOUS EFFLUENTS ~~~M e4cf 3/4.11.2.1 (DOSE RATJ .

This specification is provided to ensure that the dose at any time at the (ite s boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either 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)).

l For individuals who may at times be within the site boundary, the occupancy of l 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 mram/ year to the total body or to less than or equal to 3000 mrem / 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 mrem / 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 proportioned among the units Qharing that system. j l SAN ONOFRE-UNIT 2 B ~3/4 11-2 -

1 l-

e-L %

RADI0 ACTIVE EFFLUENTS, BASES

_ {Olt.\<Ne b

Y ;

3/4'.11.2.2[ DOSE-NOBLEGASES F

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 okerating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievabliP'. The Surveillance Requirements implement the requirements in Section III; A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

Tne dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are cor.sistent with the methodology provided in Regulatory Guide 1.109, "Calcula-tion of Annual Doses to Man from Routine Releases of Reactor Effluents for .53 Purpose of Evaluating Compliance wit 5 10 CFR Part 50, Appendix I," Revision ',

October 1977 and Regulatory Guide 1. ill, " Methods for Estimiting Aterpheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors." Revision 1, July 1977. The ODCM equations provided for determining the air doses at the site boundary are based upon the g historical average atmospheric conditions, j 4 % e4cd 7 3/4.11.2.3 M SE - RADI0 IODINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM l (AND 4M1iAUM

~

This specification is provided to implement the' requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appen-dix 1. 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 ar, is reasonably achievable." The ODCM calculational methods specifieo 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 calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calcu-lating the doses due to the actual release rates of the subject materials are

, consistent with the methodology provided in Regulatory ' Guide 1.109, J

1

[ ~j SAN ONOFRE-UNIT 2 B 3/4 11-3

I i

i RADI0 ACTIVE EFFLUENTS BASES P

f " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents i 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 specifications for radiciodines, radioactive materials in particulate form and tritium are dependent on the existing radionuclides pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1) individual inhalation uf airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent J Lexposure of man

t. jegc4,4gy. &'

3/4.11.2.4[GASEOUSRADWASTETREATMENT I '

The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be ut.ed, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, i

and the desi,n t objectives given in Section II.D of Appendix I to 10 CFR Part 50.

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

3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures cor.tained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrugen and oxygen concentrations from reaching these flammability iimits. These automatic control features include injection of dilutants to reduce the concentration belc.1 the flammability limits. Maintaining the concentration'of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

S SAN ONOFRE-UNIT 2 B 3/4 U-4 t

K -

t

(;n c og RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.6 GAS STORAGE TANKS

' Restricting the quantity of radioactivity contained in each gas storage -

tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individcal at the s ncarest exclusion area boundary will not exceed 0.5.res. This is consistent with. Standard keview Plan 15.7.1, " Waste Gas Systam Failure".

I 3/4.1*i.3 SOLID RADI0 ACTIVE WASTE

]

The OPERABILITY of the' solid redweste system ensures that the system will be available for use whenever solid radwestes require processing and packaging prior to being shipped offsite. This specification implements the requirements of'10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to.10 CFR Part 50. The pror parameters included in establishing the PROCESJ CONTROL PROGRAM may include, but are not limited to waste type, waste pH, wasta/ liquid /

solidification age..c/ catalyst ratios, waste oil content, waste. principal chemical constituents, mixing and curing times.

3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.

The, specification requires the preparation'and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix 1. For sites containing up to 4 reac-tors, it is highly unlikely that the resultant dose to a member of the public will excted the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result'in the limitation' of dose to a member of the public for 12 consecutive months to within the 40 CFR 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 cther nuclear fuel cycle- i facilities at the same site or within a radius of 5 miles must be considered.

If the dose to any member of the public is estimated to exceed the require-suents of 40 CFR 190, the Special Report with a request for a variance in accordance with the provisions of'40 CFR 190.11, is considered to be a timely request and ful fills the requirements of 46 CFR 190 until NRC staff action is completed provided the release conditions resulting in violation cf 40 CFR 190 have not already been corrected. An individual is not consideted a member of the public during any period in which he/she is engaged in carring out any .

operation which is part of the nuclear fuel cycle.

SAN ONOFRE-UNIT 2 B 3/4 11-5 h'1"%c3 a__ _ = _ _ _ _

1 Y

3/4.12 RADIOLOGICAL' ENVIRONMENTAL MONITORING l

BASES

.3/4.12.1 MONITORING PROGRAM

.The radiological monitoring program required by this speci provides measurements of radiation and of radioactive materials l lie.' those tion exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from-the station

. operation. This monitoring program thereby supplements the radiological

' effluent monitoring program by verifying that the measurable concentrations of radioactive materials'and levels of radiation are not higher than expected-en

.the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program cha ges may be initiated based on operational experience.

The detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in industrial laboratories. It should -

be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyser, shall be per-formed in such a manner that the stated LLDs will be achieved under routine

. conditions. Occasionally background fluctuations,' unavoidably small sample sizes, the' presence of interfering nuclides, or other uncontrollable circum-stances may render these LLDs unachierable. In such cases, the contributing factors will be identified and descr11ed in the Annual Radiological Environm?ntal Operating Report.

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the'use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting with local agricultural authorities shall be used. This census satisfies the requirements of lq Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to l gardens of greater than 500 square feet provides assurance that significant i exposure pathways via leafy vegetables will be identified anti monitored since

a garden of this size is the minimum required to produce the quantity l -

(25 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and caboege), and 2) a vegetation yield of 2 kg/ square meter.

B 3/4 12-1 tt. c ' ~ <

QANONOFRE-UNIT 2

_ _ _ _ _ _ _ _ _ __ - _ _ _ - - .- . - _ . 1

?: .

RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM g .

The requirement for participation in an Interlaboratory Comparison

' Program is provided to ensure that independent checks on the precision and accuracv of the measurements of radioactive material in environmental sample matric A are performed as part of the quality assurance program for environ .

mental monitoring in order to demonstrate that the results are reasonably valid.

U J h DFRE-UNIT 2 8 3/4 12-2

-x,. r m .c

ADMlli!STEATIVE CONTROLS i

b. Ir.-Flant Radiation lionitorino c

' A program which will. ensure the capability to accurately determine i the' airborne iodine concentration in vital arees under. accident .

conditions. This prcgram shall include the following:

(i) Training of personnel, ,

(ii) Procedures for monitoring,-and -

(iii) Provisions for maintenance of sanpling and analysis equipment.

c. Secondary Water Chemistry P

l* A progrta for monitoring of secondary water chemistry to inhibit steam generator _ tube degradation. .This program shall include:

4 (i) Identification of a sampling schedule for the critical variables and control points for these variables,

'(ii) Identification of the procedures used to measure the values of the critical variables, ,

(iii) Identification of process sampling points, including monitoring

. the discharge of the condensate purps for evidence of condenser in-leakage, (iv) Procedures for the recording and r.ana;ement of data, (v) Procedures defining corrective actio.., for all off-control point chemistry conditions, and (vi) .A precedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

~ ^

. d. Post- Accident Sampling A* program

which will ensure the capability to obtain and analyze reactor coolant,, radioactive iodines and particulate in plant gaseous effluents, and c'ontainment atmosphere samples under accident conditions. The programa shall include the training of personnel,

% the procedures for sampling and analysis and the provisions for y w l 5 maintenance of sampling and anclysis eqdipment.

9. 9 P.EPORTIf!G REQUIREMENTS ROUTIliE REPORTS AtiD REPORTABLE OCCURRENCES

~

6.9.1 In addition to the applicable reporting recuire ents of Tit.le 10, Code of Federal Regulations, the following reports sbE11 be submitted to the I;FC Regional Administrator unless otherwise ncted.

1:ot required to be implemented until Septecer 1, ICE 3.

, IA! 0*icrR~-U::IT 2 6-15 AIGiUSJT NO.17

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APR 2 91923

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.:~ F 4 g .Ra'sioactise Effluent Controls Program 4

A program'shall be provided conforming with-10 CFR 50.36a for the control'of. radioactive' effluents and for maintaining the doses ~to.'.

MEMBERS OF THE PUBLIC from-radioactive effluents as low as reasonably -

achievable... The program (1) shall be contained in-the ODCM, (2) shall.be implemented by operating procedures,'and (3)-shall in-clude remedial' actions to be taken whenever the program' limits'are exceeded; The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation . including surveillance tests and set-

-y point determination in accordance with the methodology in the ODCM;

2) Limitations on'the concentrations of radioactive material.

released in liquid effluents'to UNRESTRICTED AREAS conforming to n 10 CFR Part 20, Appendix B Table II, Column 2;

3) Monitoring, sampling, and analysis of radioactive liquid and '

gaseous affluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,"

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

5) Determination of cumulative and projected dose contributions from radioactive affluents for the current calendar quarter and e current calendar yearfin accordance with the methodology and Wf"I D ' parameters in the ODCM at least every 31 days',

6) Limitatichs on the operability'and use of the liquid and gaseous affluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radio-activity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose e - itment conforming to Appendix I to 10 CFR Part 50'

7) Limitations on the dose rate resulting from radioactive material.

, released in gaseous effluents to areas-beyond the SITE BOUNDARY conforming to the doses associatatt with 10 cro part 70 Annendix 8. Table II, Column 1.

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

l< 9) . Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131. Iodine-133, tritiumc and all radio-nuclides in particulate fom with half-lives greater than 8 days in gaseous affluents relea. from each unit to areas beyond the

<fTF AnUNDARY conformina to Anoendix I to 10 CFR Part 50.

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

Radiological Environmental Monitoring Program A program shall:be provided to monitor the radiation and radio-nuclides in the environs of'the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential' exposure pathways, and (2) verification of the accuracy of r the affluent monitoring program and modeling of environmental expo-sure pathways.- The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and

.(3) include the.following:

1) Monitoring,' sampling, analysis, and reporting of radiation and radionuclides in'the environment in accordance with the method-ology and parameters in the 00CM,

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

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

ADMINISTRATIVE CONTROLS-( -

-: x' fats w4-te*dk Q ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.L6f Routine radiological environmental operating reports covering the 'M 7pEration of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to ;

May l' of the year following initial critica11tv g 6.9.L7/The annual radiological environmental operating reports sr.all inc Psummaries, interpretations, and an analysis of trends of the results of the f h radiological environmental surveillance activities for the report period, ,

including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by SperiMation 3.12.2. If harmful effects or evidence of irreversible damage are aetected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8,- ,

December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiologi al environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions-from one reactor; and the results of licensee participation in the Interlaboratory Comparisoa Program, trequiretbySpecification3.12.3. ;

SEMIANNLAL RADI0 ACTIVE EFFLUENT RELEASE REPORT

6.9.1.8Mine radioactive effluent release reports covering the operation 1

.for T.ne unit during the previous 6 months of operation shall be submitted s within 60 days after January 1 and July 1 of each year. The period of the Qrstreportshallbeginwiththedateofinitialcriticality. _

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

SAA DNOFPE-UNIT 2 6-17 p

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k n Th( Annual Radiological Environmental Operating Repart covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

I,hk The Semiannual Radioactive Effluent Release Report covering the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall in-clude'a summary of the quantities of radioactive liquid and gaseous effluents and solid weste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCH and PCP and (2) in con-formance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to'10 CFR Part 50.

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6.9.1.0/ihe radioactive affluent release reports shall include a summary of '

' the quantities of radioactive liquid and gaseous effluents and solid waste

, released from the unit as outlined in Regulatory Guide 1.21, " Measuring, R Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of '

Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cocied Nuclear Power Plants," Revision 1. June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The radioactive effluent release report to be submitted 60 days after January 1 of each' year shall include an annual summary of hourly meteorological L data' collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind Speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of stability. This same' report shall include an assessment of the radiation doses due to the radioactive liquid and p seous effluents released from the unit or station during the previous calendtr. year. This same report shall also ir.;:1ude an. assessment of the radiation doses from radioactive liquid and gaseous. effluents to members of the public due to their activities inside the site boundary (Figures 5.1-3 and 5.1-4) during the report period.

All assumptions used in making these assessments (i.e. , specific activity.

. exposure time and location) shall be included in these reports. The meteoro .

logical conditions concurrent with tt: time of release of radioactive

' materials in gaseous effluents (as determined by samphng frequency and measurement) shall be used for detemining the gaseous pathway doses. The .

assessment of radiation doses shall be performed in accordance with the OFFSITE DOS: CALCULATION MANUAL (ODCM).

The radioactive effl ant release report to be submitted 60 days after Jancary 1 of t och year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases ard other nearby uranium fuel cycle sources (including doses from primary effluent pathways.and direct radiation) for the previous 12 consecutive months to show conformance eith 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose .

contribution from liquid and gaseous affluents are given in Regulatory Guide 1.109, Rny. 1.

The radioactive affluents release shall include the following information for aach type of solid waste shipped offsite during the report period:

s. Container volume,

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

c. Principal radionuclides (specify whether determined by measurement (8 or estimate), .

d. Type of waste (e.g., spent resin, compacted dry waste, evaporator j bottores),

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

f. Solidification agent (e.g., cement, ures formaldehyde).

w 1 SAN ONOFRE-UNIT 2 6-18 __

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ADMINISTRATIVE CONTROLS S

bheradioactiveeffluentreleasereportsshallincludeunplannedreleasesfrom the site to unrestricted areas of radioactive materials in gaseous and liquid effluents en a quarterly basis.

The radioactive effluent release reports shall include any changes to the-PROCESS CONTROL PROGRAM (PCP) made during the reporting period. ;

~

M0NTHl.Y OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the safety valves, shall be submitted on a monthly. basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, w*th a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effec-tive, In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted in accordance with 6.0,2.

REPORTABLE OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specifications 6.9.L12 and 6.9.1.13 below, including corrective actions and measures to prevent returrence, shall be reported to the NRC. Supplemental reports may be requ! red to fully

. describe final resolution of occurrence. In case of corrected or ippoleental reports, a licensee event report shall be completed and reference shall ce Made to the original report date.

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.12 The types of events listed below shall be reported within 24 hocrs by telephone and confirmed by telegraph, mailgram, or facsimile transmisshn u i

the NRC Regional Administrator, or his designate no 1cter than the first.

working day following the event, with a written followup report within '

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

. cvent report form shall be supplemented, as nnM W redtionsi narrative V m.ateM al to provide complete explanation of the circumstances w rrounding the tavent. ,

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 reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to comple.te the required protective

. function.

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l i-SEP 21 i984 (

SAN ONOFRE-UNIT 2 6-19 AMENDMENT NO. 25 l a

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- h.' Records of in-service inspections performed pursuant to these-

Technical Specifications.-

i.- ' Records'of Quality Assurance. activities required by the QA Manual, P N j,'

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Records;of reviews' performed for changes made to procedures'or equipment _ or reviews of tests'and experiments pursuant to~

10 CFR 50.59.

k.- : Records of meetings of.the OSRC and the NSG.

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1.. ' Records"of the service. lives of'all snubbers ~within the scope of t Technical Specification 3/4.7.6' including the date at which the- .

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service life commences,and associated installation and maintenance records.- -

m. JRe r of secondary water, sampling and_ water quality,.

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6.11 RA DROTECTIONPROGRAM y,d' R . Procedures for personnel radiation protection shall be prepared consistent.

fyn Lwith the requirements of 10 CFR Part 20_and shall be approved,' maintained and V u ! adhered to for. all operations involving personnel radiation exposure.-

_... 6.12 HIGH RADIATION AREA .

6.12.1 In lieu of the " control device" or " alarm signal" req'uired by -

paragraph.20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity.of radiation is greater than 100 arem/hr but.less than 1000 arem/hr

shall be barricaded and conspicuously posted as a high radiation' area and f~ entrance thereto shall be controlled by requiring issuance'of a Radiaticn.

Exposure Permit (REP)*. 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 g radiation dose rate in the &rea 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 knowledgeable of them.

" Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the REP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following approved plant radiation protection procedures for entry into high radiation areas.

SAN ONOFRE-UNIT 2 6-23 D N h0. 33

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

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ADMINISTRATIVE CONTROLS

c. An individual qualified in radiation protection precedures who is Lt? >

equipped with a radiation dose rate _ monitoring device who is ,

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 Radiation Exposure Permit.

6.12.2 _In addition to the requirements of_6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body cauld. receive in one-hour a dose greater than 1000 mrem shall be provided with locked doors to prevent ur, authorized entry,'and the keys shall be maintained under the administrative-

~ control of the Shift-Supervisor on duty and/or health' physics _ supervision.

Doors shall remain locked except during periods of access.by personnel under an

! approved REP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour-a dose in excess of 1000 ares ** that are located within large areas, such as PWR containment,' where no enclosure exists for purposes of locking, and no enclosure'can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a. warning device. 'In lieu of the stay time specification of the REP, direct or remote (such as use'of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation. protection procedures te provide positive

exposure control over the activities within the area.

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.#

.6.13.2 Licensee' initiated changes to the PCP: p

1. rShall be submitted to the Commission in the semi-annual Radioactive

< . --9 Effluent Releasa Report for the period in which the change (s) was j Jade. This submittal shall contain: ;

a. Sufficiently detailed information to totally support the rationale

%, =d - W[M 7 for the change without benefit of additional or supplemental I

inforasation;

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IA determination that the change did not reduce the overall 7 b.

~~S. ~ conformance of the solidified waste product to existing criteria dA for solid wastes; and ,

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9 c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to 6.5.2.

2. Shall become effective upon review and cceeptance pursuant to 6.5.2.

""Measuremeiit made at 18" from source of radioactivity.

The PCP shall be submitted and approved prior to shipment of " wet" solid radioactive waste. . .

SEP 211984 SAN ONOFRE-UNIT 2 6-24 AMENDMENT NO. 25

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., ed as required by Specification 6.10.6,n, l

SEAT 8 Suff5fent informatio' n to support the change together with the appropriate analyses or evaluations justifying the change (s);

l N-A determination that the change will maintain the overall con-formance of the solidified waste product to existing require-ments of Federal, State, or other applicable regulations.

1 l

t i ADMINISTRATIVE CONTROLS -

l 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCH:

F

1. Shall be submitted to the Commission in the Monthly Operating Report O g within 90 days of the date the change (s) was made effective. This G qubmittal shall contain:

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a. FSufficiently detailed infomation to totally support. the l rationale for the change without benefit of additional or 3 1c-N supplemental information. Information submitted should consist S of a package of those pages of the ODCM to be changed with each

.a.- w4 ll page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying dhe change (s);

J V

g b. FA determination that the change will not reduce the accuracy or "

hev4 (2 , sy[,and reliability of dose calculations or setpoint determinations;

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c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to 6.5.2.

f 2. Shall become effective upon review and acceptance pursuant to 6.5.2.

(6.15MAJORCHANGESTORADIOACTIVEWASTETREATMENTSYSTEMS(Liquid,Gaseousand l

)'

solid) 6.15.1 Licensee initiated major changes to the radioactive waste systems l (liquid, gaseous and solid):

)

1. Shall be reported to the Commission in the Monthly Operating Report I

for the period in which the evaluation was performed pursuant to 6.5.2.

. The discussion of each change shall contana

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

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

c. A detailed description of the equipment, components and ,

l processe:i involved and the interfaces with other plant systems;

d. An evaluation of the change which shows the predicted releases l

of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto:

SEP 211984

. SAN ONOFRE-UNIT 2 6-25 AMENDMENT NO. 25

}QfT.

Shall be'. documented and records of reviews performed shall be' retain-ed as required by Specification 6.10.E,e.,

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' Sufficient information-to support the change together with the appropriate aralyses or evaluations justifying the change (s))

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

Instar 13 D. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

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F ADMINISTRATIVE-CONTROLS 3

s e.' An evaluation of the change which'shows~the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously-estimated in the license application and amendments thereto; f.. A comparison of the predicted releases of. radioactive

. materials, in liquid _and gaseous effluents:and in solid waste, to the actur.1 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 pursuant to 6.5.2.

2 .Shall become effective upon review and acceptance pursuant to 6.5.2.

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(C SAN ONOFRE-UNIT 2 6-26 EP 211984 ENDMENT NO. 25

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-SECTION

. PAGE 1.0 ' DEFINITIONS L 1-

~1.2

'AXIAL ACTION......................................................

SHAPE 11 INDEX.......;................................... 1-1

.1.

1.4 3 - AZIMUTHAL POWER TILT ....................................... 1-1

1. 5 -

CHANNEL CALIBRATION...................................v..... . .1-1 CHANNEL: CHECK........,...................... .................. 1-l' 1.6- ' CHANNEL-FUNCTIONAL TEST..... .... .

.......... ............. 1-2 0

1.7 ' CONTAINMENT < INTEGRITY........................... ..... ..... 1-2 iL8 ' CONTROLLEDiLEAKAGE...... .................. .......... ...... 1-2 L9- CORE' ALTERATION........ . . . . . . . . .........,.................. 1-2 1.:10 LDOSE' EQUIVALENT I-131....................................... 3 1.11 E-AVERAGE DISINTEGRATION. ENERGY..................... . . . . . . . 1-3 1.12 ' ENGINEERED SAFETY FEATURES RESPONSE TIME.............. ..... 1-3'

L 13 yFREQUENCY NOTATION................ ... ..................... 1 1.14= GASEOUS,RADWASTE TREATMENT' SYSTEM............... . . . . . . . . . . 1- 3 1 cl.15 IDENTIFIED' LEAKAGE.......................................... 1-3

^1.16. OFFSITE DOSE CALCULATION MANUAL-(0DCM)........... .. ....... 1-4 1.17: OPERABLE' 0PERABILITY...................... ............... 1-4 11.18 OPERATIONAL MODE'- M0DE..................................... 1 :L19 PHYSICS TESTS........... a....... ........ ............... .1-4 (1.20 PLANAR RADIAL PEAKING FACTOR - F .

............... 'l-4 xy .........

L 21 PRESSURE BOUNDARY LEAKAGE..................... ............. 4.

1.22 PROCESS CONTROL PROGRAM (PCP)................ .. ...... . 1-4 1.23 PURGE - PURGING............................................. 1-5 P0WER...................

1.24 . RATED THERMAL ..................... 1-5'

- 1.25 REACTOR TRIP SYSTEM RESPONSE TIME....... .... ... .......... 1-5

~ 1.26 REPORTABLE OCCURRENCE................... . ................. 1-5 L 27 SHUTDOWN MARGIN..... .................. ................... 1-5

' 1.28 SOFTWARE.....m. .. .. . . . . .. . . . . ......................... . 1 -_3 b 1.29 [30LIDIFICATION.. . . . . . . .. ... ... . . . . . . . 1-5

1. 30 */l500RCE CHECK.................................................

1.31 FSTAGGERED TEST BASIS..................... ... ..............

1-6 1.32 THERMAL P0WER...... .......................... ............. 1-6 1.33 UNIDENTIFIED LEAKAGE...... ....................... . . . . . . 1-6 1.34 VENTILATION EXHAUST TREATMENT SYSTEM................. . . . . . . 1-6

- 1.35 VENTING................... ................ .. . . . . . . . . . . 1-6

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l ~. INDEX' LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE' REQUIREMENTS _ _ _ _ _

SECTION PAGE

'3/4'2 . POWER' DISTRIBUTION LIMITS

.3/4.2.1 -. LINEAR HEAT RATE........................................ 3/4 2-1 3/4.2.2 ^ PLANAR RADIAL PEAKING FACT 0RS........................... 3/4 2-2 3/4.2.3 AZIMUTHA L POWER TI LT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-3 3/4.2.4 DNBR MARGIN............................................. 3/4 2-5 3/4.2.5 'RCS FLOW RATE............................................ 3/4 2 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE........ ....... ... 3/4 2-10 3/4.2.7 AXIAL' SHAPE INDEX........................ .............m. 3/4 2-11 3/4.2.8 PRESSURIZER PRESSURE..................................... 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATI09...................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES. ACTUATION SYSTEM INSTRUMENTATION....................................... 3/4 3-13 3/4.3.3 MONITORING: INSTRUMENTATION RADIATION MONITORING ALARM INSTRUMENTATION... ....... 3/4 3-34 INCORE DETECTORS.............................. .. ... 3/43-41 SEISMIC INSTRUMENTATION.. ........................... 3/4 3-42 METEOROLOGICAL INSTRUMENTATION....................... 3/4 3-45 REMOTE SHUTCOWN IN3TRUMENTATION...................... 3/4 3-48 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-51 FIRE DETECTION INSTRUMENTATION....................... 3/4 3-57 V

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION............ g ....................... 3/4 3-64 I

6 lts.vc ADIDAcTIvtssASEOUS EFFLUEND MONITORING INSTtVMENTATION.................................... 3/4 3-69 LOOSE PART DETECTION INSTRUMENTATION................. 3/4 3-75 3/4.3.4 TURBINE OVERSPEED PROJECTIONS........................... 3/4 3-76 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 - REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION............................. 3/4 4-1 HOT STANDBY............................................. 3/4 4-2 ,fLa (j$iV 15198[?

SAN DNOFRE-UNIT 3 IV

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,f y INDEX-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS' p SECTION' PAGE:

3/4.11 RADIDACTIVf EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS FC0NCENTRATION...... ............. .................... 3/4 U-11 00SE..........y....................................... 3/4 11-5 4 eLIQUID WASTE = TREATMENT................................ 3/4 11-6j u

LIQUID HOLDUP TANKS.................... ..._....<..... 3/4 11-7.i I

3/4.11.2 GASEOUS-EFFLUENTS b

(DOSERATE.............................................. 3/4 11-87 DO S E - NOB L E G A S E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4'11-12 DOSE - RADI0 IODINES, RADIOACTIVE MATERIALS IN PARTICULATE FORM AND TRITIUM......................... 3/4 11-13 4ASE0VS'RADWASTE TREATHENT............................. 3/4 11-14).

EXPLDSIVE GAS MIXTURE................................. 3/4 11-15 GAS STORAGE TANKS..................................... 3/4 11-16' i b,

WASTE...............................

3/4.11.3 SOLID RADI0 ACTIVE 3/4 11-17 3/4.11.4. TOTAL 00SE......................................,..... 3/4 11-19 I3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM.................................... 3/4 12-1 ;

i 3/4.12.2. LtND USE CENSUS....................................... 3/4 12-11 I t

3/4.12.3 INTERLABORATORY COMPARISON PROGRAM....................

3/412-1]

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BASES e..

SECTION PAGE.

l 3/4.1F RADI0 ACTIVE EFFLUENT _S l:

3/4.11.1' LIQUID' EFFLUENTS.................'...................... ~B 3/4'11-1 3/4.11.2 .GASEQUS EFFLUENTS........... ........ ................ 2 B 3/4 11 P I3/4.11. 3 - SOLID RADI0 ACTIVE WASTE. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-5' 3/4.11.4 TOTAL'D0SE........................................... B3/411-5j Y

[3/4.12'RADI0ACTIVEENVIRONMENTALMONITORING 3/4.12.1 MONITORING PR0 GRAM..................... ............... , B 3/4 12-1

, 3/4.12.2 LAND USE CENSUS... ................................... B 3/4 12-1

~

3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM..................... B . 3/4 :.12-2 L j t

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INDEX-n L1 ,, ADMINISTRATIVE CONTROLS-4 SECTION PAGE tAUTHORITY;................. . ....................... 6-13 REC 0RDS........................... ................... .6-13 I

6.6 REPORTABLE OCCURRENCE ACTION.........,....................... 5 6.7 SAFETY' LIMIT' VIOLATION'. . . ................ ................... .6-14

,. '6.8^ PROCEDURES AND PROGRAMS..... ................. ........... .c 6-14 6.9 REPORTING'REQUI'REMENTS 6.9.1 ROUTIN'E REPORTS AND REPORTABLE OCCURRENCES.... ......... 6-16 STARTUP REPORT. ...... ................... ... .. ... 6-17 ANNUAL. REPORTS..... ................................. 6-17 4 ANNUALRADIOiOGICALENVIRONMENTALOPERATINGREPORT... 6-18 I -SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT....... 6-18.

MONTHLY' OPERATING REPORT............... ............. 6 R E PO RT AB LE OC C U RR EN C E S . . . . . . . . .. . . . . . . . . . . . . .. . .. . . . . . . . 6-20 PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP.... ... ... 6-20 THIRTY DAY WRITTEN-REPORTS................ ......... 6-22 HAZARDOUS CARGO TRAFFIC REPORT......... .......... ... 6-22 i ,

6.9.2 SPECIAL REPORTS...... ................. .. .... ...... . 6-22 L

6.10 RECORD RETENTION.............................. .. ......... 6-22 6.11 -RADIATION PROTECTION PR0 GRAM................................ 6-24 p 6.12 HIGH RADIATION AREA.................. ............... .. .. 6-24 6.13 PROCESS CONTROL PROGRAM (PCP). . . . . . . . . . . . . . . . . . . ........ 6-25

~

6 6.14 0FFSITE DOSE CALCULATION MANUAL.................. . ..... 6-2 h15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS... 6-29 ..

Y SAN ONOFRE-UNIT 3 XVI y15Bj@

01 L

INDEX

' LIST OF TABLES TABLE PAGE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE.

REQUIREMENTS............................................... 3/4 3-55 3.5-11 FIRE DETECTION INSTRUMENTS.................................

3/4 3-58 a,-__.

-1gdIADI[0ACTIVELIQUIDEFFLUENTMONITORINGINSTRUMENTATION..... 3/43-69 I

@ M .3-8 1e-av y (FRADIDACTIVELIQUIDEFFLUENTMONITORINGINSTRUMENTATIOR SURVEILLANCE REQUIREMENTS.. ............................. 3/43-63) 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.... 3/4 3-70

- 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

-SURVEILLANCE REQUIREMENTS.................................. 3/4 3-72 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION....................................... 3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION......................... .. 3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES........... 3/4 4-20 3.4-2' REACTOR COOLANT SYSTEM CHEMISTRY....... ................... 3/4 4-22 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS........ ...................................... 3/4 4-23 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE................... 3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE................................................... 3/4 4-29 4.6-1 SURVEILLANCE........................................

l TENDON 3/4 6-12 4.6-2 TENDON LIFT-OFF F0RCE...................................... 3/4 6-13 3.6-1 CONTAINMENT ISOLATION VALVES.................. ............ 3/4 6-21 3.7-1 STEAM LINE SAFETY VALVES PER L00P.......................... 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS................................. 3/4 7-3 1

JE ' i SAN ONOFRE - UNIT 3 XIX (AMENDMENTNO.Sg)

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B INDEX g I LIST OF TABLES TABLE PAGE

'4.7-1 SECONDARY COOLANT' SYSTEM SPECIFIC ACTIVITY SAMPLE'AND'

. ANALYSIS PR0 GRAM.................,............ ......... . 3/4 7-9 3.7-5' ' SAFETY-RELATED SPRAY AND/OR SPRINKLER SYSTEMS............. 3/4.'7-32 3.7-6' FIRE HOSE STATIONS........................................ 3/4 7-34 4.8-l' DIESEL GENERATOR TEST SCHEDULE............................. 3/4'8-7 4.8-2 8ATTERY SURVEILLANCE REQUIREMENTS.......................... 3/4 8-11

-3.8-1 . CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.................................................... 3/4 P-18 3.8-2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION-BYPASS DEVICES PERMANENTLY BYPASSED.............................. 3/48-32p h.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM.... 3/4 11-P 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM... 3/4 11-9 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.............. 3/4 12-3 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES..................................... 3/4 12-7 4 12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION..........

3/412-8) 83/4.4-1 REACTOR VESSEL T00GHilESS....................... .......... B?/4 4-8 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS...................... 5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION............................ 6-5 S

SEP 2 4 G55 SAN ONOFRE-UNIT 3 XX (AMENDMENTNO.22 l

DEFINITIONS .

OFFSITE DOSE CALCULATION MANUAL (ODCM) p 1.16 1The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology-and

( parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents.and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints.

s OPERABLE - OPERABILITY I lac.c Wkk Tr b

~~~

1.17 A system, subsystem, train, component or device shall be OPEP.ABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are i required for the system, subsystem, train, component or device to pcrform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.18 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. ,

PLANAR RADIAL PEAKING FACTOR - F xy 1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) g Gym A sw4- l 1.22 The PROCESS CONTROL PROGRAM shall conGinT he sampling, analysis, and '

ulation determination by which SOLIDIFICATION of radioactive wastes from Jiquidsystemsisassured. ;

i i

i b

SAN ONOFRE-UNIT 3 14 15 @ l l

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3 INSeptT l h

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The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodolo and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints and in the conduct of the Environ-mental Radiological Monitoring Program. ,The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Pro-grams required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1 6 and 6.9.1.6.

I A SEAT z.

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

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H . DEFINITIONS' l

PURGE ~ ' PURGING I < L 1. 23 PURGE or PURGING.is the controlled process. of. discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration'-

or other operating condition, in such a manner that replacement air or gas is-required to purify'the confinement.

-RATED THERMAL-POWER-1.24' . RATED THERMAL POWER shall be a total reactor core heat transfer rate to the' reactor coolant of 3390 MWt. -

REACTOR TRIP SYSTEM RESPONSE TIME y 1.25 The REACTOR TRIP; SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

REPORTABLE OCCURRENCE 1.26 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.12 and 6.9.1.13.

SHUTDOWN MARGIN 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which' the reactor is subcritical or would be subcritical from its present' condition

assuming (1) all fu11' length control elen it assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn, and (2) no change in part length control. element assembly position.

SOFTWARE 1.28 The digital computer SOFTWARE for the reactor protection system shall be the program codes includin their associated data, documentation and procedures.

g LIDIFICATION N 1 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid ens to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct utline on all sides (free-standing). y

~

SOURCE CHECK 1.3C A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

S SAN ONOFRE-UNIT 3 1-5 L - - - _ _ _ _ _ _ _ - - - _ _ - _ - - _ - - - - - - _ - - - - - - - - - - - _ - - - - - - - - - - -

g 5/4. 3.3.3 dcU INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING ~ INSTRUMENTATION LIMITING CONDITION FOR OPERATION l 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERA 8LE with their alare/ trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CN).

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel er declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent "

monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Exert best efforts to return the instrument to OPERA 8LE status within 30 days and, additionally, if the inoperable instrument (s) remain incperable for greater than 30 days, 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 6.9.1.13b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each radioactive 11guld effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.

.3.8.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , all pumps required to be providing dilu-2n to meet the site radioactive effluent concentration limits of Specifica-V >n 3.11.1.1 shall be determined to be operating and providing dilution to the Qi_chargestructure. J 5AN ONOFRE - UNIT 3 3/4 3-64 d b hMENOMENT NO. 4

N O

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s s L E R T i - i B U A R t s s t s s A L L a e a a a a a T F A 3 z n p p z p p F / i i y y i y y E G 2 l L B B l B B N a a D I - r t n n r n n

I D t n w w t w w U I e u e o o e u o o Q V n e u d d n e d d I O i N7 l w w i N w w L R L (1 f o o L ( o o P 8 f l l l l E t n7 E B B S t n B B V S n w E n w I R e o- s ) ) C e o ) )

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TABLE NOTATION ACTION 28 - With the number of channels.0PERABLE less than required by the Minimum ChanntCs 0PERA8LE requirement, effluent' releases may continue provided that prior to initiating _a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, 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 29 - With the number of channels OPERABLE less than required by the Minimus Channels OPERA 8LE requirement, effluent releases via this pathway _may continue provided grab samples are analyzed for

_ gross radioactivity (beta or gamma) at a limit of detection of at least 10 microcuries/ gram:

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific aci.tvity of the secondary coolant is greater than 0.01 microcuries/

gram DOSE EQUIVALENT I-131.

b. Atleastonceper24hourswhenthespecificactivjty of the secondary coolant is less than or equal to 0.01 microcuries/ gram DOSE EQUIVALENT I-131; or

c. Lock closed valve HV-3773 and divert flow to T-064 for '

processing as liquid radwaste.

ACTION 30 - With the number of channels 0PERA8LE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactlyity (beta or gamma) at a limit of detection of at least 10 microcuries/mi or lock closed valve 522U19-MU077 or 522U19-MUO78 and divert flow to the radwaste sump for processing as liquid radwaste.

ACTION 31 - With the number of channels OPERA 8LE less than required by the Minimus Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

3/4 3-66 AMENOMENT NQ. 46

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I TABLE 4.'3-8 (Continued)'

TABLE NOTATION (1).-The CHANNEL FUNCTIONAL-TEST shall also demonstrate that automatic isolation _of this pathway and control room alarm annunciation occurs if any of--the following conditiens exists:*

1. Instrument' indicates measured levels above the alarm / trip setpoint.

2.: Circuit-failure.

3. Instrument-indicates a downscale failure.

(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 CHANNEL CALIBRATION, sources that have -

been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall f.onsist of verifying indication of flow during; periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

"If the instrument controls are not in the operate mode, procedures shall require that the channel be declared inoperable.

p is mi i ONOFRE-UNIT 3 3/4 3-68

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4

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c INSTRUMENTATION'

-y- V '

, p5fDACTIVE)GAgt0US EFFLUENf)M0NITORING INSTRUMENTATION .

LIMITING CONDITION FOR OPERATION L Jgsm 9, , . U.c 3.3.3.9 Thegadioacti as6s offluent) monitori ' instrumentation. channels shown in Table 3.3-13 shall be OPERABLE wit 1ars/ trip setpoints set to ensure that the limits'of Specification C h_h J2 re not exceeded. Re larm/ trip setpoints of these channels sh_all be determined in accor~ dance wi_t h) -

l e ODCM.*/-

APPLICABILITY: As shown in Table 3.3-13 ACTION:

a. With p adioactiv s[ouseffluen onitoring instrumentation channel alarm / trip setpoint less conservative _than required _by the_ J--

above Specificatiorifismediately suspend the release of radioactive)

@aseouseffluentsmonitoredbytheaffectedchannelor/Fc~1arethe channel. inoperable a b.. With less than the minimum number of facloact_iyp t gasapus effluent /~~ .

monitoringinstrumentationchannelsIPERABLE,taketheACTIONdhown

%c<c. M in Table 3.3-la;rCExeFt best eff5rtiTo return the Einstrumentp to (w Jr. 0PERABLE status within 30 days and,jfidditionally, if the inoperable y,M [ instrument (s) remain inoperable for greater than.30 days, explai

\in the next Semiannual Radioactive Effluent Release Repor hy the inoperability was not corrected in a t ey manner.

' 9-

' c. The provisions of Specifications 3.0.3 .0.4(~ind 6. Q 13 g no_1,T applicable. ,Q y 4

& s.st A a6 pea g h s 4 6c., Cam m.o6sw pso M /

SURVEILLANCE REQUIREMENTS- pyehdat -

, (emttM@ -p 9-4.3.3.9 ETac ~ fradioactiveJgasfous affluenf monitoring instrumentation channel shall be demonstrated OPdABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

b*Continuousmonitoringandsamplingofthecontainmentpurgeexhaustdirec ;

from the purge stack shall be provided for the low and high volume (8-inch and 42-inch) containment purge prior to startup following the first refueling outage. Containment airborne monitor 3RT-7804-1 or 3RT-7807-2 and associated sampling media shall perform these functions prior to initial criticality.

From initial criticality to the startup following the first refueling outage ,

containment airborne monitor 3RT-7804-1 and associated sampling media shall Q form the above required functions. -

JAN 11 1985 SAN ONOFRE-UNIT 3 3/4 3-69 AMENDMENT NO. _20)

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.- Ic TABLE 3.3-13 (Continued) .

n TABLE' NOTATION M

C* nt n' I times > ,

)uring waste gas holdup system operation (treatment'for primary system offgases).

F*** MODES 1-4withanymainsteamisolationvalveand/oranymainsteam1solattnq valve bypass valve not fully closed.

(1) Provided 3RT-7865-1 is equipped to automatically teminate containment

~

purge release.

(2) Prior to completion of DCP53N, Containment Airborne Radiation Monitor 'l

[' 3RT-7804-1 perferas the functions of 3RT-7828. 3RT-7804-1 is not equipped to monitor purge flow.

.(3) Prior.to completion of DCPS3N,'3RT-7865-1 may perform this function for i

^

minipurge only. Otherwise comply with Action 34 if another means of continuously monitoring purge flow is not available.

(4) 3RT-7818 is not' equipped to monitor process flow. If another means of continuously monitoring process flow is not available, then comply with ACTION 36.

(5) 2/3 RT-7808 isinot equipped to monitor plant vent stack flow. If another means of continuously monitoring plant vent stack flow is not availabit;,

'then comply with ACTION 36. A ACTION 35 - With t,he number cf channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to. initiating the release: ,

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

b. At least two technically qualified meLbers of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend releases of radioactive affluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the o Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . System design characteristics may be used to estimate flow.

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analy::ed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway OR Prior to completion of DCPS3N, and with Plant Vent Stack Monitor 3RT-7865-1 not capable of terminating containment purge release, e PURGING may continue using 3RT-7865-1 provided that: y 1 SAN ONOFRE-UNIT 3 3/4 3-71 blENDMENT NO. 20 dN1_U9

g

{; ,

4 i

y _

, TABLE 3.3-13 (Continued)

TABLE NOTATION

~-

d) Plant Vent ~ Stack Monitor 3RT-7865-1 is aligre'd tolthe '

purge stack for the duration of the purge; and,

2) ' Plant Vent Stack Monitor 2/3 RT-7808 or 2RT-7865-11s OPERA 8LE and aligned to the plant vent stack;'and,

3) Whan PURGING is complete, 3RT-7865-1 is realigned to the plant vent stack; and, ,-

4)- In the event of a high activity alarm during the PURGE.

from any of 3RT-7865-1, 2RT-7865-1 or.2/3 RT-7808, an operator immediately suspends containment PURGING and realigns 3RT-7865-1 to the Plant Vent Stack., j ACTION 39 - With the number of channnels OPEr.ABLE one less than required by the Minimus Channels OPERABLE requirement, operation of this system may continue provided that-the remaining OPERABLE channel is aligned to the waste gas surge tank. -With two channels inoperable, operation of this syste.1: may continue 1 provided that grab samples are taken at ien t once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following four hours. j>

FACTION 40 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are con-tinuously cc11ected with auxiliary sampling equipment as p required in Table 4.11-2.

].

JAN 11 1985 SAN ONOFRE-UNIT 3 3/4 3-72 AMENDMENT NO. 20

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TABLE 4.3-9 (Continued) f . TABLE NOTATION S

fat'all timesJ

- During wasta gas holdup system operation'(treatment for primary system offgases).

k** MODES 1-4withanymainsteamisolationvalveand/oranymainsteam 7

, ' isolating vaibe bypass valve.not fully closed. L (1) The' CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of.this pathway and control'r any of the followir.g conditions-exists:gon alarm annunciation occurs if 1.- Instrument. indicates measured levels above the' alarm / trip setpoint.

2. Circuit failure..

3. Instrument indicates a downscale failure.

(2) TheCHANNELFUNCTIONALTESTshallalsodemonstratethatcontrolroog

-alarm annunciation occurs if any of the following conditions exists :

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrcment indicates a downscale failure.

(3)_ The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using

'in meastandards-that surement assurancehaveactivities been obtained with NBS. from suppliers that participate These standards shall permit-calibrating the system over its intended range of energy and

l measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial. calibration shall be used. 3 (Q e CHANNEL CALIBRATION shall include the use of standard gas samples v containing a nominal:

, 1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

[g he CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

F(6) Frior to each release and at least once per month.

7l (7) Prior to completion of DCP53N, these surveillance requirements are to be '

performed on the instruments indicated by Table 3.3-13.

If the instrument controls are not set in the operate mode, procedures shall (callfordeclaringthechannelinoperable. )) g SAN ONOFRE-UNIT 3 3/4 3-75 hMENDMENTNO. 20j

_ _ _ _ _ _o

b

//././ --

4.11 RADI0 ACTIVE EFFLUENTS I

3/4.11.'l LIQUID EFFLUENTS CONCENTRATION LIMITING COND_I. TION FOR CPERATION 3.11.1.1. The concentration of ratiicactive i.aterial released froin the site (see Figure 5.1-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for-radionuclides other than dissolved ar-entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10d microcuries/ml total activity.

APPLICA8ILITY: At all times.

ACTION: l With the concentration of radioactive material released from the site exceeding the above limits, immediately restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS ,

4.11.1.1.1- The radioactivity content of each batch of radioactive liquid waste sh311 be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods ir, the ODCM to assuie that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.

4.11.1.1,2 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 4.11-1. The results of the previoc:; post-release analyses shall be used with the calculational methods in the ODCM to assure that th concentrations at the point of release were maintained within the limits of Specification 3.11.1.1.

4.11.1.1.3 The radioactivity concentration of if quids disc.harged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods in the ODCM to assure that the concen-trations at the point of release are maintained within the limits of (Specification 3. ' 11.1.1. y CIOV 151982 SAN ON0FRE-UNIT 3 3/411-II

]

L; , , 3

. , ' . *2 . . . r,

,y .

k j TABLE 4.11a1

~

)-

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS ,PROGRkA ,

Lower Limit" i

. MiMaus .

of Detection l Type of Activity..

!. Liquid Release- Sampling Analysis (LLD) 1 Type- Frequency Frequency Analysis (pC1/ml)"-

u. ' A..Calch Waste P. P

~7

' Released . Each Batch Each Batch Principa}Gamsa $x10 Tanks - Emitters

1. Primary Plant I-131 1x10

Makeup Storage

-5

, Tanks

.P- M Dissolved and 1x10

2. Radwasta Primasy One Batch /M Entrained Gases 4

. Tanks- (Gamma emitters)

3. Radweste P M b Secondary Tanks Each Batch Composite H-3 1x10 I'

4.. Miscellaneous- 7 Wasta Condensate Gross Alpha 1x10 .

'Moniter Tanks _

~

5. Neutralization ~8 P Q Sr-89, Sr-?D 5x10 b i

........; Sump. . .._ . _._. Each Batch Composite

-6 L Fe-55 1x10 L,

L D W Principa} Gamma 5x1d"7 B. Continuous#

Releases Grab Sample Composite

Emitters

1. Steam Generator -6 I-131 1x10

.._ _. Blowdown _. _ _

_ r, -

2. Turbine Building M M Dissolved and 1x10 '

Sump .

Grab Sample' Entrained Gases (Gamma Emitters)

3. Misca1]anecus_. __ ~

Wasta Evaporator M Condentate*

D c -5 1x10 Grab Sample Composite H-3

~7

4. ' Salt Water Gross Alpha 1x10 Discharge From Component" .g Cooling Heat D Q Sr-89, Sr-90 5x10 C

Exchanger Grab Sample Composite

-6 3.- Steam Generator Fe-55 1x10 Blowdown BynassM** ,

AMEND"EMT NO. 7 SAN ONOFRE-UNIT 3 3/4 11-2 2 -M_LsR A

_N TABLE 4.11-1 (Continued)

TABLEJ0TATION

]>

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

]

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

4.66 s !

0* E V 2.22 x 106 Y - exp (-Aat)

Where:

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

shis the standard deviction of the background cot.nting rate or of tne counting rate of a blarik sample as appropriate (as counts per minute),

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

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

2.22 x 105 is the number of transformations per minute per sticrocurie, Y is the fractional radiocheinical yield (when applicable),

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

The value of suused in the calculation of the LLD for a particular measurement system shall bM based on the actual observed variance of the background counting '

l rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.

IncalculatingtheLLDforaradionuclidedeterminedbygammarayspectrometry,f the background should include the typical contributions of other radio- r i nuclides normally present in the samples. Typcial values of E, V, Y and i l at shou' be used in the calculation.

I It shoul, je recognized that the LLD is defined as an a priori (before the fact) liniit representing the capability of the measurement system and not as a posteriori (after the fact) limit for a particular measurement.*

WFor a more complete discussion of the LLD, and other detection limits, see the following:

(1) HASL Procedures Manual, HASL-300 (revised annually).

(?) Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).

(3) Hartwell, J. K. , " Detection Limits for Radioisotopic Counting Techniques,"

Atlantic Richfield Hanford Company Report ARH-2537 (June 22, 1972).

l l .NOV 151982 g0N0FRE-UNIT 3 3/4 11-3 --

j

_- - _ - _ _ D

H . . . I r L ..

J j

h TABLE 4.11-1 (Continuedi 1

~ '

. ' TABLE NOTATION )

4 d

be A contosite sample is one in which the quantity of liquid sampled is' proportional.to the quantity of liquid waste discharged and in which ,

the method of sampling employed results in a specimen which is representative of the liquids released.

c. To be representative of the quantitier'and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the affluent stream. Prior to analyses, all samples taken for-the composite shall be thoroughly mixed in order for,the composite

. sample to be representative of the affluent release, d.. A batch release is the' discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, esch batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure represent 1ttive sampling.

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

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, In-65 Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, b together with the above nuclides; shall also be identified and,

' reported.

Sampling of this flow is not required if, at least once per 31 days, condensate monitor' tank bypass valve, SA 1415-2%"-200, is verified locked shut.

Admin'istrative controls shall provide for composite sampling of the continuous releases per note b vice note c until January 1, 1983.

Continuous proportional sampling shall be in accordance with note c

~

from January 1, 1983 and all times subsequent as required by Table 4.11-1.

M Administrative controls shall provide for composite s apling of the continuous releases per note b vice note c until January 1,1984.

Continuous proportional sampling shall be in accordance with note c from January 1, 1984 and all times subsequent as required by Table 4.11-1.

- Sampling of this flow is not required if at least once per 31 days blowdown bypass isolation valve (531301MU618 for Steam Generator i E088 and 531301MU619 for Steam Generator E089) is verified locked shut. .

T*,P 0 4 E 3/4 11-4 (SANONOFRE-UNIT 3

._.______._.___._________-___-____mm_--

.-.--- - - =

1

t. :n t .-

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?

U RADIOACTIVE EFFLUENTS

' ')

-DOSE i LIMITING CONDITION FOR OPERATION 3.11.1.2. The dose or dose commitment to an individual from radioactive materials in liquid effluents released, from each reactor unit, from the site (see Figure 5.1-4)-.shall be limited:

a. During any calendar quarter to less than or equal to 1.5'arem to the total body and to-less than or equal to 5 mrem to any organ, and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commissien within 30 days, pursuaint to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions'taken to reduce the releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with Specification 3.11.1.2 '

b. The' provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

s SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose calculations. Cumulative dose contributions from liquid effluents shall be determined in acccrdance with the ODCM at least once per 31 days.

t;gy 151982 3/4 11-5 IQANONOFRE-UNIT 3

da i

RADI0 ACTIVE EFFLUENTS.

]

LIQUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the 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 (see Figure 5.1-4) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.*

APPLICABILITY: At all times.

ACTION:

a. With the liquid radwaste treatment system inoperable for more thar; 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

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

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

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days, in accordance with the ODCM.

4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 15 minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.

A Per reactor unit NOV 151982 l

(SANONOFRE-UNIT 3 3/4 11-6

i i

s 3N, d. 2.1 -

3/4.Il.24 c.6LJ ADI0 ACTIVE EFFtVENTS .

3 3/4.11.2 GASEOUS EFFLUENTS l DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site (see Figure 5.1-3) shall be limited to the following:

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

b. For all radioiodines, tritium and for all radioactive materials in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.

APPLICABILITY: At all times.

ACTION: ,

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

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.2 The dose rate due to radiciodines, tritium and radioactive materials in particulate form with half lives ' greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and I performing analyses in accordance with the sampling and analysis program QpecifiedinTable4.11-2. J l

l (ky 1s m2) r SAN ONOFRE-UNIT 3 3/4 11-8 L g dp )

{- 3 TABLE 4.11-2

-RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM' M1ntmum Lower Limit of

' Sampling Analysis ' Type of Detection (gLO)

Activity Analysis

' ' ~

Gaseous Release Type Frequency Frequency (pCf/ml)

P P A. Wasta Gas' Storage Each Tank Each Tank- Principal Gamma Emittersg 1x10 4 Tak Grab

' Sample B. Containment Purge 'P' P b

Principal Gamma Emitters 8 1x10

42 inch Each Purgeb 'C Each Purge H-3 1x10-=

D b 8 inch' M M ' Principal Gamma Emmittersh 1x10 4 Grab Sample H-3 1x10-*

b gb C. 1. Condenser M Principal Gamma Emitters 9 1x10 4 Evacuation Grab System Sample H-3 lx10-'

2. Plant Vent Stack- d d D. All Release Types Continuous f

/ I-131 1x10 12 as listed in B and Sampler Charcoal C above. Sample I-133 1x10 10 continuous f

/ Principal Gamma Emitters 9 1x10 22 Sampler Particulate (I-131,others)

Sample f

Continuous M Gross Alpha 1x10 11 Sampler Composite Particulate '

Sample f

L Continuous Q Sr-89, Se-90 1x10 12 Sampler Composite Particulate Sample f

Continuous Noble Gas Noble Gases 1x10 6 Monitor Monitor Grass Beta or Gamma i i E. Incinerated Oil h Each batch Each batch Principal Gamma Emitters 9 5x10 7 -1 Grab Sample 1 l .

SAN ONOFRE - UNIT 3 3/4 11-9 AMENDMENT NO. 41 lw_ ..

1

I Y

TABLE 4.11-2 (Continued)

TABLE NOTATION i

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

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

4.66 s D= E V 2.22 x 106 Y exp (-Aat)

Where:

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

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

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

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

2.22 x 108 is the number of transformations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclides, and at is the elapsed time between midpoint of sample collection hnd time of counting (for plant effluents, not environmental samples).

The value of sg used in the calculation of the LLD for a particular

measurement syEtem shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rath2r than on an unverified theoretically predicted variance.

In calculating the LLD for a radionuclides determined by gamma ray spectrometry, the background should include the typical contributions of f

other radionuclides normally present in the samples. Typical values of

! E, V, Y and at should be used in the calculation.

I J It should be recognized that the LLD is defined as an a priori (before j the fact) limit representing the capability of the measurement system and

not as a posteriori (after the fact) limit for a particular measurement.*

l

Fur a more complete discussion of the LLD, and other detection limits, see the following:

(1) HASL Procedures Manual, HASL-300 (revised annually).

l (2) Currie, L. A. , " Limits for Qualitative Detection and Quantitative l Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (196S).

(3) Hartwell, J. K. , " Detection Limits for Radioisotopic Counting Techniques,"

Atlantic Richfield Hanford Company Report ARH-2537 (June 22, 1972).

t SAN ONOFRE-UNIT 3 3/4 11-10

I TABLE 4.11-2 (Continued) p TABLENOTAT[0N :

' Analyses shall-also be performed following shatdown, startup, or a

~

b.

THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period,

c. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refue',fng canal is flooded.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or a.fter removal from sampler). Samp1'sig shall also be performed at le&st once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at.least 7 days'following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10. -

e. 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. .

f. The ratio of the samp'io flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation'made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

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

h. Incinerated oil may be discharged at points other than the plant vent stack. Release shall be accounted for based on pre-release grab sample data.

i. Samples for incinerated oil releases shall be collected from repre-l sentative samples of filtered oil in liquid form. ,

^

~_ _

A l

3/4 11-11 hANONOFRE-UNIT 3 AMEN 0 MENT NO.

4) u l

y

[

i.

!T ,bRADI0ACTIVEEFFLUENTS 7 DOSE - NOBLE GASES 5 LIMITING CONDITION FOR (PERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site (see Figure 5.1-3) shall be limited to the following-l- a. During any calendar quarter: . Less than or equal to 5 mrad for gamma radiation and.less than or equal to 10 mrad for beta radiation and,

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

APPLICABILITY: At all times. ,

ACTION

]

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the;above limits, in lieu of any .other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the-limit (s) and l- defines the corrective actions taken to reduce releases and the proposed corrective actions to be taken to assure that subsequent relqases will be in compliance with Specification 3.11.2.2.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculation, Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

i i

.HDV151982 SAN ONOFRE-UNIT 3 3/4 11-12 __ ]

RADI0 ACTIVE EFFLUENTS, DOSE - RADI0 IODINES; RADI0 ACTIVE MATERIALS IN PARTICULATE FORM AND TRITIUM I LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from tritium, radiciodines and radioactive materials in particulate form with half-lives greater than 8 days.-in gaseous effluents released, from each reactor unit, from the site-(see Figure 5.1-3) shall be'Ifmited to~the folicwing:

a. During any calendar quarter: Less than or equal t'o 7.5 mrem to any organ and, l

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

c. Less than 0.1% of the limits of 3.11.2.3(a) and (b) as a result of burning contaminated oil.

APPLICA8!LITY: At all times.

ACTION:

a. With the calculated. dose from the release of tritium, radiciodines, and radioactive materials in particulate form, with half Ifves .

greater than 8; days, in gaseous effluents exceeding any of the above limitt,lin lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification'6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions taken to reduce releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with Specification 3.11.2.3.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS ,

4.11.2.3 Dose Cal _culations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the 00CM at least once per 31 days.

hANON0FRE-UNIT 3 3/4 11-13 AMENOMENT NO. 4lj

- __d

c .

en 4 m ,

n >

y ,

P 1

s

[ADI0ACTIVEEFFLUEKTS: ] l

[ GASEOUS'RAQWASTETREATMENT : e.

L

-LIMITING CONDITION FOR OPERATION

3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST: 1

. TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASE0US-o .RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in o.

l . gaseous. waste prior to their discharge when the projected gaseous effluent air; e doses due to gaseous effluent' releases from the site-(see Figure 5.1-3), when, averaged over 31 days, would exceed 0.2 mrad for' gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of 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 (see Figure 5.1-3) when averaged over 31 days

-would exceed 0.3 mrem to any organ.*'

APPLICABILITY: At all times'.

ACTIONi J

a. With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the aboveLlimits, in lieu of any other report required by Specifice.-

tion 6.9.1, prepare and' submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information: ' -'

1. Identification of the inopercble equipment or subsystems and '

the reason for inoperability,

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

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

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ,

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous re7 eases from.the site shall be projected at leitst once per 31 days, in accordance with the ODCM.

4.11.2.4.2 Tte GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the GASEOUS

.RADWASTE TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

I Th ese doses are per reactor unit. NOV 15198L t @AN @NDFRE-UNIT 3 3/4 11-14 ___

t.

3 p 3 }. fl.}~ ~ -lIk h CY Y fRADI0ACTIVEEFFLUENTS

]

3/4.11.3 SOLIO RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be OPERABLE and used, as applicable in accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times.* .

ACTION:

a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Pa-t 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other repor' required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specificae tion 6.9.2 a Special Report which includes the following information:

1. Ida. notification of the inoperable equipment or subsystein and the reason for inoperability,

2. Action (s) taken to restore the inoperable equip.nent to OPERABLE status,

3. A description of the alternative used for SOLIDIFICATION and ,

packaging of radioactive wastes, and

4. Summary description of action (s) taken to prevent a re:urrence. l l

c. Tiie provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS . _ _ _

4.11.3.1 The solid radwaste system shall De demonstrated OPERAl4LE at least once per 92 days by:

a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or

b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

(*See Specification 6.13.1. gy 151982 ]

SAN ONOFRE-UNIT 3 3/4 11-17

2 hi c ,

J.i ,

r

' ),:

4 r '

N

RADI0 ACTIVE EFFLUENTS-

') j SURVEILLANCE' REQUIREMENTS (Continued)' _

4.11.3.2' 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, other than dowatered bead type, evaporator bottoms, boric acid -

solutions, and. sodium' sulfate solutions).

a. ~If any test-specimen fails to verify SOLIDIFICATION, the-SOLIDIFICATION of the batch under test shell be suspended until such time as additional test specimens can be-obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a. subsequent test verifiesLSOLIDIFICA-TION. SOLIDIFICATION of the batch !nay then be resumed using the

. alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste. fails to verify.

' SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection'and testing of representative test specimens from each consecutive batch of the same type of wet waste until .at least.

3 consecutive initial test specimens demonstrate SOLIDIFICATION.

The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste. ,

i.

l l

[ OV15sef N 5 ONO, .c m > su 1 1. , J

1 I ,RADI0 ACTIVE EFFLUENTS ;

i 3/4.11.4 TOTAL DOSE l 1

LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to any member of the public, dug to releases of radioactivity and radiation, from uranium fuel cycle souices shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid.. which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materiais in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2.a. 3.11.1.2.b, 3.11.2.2.a. 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of eceeding the limits of Specification 3.11.4.

This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dese(s) exceeds the limits of Specification 3.11.4, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, !

the Special Report shall include a request for a variance in accordance with the provisicns of 40 CFR 190 and including the specified information of 5 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff l action on the request is complete. The variance only relates to the j limits of 40 CFR 190, and does not apply in any way to the require-ments'for dose limitation of 10 CFR Part 20, as addressed in other i

sections of this technical specification.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1

1 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous ef fluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, id 4.11.2.3, and in accordance with the ODCM.

NOV 15 52 t SAN ONOFRE-UNIT 3 3/4 11-19 ___ ]

/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM i

LIMITING CONDITION FOR OPERATION '

3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

1 APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Com-mission, in the Annual Radiological Operating Report,'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 in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to Specification 6.9.1.13. When more than one of the radionuclides in Table 3.12-2 ara detected in the sampling medium, this report shall be submitted if:

concentration (1) limit level (1) concentration (2) + ***> 1*0 limit level (2) -

When radionuclides other than those in Table 3.12-2 are detected and -

are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equsi to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in i such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With fresh leafy vegetable samples or fleshy vegetable samples unavail- l able from one or more of the rample locations required by Table 3.12-1, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specific-ation 6.9.2, a Special Report which identifies the cause of the unavail-ability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then l be deleted from these required by Table 3.12-1, provided the locations 4 from which the replacement samples were obtained are added to the '

environmental monitoring program as replacement locations. ;

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. l l

Y -

)

SAN ONOFRE-UNIT 3 3/4 12-1 (Iov151 [ l l

j

I4 1:

RADIOLOGICAL' ENVIRONMENTAL MONITORING l SURVEILLANCE REQUIREMENTS ,

i 4.12.1 The radiological environmental-monitoring samples shall be collected

. pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4 Y r.12-1. Y e

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0

_ k TABL .12-1 (Continued)

TABLE NOTATICN

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

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

4.66 s

E -

V -

2.22

Y -

07p(-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above (as picocurie por unit mass or volume),

shis the stan'.fard deviation of the background counting rate or of tne counting rate of a blank sample as appropriate (as counts per minute),

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

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

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

A is the radioactive decay . constant for the particular radionuclides, and At is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).

The value of sbused in the calculation of the LLD for a detection system-shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. in calculating the LLD for a radionuclides determir.ed by gama-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g. , potassium-40 in milk samples).

Typical values of E, V, Y and at shall be used in the cciculations.

In calculating the LLD for a radionuclides determined by gamma-ray spectro-metry, the background should include the typical contributions of other radionuclides normally present in the stanples (e.g. , potassium-40 in milk samples). Typical values of E, V, Y and At should be used in the calculation.

I NOV 151982 gNONOFRE-UNIT 3 3/4 12-9

(:

[ ::

b s TABLE 4.12-l'(Continued)

'l

. TABLE NOTATION l

It should be recognized'that the_LLD is. defined.as an a priori'(before theLfact)-

limit representing the capability of a measurement system and.not as a p_osteriori I; .(after the fact) limit for a particular measurement."

l-

b. LLD for drinking water,

c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1,'shall be identified and reported.

For'a more compiete discussion.of the LLD, and other detection limits,'see the following:

(1): HASL Procedures Manual, HASL-300'(revised annually).

(2) Currie, .L. A. , " Limits for-Qualitative Detection and Quantitative Determin-ation - Application to Radiochemistry" Anal. Chem. 40, 566-93 (1968).

(3)~ Hartwell, J. K., " Detection Limits for Radioisotopic Counting Techniques,"

Atlantic Richfield Hanford Company Report ARH-2537 (June 22,1972). .

La t

y 6v WS8D hANONOFRE-UNIT 3 3/412-10) ,

l RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS j LIMITING CONDITION FOR OPERATION ,

j 3.12.2 A land use census shall be conA cted and shall identify the location j of the nearest milk animal, the nearest residence and the nearest garden

of greater than 500 square feet producing fresh leafy vegetables in each of the { .

16 meteorological sectors within a distance of five miles. For elevated !

releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify the locat10ns of all milk animals and all gardens of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of three miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of any other report required by Specification 6.9.1., prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location (s),

b. With a land use census identify'ing a location (s) which yields a calculeted dose or dose commitment via the same exposure pathway 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, in lieu of any other report required by Specification 6.9.1, pregare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding -

the control station location, having the lowest calculated dose or !

dose commitment via the same exposure pathway may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0'UIREMENTS 4.12.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 using that information which will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

A Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

t@l 151982 SAN ONOFRE-UNIT 3 3/4 12-11 J

9 -

lf '\

il y RADIOLOGICAL ENVIRONMENTAL MONITORING

'3/4.12.3 'INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials. supplied as part of an5Interlaboratory Comparison Program which has been approved by the Commission.

r APPLICABILITY: At all times.

~A CTION:

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.

b. .The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS-4.12.3. A summary of the results obtained as part of the above' required Interlaboratory Comparison Program and in accordance with the ODCM shall be I included in the Annual Radiological Environmental Operating Report.

O k '

(40V 1519@

.hANONOFRE-UNIT 3 3/412-12)

p: '

l

INSTRUMENTATION BASES E

3/4.3.3.7 ' FIRE DETECTION INSTRUMENTATION OPERABILITY-of the fire detection instrumentation ensures that adequate.

warning capability is available for the prompt detection of fires. This' Jcapability is required in order to detect and locate. fires-in their- early stages. Prompt detection of fires will reduce the potential for damage to sefs shutcown and/cr safety-related equipment and is an integrel element in the lL .overall facility fire protection program.

In the event that less than 50% of.the fire detection instrumentation is inoperable in any fire area / zone, the establishment.cf frequent fire pacrols in the affected areas is required to provide detection capability until the y i_noperable instrumentation is restored to OPERABLE. ~

Since the fire detectors are non-seismic, a plant visual' inspection for. '

fires is required within two hours following an earthquake (>0.05g). Since.

safe shutdown systems are protected by seismic Category I barriers, any fire after an earthquake should be detected by this inspection before safe shutdown systems'would be affected. Additionally, to verify tha continued OPERABILITY

-of fire: detection systems after an earthquake -an engineering evaluation of the '

fire detection instrumentation in the required' zones is required to be per-formed within-72 hours following an earthquake.

3/4.3.3.8[RADIDACTIVELIQUIDEFFLUENTINSTRUMENTATION i

..,The radioactive liquid effluent instrumentation is provided to monitor and control, as applic6ble, the releases of radioactive materials in liq,uid 4

effluents during actual or potential releases of liquid effluents. The alare/

trip setpoints for these instruments shall be calculated in accordance with "

the procedures in.the ODCM to' ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. _The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design

'driteria,60,_63 and 64 of Appendix A to 10 CFR Part 50. _ J ceipi nwev w- m- -

3/4.T. 3T. n EADIDAC TIVB GA!EDUS EFFLUENTJ INSTRUMENTATION QptoF@ q., b. V

.The Xadioactivs ga mous affluonn instrumentation is providedJto monitor 3 mna contros, as applicacie, tne re eases or raaicactive materials in gaseous effluents during actual or. potential releases af gaseous affluents. The alare/ trip setpoints for these instruments shall be calculated in accordanca with the precedures in the ODCM tu ansure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. Thisinstrumentationalsoincluded ,

sronisionsp or monitoring anc controlling tne concentrations or potentially exp'osive gas mixtures in the waste gas holoup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General

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

i SAN ONOFRE - UNIT 3 B 3/4 3-4 8dNOMENTNO.53 l

l ,

f

, 3/4.11 RADI'0 ACTIVE EFFLUENTS

' BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1} CONCENTRATION , l 1

F This specification-is provided to ensure that the concentration of ,

I radioactive materials released in liquid waste effluents from the site will be l 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 bodies of water outside .the site will result in exposures within (1) the Section II.A design objectives of Appen-dix I,.10 CFR 50, to an individual, and (2) the limits of 10 CFR 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 l and its MPC in air (submersion) was converted to an equivalent concentra-tion in water using the methods described in International Commission on g Radiological Protection (ICRP) Publication 2. _

j 3/4.11.1.2 k (c 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 } i i Condition for Operation implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid offluents 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 conformance with the guides of Appendix I be shown by calcula-

tional procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents L are consistent with the methodology provided in Regulatory Guide 1.109,

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

Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion'of Effluents frc;n Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

This specification applies +o the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from.the shared system are proportioned among the units Qharingthatsystem. J b

(fgy 151989

~

SAN ONOFRE-UNIT 3 B 3/4 11-1

__ __- -_ ____ - __ A

);

s,

F

.1 J l s

RADI0 ACTIVE EFFLUENTS C' BASES 3/4.11.1.3 A IQUID WASTE TRfATMENT Y d .

C . The OPERABILITY of the liquid radweste treatment system ensures that this ) '.

system will be available for use whenever liquid effluents require treatment prior to release to.the environment. The requirement that the appropriate

portions of this system be used when specified provides assurance that the.

releases of radioactive materials in liquid' effluents will be kept "as low as

-is reasonably. achievable". Thisl specification-implements:thie requirements of

/10 CFR Par' c 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. ~ The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Qart50,forliquideffluents. J 3/411.1.4 LIQUID HO'LDUP TANKS

. Restricting the quantity of radioactive material contained in the specified

. tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, 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.

. 3/4.11.2 GASEOUS EFFL9ENTS f De\ckeb 3/4.11.2.1 ICOSE RATE A -

This specification.is provided to ensure that the dose at any time ht the (ites boundary from gaseous effluents from all units on the site will be within l the annual. dose limits of 10 CFR Part 20 for unrestricted areas. The annual

. dose limits are the doses associated with the concentrations of 10 CFR Part 20,-

. Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside

.the site boundary, to annual average concentrations exceeding the limits sisecified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

For individuals who may at times be within the site boundary, the occupancy of the indiv1 dual 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 i dose iates above background to an individual at or beyond the site boundary to less than or equal to 500 arem/ year to the total body or to less than or equal to 3000 ares / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate abeve background to a child via the inhalation pathway to less than or equal to 1500 mrem / year.

This specification applies to the telease of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units haring that system.

l diDV 1519f L SAN ONOFRE-UNIT 3 B 3/4 11-2

p ' '

3s' t i

}

'RADI0 ACTIVE EFFLUENTS o.

BASES- 1 3/4.11.2.2 IDOSE - NOBLE GASES N

']

f HThis 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 Ol Condition for Operation implements'the guides set.forth in Section II.B of

' Appendix I.- ;The' ACTION statements provide, the required operating flexibility l and'at the.same time implement the guides set forth-in Section IV.A of:

~

Appendix I to' assure that the releases of radioactive" material in gaseous effluents'will be. kept "as low us is reasonably achievat,le". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that _ l. _ _ .

1 conformance with the guides ofLAppendix I be shown.by calculational ~ procedures based on'models-and' data such that'the actual exposure of an individual' y through appropriate pathways ir, unlikely to'be substantially underestimated.

The dose calculations established in-the ODCM for calculating the doses due to the actual release rates of radios.ctive noble gases'in gaseous effluents are consistent with"the methodology provided in Regulatory Guide 1.109,."Calcula-

'cion of Annual Doses to Man from Routine Releases 'of Reactor Effluents for the

. Purpose' of Evaluating Compliarce with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric L Transport and Dispersion of. Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. . The ODCM equations provided for: determining the air doses at the site boundary are based upon the Q istorical average atmospheric conditions.

2/4; ' 1 2. -

A 0 INES, RADI0ACTIV.E_ MATERIALS IN PARTICULATE FORM ]

{ANDFRITILM

.. This specification is provided to implement the requirements of Sections'II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appen-dix I. The-ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Sectior. IV.A of Appendix I to assure that the releases of. radioactive materials in gaseous effluents will

'L .be kept "as low as is-reasonably achievable." The ODCM 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 calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to i be substantially underestimated. The ODCM calculational methods for calcu-lating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, J

i SAN ONOFRE-UNIT 3 B 3/4 11-3 p

, 4 M'

1 5 i i t RADIOACTIVE EFFLUENTS . i BASES.

fCalcu1'ation of Annual Doses to Man from Routine Releases of Reactor Efflu 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 -l-

, from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations 7 also provide for determining the actual' doses. based upon the historical average. t

. atmospheric conditions. The release rate specifications for radiciodines, l' radioactive materiais in particulate form _and tritium are dependent on the existing radionuclides pathways to man,-in the unrestricted area. The pathways which were examined in the development of. these' calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green. , ,

leafy vegetation with subsequent consumption by man,'3) deposition onto grassy

"=

areas where milk: animals and meat producing animals graze 1with consumption of the milk and meat by man, and 4) deposition on the ground'with subsequent Lexposure of man, j 3/4.11.2.4fGASE005RADWASTETREATMENTX F

3 k

.The OPERABILITY of.the GASEQUS RADWASTE TREATMENT SYSTEM and the i VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available 1 for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems r be used, when'specified, provides reasonable-assurance that the releases of H radioactive materials in gaseous effluents will be kept "as low as is

reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to.10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems.

were specified as a suitable fraction of the dose design objectives set forth Q Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations-from reaching these flammability limits. These automatic control features include injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of

. radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

I i

~

SAN ONOFRE-UNIT 3 B 3/4 11-4 l

h:$

N$l g <

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m Qii , l m

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q , RADI0 ACTIVE EFFLUENTS. I w

BASES I

3/4.11.2.6 GAS STORAGE TANKS-

.. Restricting the quantity'of radioactivity contained in each gas storage ,

tank provides assurance = that in the' event of-an uncontrolled release of the tank's contents, the resulting total body exposiJre to an individual at the '

nearest exclusion area' boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure".

4.1'1.3 : SOLID RADI0 ACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system'will.

be available for use whenever sclid radwastes require processing and packaging' e prior to.being shipped offsite. This specification implements the requirements- J ofr10 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 constituents,; mixing and curing times.

3/4.11.4 TOTAL DOSE-This specification is provided to meet the' dose limitations of 40.CFR 190.

The specification requires the preparation and. submittal of.a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I .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 190 if the individual reactors remain within the reporting requirement-level. The Special Report will describe a course of action which should result in the limitation of dose to a member of the public for:12 consecutive months to witbin the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to L' 'the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle

'acilities at the same site or within a radius of 5 miles must be considered.

If the dose to any member'of the public is estimated to exceed the require-ments of 40 CFR 190, the Special Report with a request for a variance in accordance with the provisions of 40 CFR 190.11, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is

' completed provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected. An individual is not considered a member of

the public during any period in which he/she is engaged in carring out any Qerationwhichispartofthenuclearfuelcycle. ]

JL-Gov i5 m3 8 3/4 11-5 5AN ONOFRE-UNIT 3

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/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l l i

BASES 3/4.12.1 MONITORING DROGRAM The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the st& tion operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental 4 exposure pathways. The initially specified monitoring program will be '

effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational i experience. '

The detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit

~

representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyses shall be per-formed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample ,

sizes, the presence of interfering nuclides, or other uncontrollable circum-stances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological ,

Environmental Operating Report.

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting with local agricultural ~1 authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to !

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

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-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that' independent checks on the. precision and; accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environ-

'; . , mental monitoring in order to demonstrate that the results are reasonably ut'*- _

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' * ' b.. In-Plant Radieticn Monitorino T '

.A program which will. ensure the capability to accurately determine L the-airborne iodine concentration in vital: areas under accident L conditions. Th'is program shall. include the followihg:

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-(i) Training of personnel, s: (ii) Procedures for monitoring, and .

R, (iii)' Provisions for maintenance of sampling and analysis-equipment.-

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c. Secondary Water Chemistry .

Ly- . A program for. monitoring of secondary water chemistry to inhibit k

steam generator tube degradation. This program shall include:

(i) ' Identification of a sampling. schedule for the critical variables -

and; control points for these variables,

.(ii) Identification'of the procedures used to' measure the values 'of the critical variables,

.(iii) Identification of1 process sampling _ points, including monitoring 1 the discharge of'the condensate pumps for evidence of condenser in-leakage,

.(iv) Procedures.for the recording and management of data, (v)- Procedures defining. corrective actions for all off-control

, point' chemistry conditions, and l

-(vi) A procedure identifying (a) the authority responsible f,or the -

interpretation of the data, and (b) the sequence and' timing of administrative events required to initiate corrective action.

d. Post-Accident Samolino A program

which will ensure the capability to obtain and analyze

, reactor coolant, radioactive iodines and particulate in plant

l. gaseous effluents, and containment atmosphere samples-under accident conditions. The program

shall include the training of personnel, the procedures for sampling and cnalysis and the provisions for

$ maintenance of sampling and analysis equipment.

. .'9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES -

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

NRC Regional. Administrator unless otherwise noted.

. *Not required to be implemented until September 1, 1983.

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A program sh'all'be provided. conforming with 10 CFR 50.36a for the contro b f radioactive effluents and.for maintaining the doses to

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, MEMBERS achievable. OF THE PUBt.IC.from radioactive. effluents as low as reasonaoly '

The program (1) shall be contained in.the 00CM, (2)-shall be implemented by operating procedures, and (3) shall in-clude remedial actions to be taken whenever'the program limits are exceeded. The program shall include the-following elements:

1) Limitations on. the operability of radioactive liouid and gaseous monitoring' instrumentation including survei?. lance' tests and set-point determination in accordance with the methodology in tha ODCM; 2)

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

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

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER'0F THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conform- ] 4 ing to Appendix'I to 10 CFR Part 50',

5)- Determination of cumulative and projected dose contributions

.from ' radioactive effluents for the current calendar quarter and ye sf gch dal , current calendar year $in accordance with the methodology and parameters in the ODCM at least every 31 days;

6) Limitations on the operability and use of the liquid and gaseous

effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radio-activity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conformina to Appendix I to 10 CFR Part 50; 7)- Limitations on tne dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II. Column 1

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i 8)- Limitations on the annual and quarterly air doses resulting from .

. noble gases ~ released in gaseous effluents from each unit to areas beyond the,JME ROUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the' annual and quarterly doses to a MEMBER OF -

THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives' greater than 8 days in gaseous effluents released from each unit to areas'beyond the

.iLTE_AQHh0ARY confarmina to Appendix I. to 10 CFR Part 50,

10) L1mitations'on tne anniTal dose or dose commitment to any MEMBER 0F THE~PUBLIC due-to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

I Radiological Environmental Monitoring Program A program shall t',e provided to monitor the radiation and radio-nuclides in the environs of-the plant. The program shall provide (1) representative measurements of radioactivity in the' highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental expo ~

sure' pathways. The prograi shall (1) be contained in the 00CM, (2)' conform to the guidance of Appendix I to 10 CFR Part 50, and (3), include the following:

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

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

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance pro- ;

gram for environmental monitoring.

ADMINISTRATIVE CONTROLS _

y Uhnu L55 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

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6.9.1.6[Routineradiologicalenvironmentaloperatingreportscoveringthe Foperation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to j mMay 1 of the year following initial criticality. J 6.9 M he annual radiological environmental operating reports shall include l p"radiologicalenvironmentalsurveillanceactivitiesforthereportperiod,ummaries, interpre including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment 7 of the observed impacts of the plant operation on the environment. The

,% reports shall also include the results of land use censuses required by I Specification 3.12.2. If harmful effects 'or evidence of irreversible damage (g4 are detected by the monitoring, the report shall provide an analysis of the N problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatury Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the -

reasons for the missing results. The missing data shall be submitted as scon as possible in a supplementary report.

The reports chall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the

] results of licensee participation in the Interlaboratory Comparison Program, j trec,uired by Specification 3.12.3. j SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT *

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6.9.3.'[d Routine radioactive effluent release reports DveTing the operationl fof the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the (first report shall begin with the date of initial criticality. ;

7 1

A single submittal may foe rude for s multiple unit station. The submittal should combine those sections that are common to all units at the *,tation; however, for units with ceparate radwaste systems, the submittal shall specify the releases of radioactive material from sach unit.

1 I

, SAN ONOFRE-UNIT 3 6-18

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N INTERT 4 The Annual Radiological Environmental Operating Report covering the -

4 operation of the unit during the previous caltadar yesr shall be submittyd before May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the resbits of the Radiological Environmental Monitoring Program for the reporting period. Th2 material provided shall be

, consistent with the objectives outlined in (1)'the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

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. The $nmiannual Radioactive Effluent Release Report covering the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall in-clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released fica the unit. The material provided shall be (1) consistent with the objectives outlined in the 00CM and PCP and (2) in con-formance with 10 CFR'50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

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ADMINISTRATIVE CONTROLS ,

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W l 6.9 he radioactive eff'.uent release reports shall include a summary of f"EFe quantities of radioactive liquid and gaseous effluents and solid waste f released from the unit as outlined in Regulatory Guide 1.21, "Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of ,

4 Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled j Nuclear Power Plants," Revision 1, June 1974, with data summarized on a i quarterly basis fo? lowing the format of Appendix B thereof. j The radioactive effluent release report to be submitted 60 days after January 1 of each year shall include an annual summary of hourly meteorological i data lollected over 4he previous year. This annual summary may be either in l the form of an hour-by-hour listing of wind speed, wind direction, and l atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of Jtability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseon effluents released from the unit or station during the previous calendar year. This same report j shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figures 5.1-3 and 5.1-4) during the report period.

All assumptions used in making these assessments (i.e. , specific activity, exposure time and location) shall be included in these reports. The meteoro-logic.al conditions concurrent with the time of release cf radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

l The radioactive effluent release report to be submitted 60 days after i January 1 of each year shall also include an assessment of radiation doses to f th@ likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose j contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

The radioactive effluents release shall include the following information for each type of solid waste shipped offsite during the report period:

a. Container volume, l b. Total curie quantity (specify whether determined by measurement or estimate),

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

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

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

f. Solidification agent (e.g., cement, urea formaldehyde),

6-19 wm s a SAN ONOFRE-UNIT 3

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ADMINISTRATIVE CONTROLS -

,.be1radioactiveaffluentrelease'.reportsshallincludeunplannedreleasesf

.p the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on'a quarterly basis.

The radioactive effluent release reports shall include any changes to.the JROCESSCONTROLPROGRAM(PCP)madeduringthereportingperiod.

M'ONTHLY' OPERATING REPORT

6.9.1.10 Routine reports .of operating statistics and shutdown experience,

. . including documentation of all challenges to the safety valves, shall be submitted on s'eonthly basis to the Director, 0ffice.of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a. copy to the Regional < Administrator of the Regional Office of the.NRC, no later than the-H '15th oi each month following the calendar month covered by the report.

'Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the

Monthly Operating Report'within 90 days in.which the. change (s) was_made effec-

<tive. In addition,'a: report of any major changes to the radioactive-waste treatment systems'shall be. submitted with the Monthly Operating Report for the period-in which the evaluation was' reviewed and-accepted in accordance with 6.5.2.

REPORTdBLE' OCCURRENCES-y 6.9.1.11 The REPORTABLE' OCCURRENCES of Specifications 6.9.1.12 and 6.9.1.13 below including corrective actions and measures to prevent recurrence, 'shall be reported to the NRC. 5' supplemental' reports may be' required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a. licensee event report shall be completed and reference shall be made to the original report date.

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.12 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephtne and confirmed by telegraph, mailgram, or facsimile transmission t6

'the Regional Administrator'of the Regional Office or his designate no later than the first working day following the event, with a written followup report '

within 14 days. The written followup 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 material to provide complete explanation of the circumstances surrounding the event.

-a. Feilure 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 reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.

l SEP 2119M SAN ONOFRE-UNIT 3 6-20 (AMENDMENTNO.14

i

ADMINISTRATIVE CONTROLS-

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA Manual.

J. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the OSRC and the NSG.

1. Records of the service lives of all snutbers within the scope of Technical Specification 3/4.7.6 including the date at which the ser-vice life commences and associated installation and maintenance records.

m. Records of secondary water sampling and water quality.

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6. 1 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and

adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 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 greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation ,

Exposure Permit (REP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

l l 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 knowledgeable of them. l

%calth Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the REP issuance requirement during the performance of their assigned radiation prctection duties, provided they are otherwise following approved plant radiation protection procedures for entry into high radiation areas. _

kEP 2 41985 SAN ONOFRE-UNIT 3 6-24 (AMENDMENT NO. 22

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

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'< c. . An individual' qualified in radiation protection procedure:: who'.is.

- equipped with a . radiation dose rate monitoring device who is~

responsible'for providing positive control over the activities l . within.the' area and shall. perform periodic radiation l surveillance'at L the frequency specified by the facility Health Physicist in the L Radiation Exposure' Permit.

6.12.2. In' addition to the requirements of 6.12.1, areas accessible:to personnel D . with radiation.levelsLWuch that a major portion of the body could receive in.

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one. hour a.' dose greater than 1000 mrem shall be'provided with-locked. doors!to w .

prevent unauthorized entry, and the keys shall. be maintained under the adminis- /

trative control of'the Shift Supervisor on. duty and/or health physics supervision.

' Doors sha11' remain locked except.during periods'o.f access.by personnel.under an approved. REP which shall,specify the dose rate levels in the:immediate work area and the maximum allowable stay time for. individuals in that area. For..

individual. areas accessible to personnel with. radiation-levels such that a major >

. portion of the. body could receive in one hour a dose in excess of 1000' mrem **

'that'are located within large areas, such as.PWR. containment,.where no enclosure exists for purposes. of locking, and no enclosure can' be reasonably constructed around the' individual aress, then that area'shall be roped off, conspicuously-posted and a flashing light shall.be activated as a warning' device. In lieu .

of the stay time specification.of the REP, direct or remote (such as use of

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closed circuit TV' cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the ares.

6.13' PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.#

6.13.2 Licensee initiated changes to the PCP:

1. Thall be submitted to the Commission in the semi-antiual Radioactive' T A Effluent Release. Report for the period in which the change (s) was .

y <made. This submittal: shall contain: - j

a. Sufficiently detailed information to totally support the rationale' '

N:w h s for the change without benefit of additional or supplemental

\information; a

. b. FA determination that the change did not reduce the overall Nk conforinance of the solidified waste product to existing criteria 3 4q $ for solid wastes 1 and ;

c. Documentation of the fact that the change has been reviewed and found acceptable pursuant to 6.5.2. ,

'2. Shall become effective upon review and acceptance pursuant to 6.5.2. !

-** Measurement made at 18" from source of radioactivity.

The PCP shall be submitted and approved prior to shipment of " wet"_ solid , i a ~_ __

radioactive waste.

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1 Shall be documented and records of reviews performed shall be retain- l ed as required by Specification 6.10.'Lo,

', contain:

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Sufficient information to support the change together with the .!

appropriateanalysesorevaluationsjustifyingthechange(s);

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A determination that the change will maintain the overall con-formance of the solidified waste product to existing require-ments of Federal, State, or other applicable regulations.

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ADMINISTRATIVE CONTROLS 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6,14.2 Licensee initiated changes to the ODCH:

0 hl. kha11besubmittedtotheConaissionintheMonthlyOperatingReporN :

34 hy within 90 days of the date the change (s) was made effective. This

-r_n w t it qubmittal shall contain:

a. hficientlydetailedinformationtototallysupportthe b !

rationale for the change without benefit of additional or supplemental information. Information submitted should consist 7[6 gg y t,

0 of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, M together with appropriate analyses or evaluations justifying j

1he change (s);

bc b. IA determination that the change will not reduce the accuracy orr b '

w4k reliability of dose calculations or setpoint determinations; A*4 3> and j .

c. Documentation of the fact that the change has been .eviewed and found acceptable pursuant to 6.5.2. p M+QQ 4i ShallbecomeeffectiveuponreviewandacceptancegbytheO p 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gasoous and D solid) 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

1. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was performed pursuant to 6.5.2.

The discussion of each change shall contain:

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

b. Sufficient detailed information to totally support the reason for the change without benefit of additiondi or supplemental information;

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

d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously

( predicted in the license application and amendments thereto; 1 SEP 211$4 s e @me n t- e v s @-86 [AMENDMFNT d

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s Shall'be' documented and records of reviews performed shall be. retain-ed as required by Specification 6.10.2.r),

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l luseg u Sufficient information? to' support the change together with the appropriate analyses or evaluations justifying the change (s).,

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A determination that; the change will maintain the level of

-radioactive effluent control required by 10 CFR 20.106, 40'CFR

'Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and

'not: adversely impact the accuracy or reliability of effluent, dose, or setpoint' calculations.

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I ), Shall be submitt.ed to the Ccomission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to tha 00CM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change wts implemented.

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hDMINISTRATIVECONTROLS ee- '.An evafuation'of-the change which shows the expected maximum exposures to individual'in the unrestricted area and to the general population:that differ from those previously estimated

] in the license application and amendments thereto; f.. A comparison of the predicted releases of radioactive

. materials, 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 reviewe'd and.

found acceptable pursuant'to 6.5.'2.

L 2. Shall-become effective.upon review and acceptance pursuant to 6.5.2 l

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[AN-ONOFRE-UNIT 3 6-27 AMENDMENT NO. 14 J

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