ML20043D855
| ML20043D855 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 05/29/1990 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20043D812 | List: |
| References | |
| GL-89-01, GL-89-1, NUDOCS 9006110230 | |
| Download: ML20043D855 (521) | |
Text
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1 t INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.22 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 1.23 MINIMUM CRITICAL POWER RATI0..................................... 1-4 1.24 0FFSITE DOSE CALCULATION MANUAL.................................. 1-4 1.25 OPERABLE - OPERABILITY........................................... 1-4 1.26 OPERATIONAL CONDITION - CONDIT!0N................................ 1-4 1.27 PHYSICS TESTS.................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE........................................ 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY.................................... 1-5 1.30 PROCESS CONTROL PR0 GRAM.......................................... 1-5 1.31 PURGE - PURGING.................................................. 1-5 1.32 RATED THERMAL P0WER.............................................. i 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY................ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME.......................... 1-6 i l 1.3b REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY.................. 1-6 1 1.36 REPORTABLE EVENT................................................. 1-7 1.37 ROD DENSITY...................................................... 1-7 1.38 SHUTDOWN MARGIN.................................................. 1-7 1.39 SITE B0VNDARY.................................................... 1-7 1.40 (Deleted)........................................................ 1-7 1.41 SOURCE CHECK..................................................... 1-7 1.42 STAGGERED TEST BASIS............................................. 1-8 1.43 THERMAL P0WER.................................................... 1-8 1.44 UNIDENTIFIED LEAKAGE............................................. 1-8 LlHERICK - UNIT 1 11
_INDEX LIMITING CONDITIONS FOR OPERATION AND SURVE1LLANCE REQUIREMENTS SECTION PAGE 't INSTRUMENTATION (Continued) Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements ........................ 3/4 3-66 Seismic Monitoring Instrumentation ........................ 3/4 3-68 : Table 3.3.7.2-1 Seismic Monitoring Instrumentation ..................... 3/4 3-69 Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirements ........................ 3/4 3-71 The information from pages 3/4 3-73 through 3/4 3-75 has been intentionally omitted. Refer to note on page 3/4 3-73 ......... 3/4 3-73 Remote Shutdown System Instrumentation and Controls ....... 3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls......... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance Requirements ........................ 3/4 3-83 Accident Monitoring Instrumentation ....................... 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation .............................. 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumentation Surveillance Requirements ........... 3/4 3-87 Source Range Monitors ..................................... 3/4 3-88 Traversing in-Core Probe System ........................... 3/4 3-89 Chlorine Detection System ................................. 3/4 3-90 Toxic Gas Detection System ................................ 3/4 3-91 Fire Detection Instrumentation ............................ 3/4 3-92 LIMERICK - UNIT 1 ix
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE _ REQUIREMENTS ifC110N PAGE
. INSTRUMENTATION (Continued)
Table 3.3.7.9-1 Fire Detection Instrumentation ...... 3/4 3-93 Loose-Part Detection _ System ............................... 3/4 3-97
.The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted.. Refer to note on page 3/4 3-98 ............. 3/4 3-98 Off9as Monitoring Instrumentation ......................... 3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation ......... 3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements .......... 3/4 3-107
;3/G.3.8 :. TURBINE OVERSPEED PROTECTION SYSTEM .................. 3/4 3-110 3/4.3.9 FEEDWATER/MA1H-TURBINE TRIP SYSTEM ACTUATION INSTRUMENTAlTON ....................................... 3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation ...... 3/4 3-113
. Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation In'strumentation Setpoints ........... . ................. 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation Surveillance Requirements ........... 3/4 3-115
~3/4'.4 REACTOR COOLANT SYSTEM 3/d.4.1- RECIRCULATION SYSTEM Recirculation Loops ................................... 3/4 4 )
LIMERICK - UNIT 1 x
um. INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f4GE REFUELING _ OPERATIONS (continued) 3/4.9.10 CONTROL R00 REMOVAL Single Control Rod Remova1.................................. 3/4 9-13
's Multiple Control Rod Removal................................ 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1..........................................., 3/4 9 17-Low Water Leve'i............................................. 3/4 9-18 I
3/4.10 SPECIAL TEST EXCEPTIONS-3/4.10.1 PRIMARY CONTAINMENT INTEGRITY............................... 3/4 10-1 3/4.10.2 R00 WORTH MINIMlZER.......................................... 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS............................. 3/4 10-3 3/4.10.4 RECIRCULATION L00PS......................................... 3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION........................................ 3/4 10-5 - 3/4.10.6 TRAINING STARTUPS........................................... 3/4 10 6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information from pages 3/4 11-1 through 3/4 11-6 has been intentionally omitted. Refer to note on page 3/4 11-1 ................... 3/4 11-1 Liquid Holdup Tanks......................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS The information from pages 3/4 11-8 through 3/4 11-14 has been intentionally omitted. Refer to note on page 3/4 11-8 ................... 3/4 11-8 I UhERlrk - UNIT 1 xvi
INDEX LIMITING CONDITIONS'FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADIOACTIVE EFFLUENTS (Continued) Explosive Gas Mixture...................................... 3/4 11-15 Main Concenser.............................,............... 3/4.11-16 The information on page 3/4 11-17 has been intentionally omitted. Refer to note on this page................................................. 3/4 11-17 '3/4.11.3- (Deleted) The information on pages 3/4 11-18 through 3/4 11-20 has been intentionally omitted. Refer to note on page 3/4 11-18. 3/4.11.4 (0eleted)................................................ 3/4 11-18 3/4.12 (Deleted) The information on pages 3/4 12-1 through 3/4 12-14 has been intentionally omitted, refer to note on page 3/4 12-1................... 3/4 12-1 LIMERICK - UNIT 1 xvii
1
-A o - -.- > j INDEX
' BASES SECTION PAGE-l_NSTRUMENTATION (Continued)
Seismic Monitoring Instrumentation.......................... B 3/4 3-4 (Deleted) .................................................. B 3/4 3-4 Remote. Shutdown System Instrumentation and Controls......... B 3/4 3-5 > Accident Monitoring Instrumentation............... ,........ B 3/4'3-5 Source Range Monitors....................................... B 3/4 3-5 . Traversing in-Core Probe System............................. B 3/4 3-5 . Chlorine and Toxic Gas Detection Systems.................... B 3/4 3-6
. Fire Detection Instrumentation.............................. B 3/4 3-6 s
Loose-Part Detection System................................. B 3/4 3-6 . (Deleted) .................................................. B 3/4 3-6 Offgas Monitoring Instrumentation .......................... B 3/4 3-7 e 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM......................... B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION ( l' . I N S T RUM E N TAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : B 3/4 3 Bases Figure B 3/4.3-1 Reactor Vessel Water-Level........................... B'3/4 3-8 3/4.4 - REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM........................................ B 3/4 4-1 t-l 3/4. 4. 2 S AF ETY /P' t I E F VALVE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-2 l 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................... B 3/4 4-3 L Operational Leakage......................................... B 3/4 4-3 3/4.4.4 CHEMISTRY................................................... B 3/4 4-3 LIMERICK - UNIT 1 xix
4 INDEX BASES ( < SECTION PAGE ~ REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATF0RM.................................... 8 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L................ B 3/4 9-2 {- 3/4.9.B and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L............. B 3/4 9-2 3/4.9.10 CONTROL ROD REM 0 VAL................................... B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION......... B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY......................... B 3/4'10-1 3/4.10.2 ROD WORTH MINIMlZER.................................... B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS........................ B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS................................... B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.................................. B 3/4 10-1 b 3/4.10.6 TRAINING STARTUPS..................................... B-3/4 10-1
' 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS E The information on page B 3/4 11-1 has been intentionally omitted. Refer to note on this page.... B 3/4 11-1 (Deleted) ............................................ B 3/4 11-2 m
= Liquid Holdup Tanks................................... B 3/4 11-2 = 3/4.11.2 GASE0US EFFLUENTS (Deleted) ............................................ 8 3/4 11-2 The information on page B 3/4 11-3 has been = intentionally omitted. Refer to note on this page.... B 3/4 11-3 _ (Deleted) ............................................ 8 3/4 11-4 LIMERICK - UNIT 1 xxii
I LNUEX BASES SECTION PAGE ! RADI0 ACTIVE EFFLUENTS (Continued) Explosive Gas Mixture................................. B 3/4 11-4 1 I Main Condenser........................................ B 3/4 11-5_
-(Deleted) .. ......................................... B 3/4 11-5
'3/4.11.~3-(Deleted) ............................................ B 3/4 11-5 -a 3/4.11.4L (Deleted) ............................................ B 3/4 11-5 j i
3/4.12 (Deleted) The information from pages B 3/4 12-1 ) through B 3/4'12-2 has been intentionally omitted. Refer to note on page B 3/4 12-1 ........... B 3/4 12-1 i l 1 i 1 1 I i 1. i l l LIMERICK - UNIT 1 xxiii D
-lNDEX r 'bESIGN FEATURES t
f SECTION PAGE 5.1 SITE Exclusion Area.................................................. 5-1 Figure 5.1.1-1 Exclusion Area........................ 5-2.
~ Low Population Zone............................................. 5 Figure 5.1.2-1 Low Population Zone.................... 5-3
-(Deleted)....................................................... 5-1 The figures on pages 5-4 through 5-6 <
have been intentionally omitted.- Refer to note on page 5-4................................... 5-4 r (0eleted)........................................................ 5-1 5.2 CONTAINMENT i
' Configuration................................................... 5-1 Design Temperature and Pressure................................. 5-1 ;
Secondary Containment........................................... 5-7 5.3 REACTOR CORE - Fuel Assemblies................................................ 5 o Contro! Rod Assemblies......................................... 5-7 5.4 REACTOR COOLANT SYSTEM
-Design Pressure and Temperature................................ 5-7 Volume.........................................................
5-8 5.5 FUEL STORAGE u iticality.................................................... 5-8 LIMERICK - UNIT 1 xxiv
INDEX ! t ADMINISTRATIVE CONTROLS
~SECTION- PAGE i s 6.13' PROCESS CONTROL PR0GRAMJ P,Cfl................................ 6-21 1
6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................... 6-22 6.15(0eleted)..................................................... 6-22 l r r , F k-
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-LIMERICK - UNIT 1 xxviii w
DEFINIT 10NS LOGIC' SYSTEM FUNCTIONAL TEST-1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested. MAXIMUM FRACTION OF LIMITlHG POWER DENSITY 1.21 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest value of the FLPD which exists in the core. MEMBER (S) 0F THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category-does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant. MIN' MUM CRITICAL POWER RATIO T.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE 00SE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The 00CM shall also contain (1) the Radioactive Effluent Controls r.nd Radiological Environmental Monitoring Programs 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 Radioacti/e Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8. OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE 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 required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2. PHYSICS TESTS 1.27 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 of the FSAR, (2) aut%rized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. . LIMERICK - UNIT 1 1-4
DEFINITIONS-
~
PRESSURE B0UNDARY LEAKAGE-1.28 PRESSURE B0VHDARY LEAKAGE shall be leakage through a nonisclable fault in a reactor coolant system component body, pipe wall or vessel wall. PRIMARYCONTAINMENTINTEGRIH 1,29 PRIMARY CONTAINMENT INTEGRITY shall exist when:
- a. All primary containment penetrations required to be closed during.
accident conditions are either:
- 1. Capable of t.emg closed by an OPERABLE primary containment automatic isolation tr u em, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided
-in Table 3.6.3-1 of Specification 3.6.3.
- b. All primary containment equipment hatches are closed and sealed..
- c. The primary containment air lock is in compliance with.the requirements of Specification 3.6.1.3.
- d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.
- e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
- f. The sealing mechanism associated with each primary containment-penetration; e.g., welds, bellows, or 0-rings, is OPERABLE.
PROCESS CONTROL-PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP)_shall contain the provisions.to assure that the solidification or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for transportation to
-the disposal site and receipt at the disposal site. With solidification or dewatering, the PCP shall identify the process parameters influencing solidification or dewatering, based on laboratory scale and full scale testing or experience.
PURGE - PURGING
' 1.'31 PURGE or PURGING shall be 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 oas is required to purify the confinement.
.' LIMERICK - UNIT 1 1-5 I
DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT-INTEGRITY (Continued)- [, b.- All-refueling floor secondary containment hatches and blowout panels-are closed and sealed,
- c. The-standby gas treatment system is in compliance with the requirements
-of Specification 3.6.5.3. '
- d. At least one door in each access to the refueling floor secondary containment is clon9d.. !
i
- e. .The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
- f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a. ,
REPORTABLE EVENT 1.36 A REPORTABLE EVENT-shall be any of those. conditions specified in Section 50.73 to 10 CFR Part 50. - i
-R00 DENSITY ! -1,37 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% R00 DENSITY.
SHUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown- ' condition; cold, i.e. 68 F; 0 and xenon free. SITE BOUNDARY 1.39 The SITE B0UNDARY shall be that line as defined in the 00CM, figure I.2.2-la. - 1.40- (Deleted) SOURCE CHECK-1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the! channel sensor is exposed tc a radioactive source. i LIMERICK - UNIT 1 1-7
- ,f, 4 i Section3.3.7.31(Deleted)
I I i THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE 00CM. . TECHNICAL SPECIFICATIONS .; PAGES 3/4 3.~/4 (HROUGH 3/4 3-75 0F THIS SECTION GVE BEEN INTENTIONALLY OMITTED. t' I s C LIMERICK - UNIT 1 3/4 3-73
. ..q Section3.3.7.11(Deleted)'
i .3.. THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS-PAGES 3/4 3-99 THROUGH 3/4 3-102 0F THIS SECTION HAVE BEEN INTENTIONALLY OMITTED. I I i I l 1 i i. If m I i (
/ ' RICK '- Ut'!T 1 3 '* b98
-INSTRUMENTATIONL 10FFGAS GAS MONITORING INSTRUMENTATION 1 LIMITING CONDITION FOR OPERATION 3.3.7.12 The offgas monitoring instrumentation ;nannels shown in Table 3.3.7.12-l'shall be OPERABLE with their alarm / trip setpoints set ~to ensure that the limits of Specifications 3.11.2.5 and 3.11.2.6 respectively, are not-exceeded.
APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:
- a. With an offgas monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.12-1.
- b. With:less-than the minimum number of offgas monitoring instrumentation l channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.- Restore the-inoperable. instrumentation.to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Semiannual Radioactive Effluent Release Report. ,
- c. The provisions of Specifications 3.0.3 and 3.0.4 are n.. applicable. .
1 SURVEILLANCE-REQUIREMENTS y i s 4.3.7.'12 Each offgas 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 l
' 4.'3. 7.12- 1.
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. ja OFFGAS MONITORING INSTRUMENTATION-E 7:
MINIMUM CHANNELS
& INSTRUMENT OPERABLE APPLICABILITY ACTION U
- 1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Hydrogen Monitor 1 ** 110-
- 2. (Deleted)
- 3. (Deleted) w 4. MAIN CONDENSER OFFGAS PRE-TREATMENT ^
2 RADI0 ACTIVITY MONITOR
$ a. Noble Gas Activity Monitor 1 ** 115 E
- 5. (Deleted)
- . . . z;.: . - :- . - . . .= . . . . - - . - . . . .
TABLE 3.3.7.12-1 (Continued) TABLE NOTATIONS (Deleted)_ l
** During_ operation of the main condenser steam jet air ejector and offgas treatment system.
***(Deleted)
ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue for up to 30 days provided
- grab samples are collected at least once per 4 hours and analyzed within the following 4 hours.
ACTION 111-114 (Deleted) ACTION 115 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, releases to the environment may continue for up to 72 hours provided that the North Stack Effluent Noble Gas Activity Monitor is OPERABLE; otherwise, be in at least HOT SHUT 00WN within 12 hours. i LIMERICK - UNIT 1 3/4 3 '05
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- LIMERICK - UNIT 1 3/4 3-106 i
d,fI ii I TABLE 4.3.7.12-1
- .OFFGAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS si CHANNEL. MODES IN WHICH; k
3 SOURCE CHANNEL FUNCTIONAL SURVEILLANCE CHANNEL
' INSTRUMENT ~ CHECK CHECK CALIBRATION TEST 'IS REQUIRED-E
- 3
- 1. MAIN CONDENSER OFFGAS TREATr1ENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM M **
I a. Hydrogen Monitor D N.A. Q(3)
- 2. (Deleted) l l 3. (Deleted) m 4. MAIN CONDENSER OFFGAS y PRE-TREATMENT RADIOACTIVITY v MONITOR (STEAM JET AIR
,8 EJECTOR)
- a. Noble gas activity monitor D M- R(2) Q(1)
- 5. (Deleted) 1
TABLE 4.3c7.12-1.(Continued) c. TABLE NOTATIONS
* (Deleted)
** During operation of the main condenser steam jet air ejector and offgas treatment system.
(Deleted) (1)- 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 / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(2)_ The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute-of Standards and Technology (NIST), previously National Bureau of Standards, or.using standards that have been obtained from suppliers that participate in measurement assurance activities with HIST. -These standards shall permit-calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CAllBRATION, sources that have been related to the initial calibration shall be used. (3) The CHANNEL CAllBRATION shall include the use of standard gas samples containing a nominal:
- 1. 0.0 volume percent hydrogen, balance nitrogen, and
~
- 2. 4 volume percent hydrogen, balance nitrogen.
'(4) . -(Deleted)
LIMERICK - UNIT 1 3/4 3-108
6 9 INTENTIONALLY LEFT BLANK LIMERICK - UNIT'1 3/4 3-109 i
.t eSe'ction.3/4'11.1.1'through 3/411,1.4(Deleted)~
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t j 'l \- h THE INFORMATION FROM THESE TECHNICAL- L SPECIFICATIONS SECTIONS HAS BEEN l RELOCATED TO THE 00CM. TECHNICAL-
. SPECIFICATIONS PAGES-3/4 11-2 THROUGH 3/4 3 11-6 0F THESE SECTIONS HAVE BEEN L INTENTIONALLY OMITTED.
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- Section 3/4J11.2.1 through Section-3/4 11.2.4 (Deleted)
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1 THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE ODCM. TECHNICAL , SPECIFICATIONS PAGES 3/4-11-9 THROUGH 3/4 11-14'0F-THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED. t-( H i
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LIMERICK'- UNIT 1 3/4 11-8
,Secticn 3/411-2.7:(Deleted).
I THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED
.TO THE 00CM, s
1 g'j-i t-' h A\h LIMERICK - UNIT l' 3/4 11-17
Section 3/4 11.3 and Section 3/4'11.4'(Deleted) s V
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I a THE INFORMATION FROM THESE TECHNICAL -3 SPECIFICATIONS SECTIONS HAS BEEN , RELOCATED TO THE PCP. TECHNICAL
= SPECIFICATIONS PAGES 3/4 11-19 THROUGH 3/4 11-20 0F'THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
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'Section 3/4. R (Deltted) .,
ri i l p 3. THE'INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED .! TO THE ODCM. TECHNICAL SPECIFICATIONS 'I PAGES 3/4 12-2 THROUGH 3/4 12-14 0F THIS i' SECTION HAVE BEEN INTENTIONALLY OMITTED. l . l: , 1: ! I l l 1
-LIMERICK - UNIT 1 3/4 12-1
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I iIIk bIa b 'dl Wk' ' INSTRUMENTATION
. BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor-isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of-the emergency core cooling equipment.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for. instrument drift specifically allocated for each trip in the safety
- analyses.
3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION The control rod block funct. ions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section-3/4.2' Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block. Operation with a trip set less conservative than its Trip Setpoint but-witnin its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures'that; (1) the radiation levels are continually measured in the areas served by'the individual channels, and-(2) the alarm or automatic action is initiated when the radiation level trip setpoint'is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables
. following an' accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.
' 3/4;3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of-those features important to safety.
This capability is required to permit comparison of the measured response to that used in the design basis for the unit. 3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM. LIMERICK - UNIT 1 B 3/4 3-4
INSTRUMENTATION-BASES' 3/4.3.7.4 REMOTE SHUTDOW SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and maintenance of . HOT SHUTDOWN of the unit from locations outside of the control room. This capability ' is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix A. 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the-accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled
'Huclear Power Plants to Assess Plant Conditions During and Following an Accident,"
December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. 3/4.3.7.6 SOURCE RANGE MONITORS The source range' monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. 'At these power _ levels, reactivity additions shall not be made without this flux level , information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted. 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM ' The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor Core. The TIP system OPERABILITY is demonstrated by normalizing all pr'obes (i.e., detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1). LIMERICK - UNIT 1 B 3/4 3-5
INSTRUMENTATION BASES 3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlorine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will. manually be placed in the chlorine isolation mode of operation to provide the required protection. The detection systems required by this specification are consistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975. 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extinguishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification. Consequently, the minimum number of OPERABLE fire detectors must be greater. l The loss o? detect on capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133 " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE 00CM. i LIMERICK - UNIT 1 B 3/4 3-6
[NFRUMENTATION BASES 3/4.3.7.12 0FFGAS MONITCalNG INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas System. 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTER This specification is providtd to enswe that the turbine overspeed protection system instrumentation and the turbine speeo control valves are OPERABLE and will protect the turbine from excessive overtpeed. Pr~otection from turbir.e excessive overspeed is required since excessive over$ peed of_the turbine could generate potentially damaging missiles which could impact arri damage thfety related l components, equipment or structures. 3/4.3.9 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUA110N INS 1RUMENTATION The feedwster/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand. LIMERICK - UNIT 1 B 3/4 3-7
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RADIDACTIVE EFFLUENTS. BASES 3/4.11.1.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM. 3/4.11.1.4 LIQUID H0 LOUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. 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 11 Column 2, at the nearest portable water supply and the nearest surface water supply in an UNRESTRICTED AREA. '3/4 11.2.1 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM. I i LIMERICK - UNIT 1 B 3/4 11-2
RAD 10AC1!VE EFFLUENTS BASES 3/4 11.2.2 through 3/4 11.2.4 (Deleted) THE INFORMATION FROM THESE SECTIONS HAS BEEN RELOCATED TO THE ODCM. k i i i LIMERICK - UNIT 1 B 3/4 11-3
- RADIOACTIVE EFFLUENTS BASES.
l t i l 3/4.11.F.5 EXPLOSIVE GAS MIXTURE ! This specification is provided to ensure that the concentration of potentially I explosive gas mixtures contained in the main condenser offgas treatment system is maintained below the flammability limits of hydrogen and oxygen, Maintaining the concentration of hydrogen below its flammability limit provides assurance that the i releases of radioactive materials will be controlled in conformance with the j requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. ~ l
.[
l LIMERICK - UNIT 1 B 3/4 11-4 I
RADIOACTIVE EFFLUENTS BASES 3/4.11.2.6 MAIN CONDENSER Restricting the gross r.idioactivity rate of noble gases from the main condenser provides reasonable assuranct that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of ' General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. 3/4 11.2.7. 3/4 11.3 and 3/4 11.4 (Deleted) - INFORMATION FROM THESE SECTIONS RELOCATED TO THE ODCM. i l l LIMERICK'- UNIT 1 B 3/4 11-5
F 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES i Section 3/4.12 (Deleted) i b THE INFORMATION FROM THIS SECTION HAS BEEN RELOCATED TO THE ODCM. BASES PAGE B 3/4 12-2 HAS BEEN INTENTIONALLY OMITTED. t i t
- f. i LIMERICK - UNIT 1 B 3/4-12-1 J
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5.0 DESIGN FEATURES 5 5.1 SITE l EXCLUSION AREA i 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1. ' 5.1.3(Deleted) 5.1.4(Deleted) 5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel lined reinforced concrete structure consisting of a drywell and suppression chamber. The drywell is a steel-lined reinforced concrete vessel in a shape of a truncated cone on top of a water filled suppression chamber and is separated by a diaphragm slab and connected to the suppression chamber through a series of downconer vents. The drywell has a maximum free air volume of 243,580 cubic feet at a minimum suppression pool level of 22 feet. The suppression chamber has a maximum air region of 159.540 cubic feet and a minimum water region of 122.120 cubic feet. I DESIGN TEMPERATURE AND PRESSURE i 5.2.2 The primary containment is designed and shall be maintained for: "
- a. Maximum internal pressure 55 psig,
- b. Maximum internal temperature: drywell 3400F.
suppression pool 2200F. i
- c. Maximum external to internal differential pressure 5 psid.
- d. Maximum floor differential pressure: 30 psid, downward.
20 psid, upward. LIMERICK - UNIT 1 5-1
THE flGURES ON PAGES 5-4 THROUGH 5-6 HAVE BEEN RELOCATED TO THE ODCM. PAGES 5-5 AND 5-6 ARE INTENTIONALLY OMITTED. I LIMERICK - UNIT 1 5-4
ADMINISTRATIVE CONTROLS PROCEDURESANDPROGRAMS(Continued)
- d. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMERS OF THE , ,
PUBLIC from radioactive effluents as low as reasonably achievable. The ' program (1) shall be contained in the 00CM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the' following elements: 1
- 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the 00CM,
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS confoming to 10 CFR Part 20 Appendix B. Table II, Column 2,
- 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,
- 4) Limitations on the annual and quarterly doses or dose comitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid ,
effluents released from each unit to UNRESTRICTED AREAS conforming I to Appendix I to 10 CFR Part 50, S) Determination of cumulative and projected dose contributions from l radioactive effluents for the current calendar quarter and current ! l calendar year in accordance with the methodology and parameters in i the 00CM 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 radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose comitment conforming to Appendix ! to 10 CFR Part 50,
- 7) Limitations on the dose rate resulting from radioactive material '
released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, ' Table II, Column 1 l LIMERICK - UNIT 1 6-14a
ADMINISIRAllVE CONTROLS PROCEDURES AND_PkOGRAMS (Continued)
- 8) Unitations on the annual and quarterly air doses resuiting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
- 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131. Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
- 10) Limitations on venting and purging of the Mark II containment through the
, Standby Gas Treatment System to maintain releases as low as reasonably achievable, and
- 11) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel ,
cycle sources conforming to 40 CFR Part 190. ;
- e. Meteorological Monitoring Program ;
i A program shall be provided to provide meteorological information in the ! environs of the plant. The program shall provide sufficient meteorological ' data for estimating potential radiation doses to the public. The program shall (1) be contained in the 00CM, (2) conform to the guidance of Regulatory Guide 1.23. " Safety Guide 23 - Onsite Meteorlogical Program", and (3) include limitations on the operability of meteorological monitoring instrumentation including surveillance tests in accordance with the methodology in the 00CM.
- f. Radiological Environmental Monitorina Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the 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 methodology 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 modifications 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 program for environmental monitoring.
LIMERICK - UNIT 1 6-14b
.- .. . . .. ~- - . - - - . - . - - . . . - - . .- . - - . . .- _-
ADMINISTRATIVE CONTROLS I A AL REPO,Ry (Continued) g , dosimeter, thermolu.minescent dosimeter (TLD), or film badge ! measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole-body dose received f rom external sources should be assigned to specific major i work functions; '
- b. Documentation of all challenges to safety / relief valves; and
- c. Any other unit unique reports required on an annual basis. [
- d. The results of specific activity analysis in which the primary coolant exceeded the lirits of Specification 3.4.5.
The following information shal] be included: (1) Reactor ; power history. starting 48 houra prior to the first sample in , which the limit was exceeded; (2) Results of the last isotopic analysis for radioirsdine performed prior to , exceeding the limit, results of analysis while limit was exceeded and feaults of one analysis after the radiolodine ! activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiolodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiolodine limit. MONTHLY OPERATING REPORTS 6.0.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main ' steam system safety / relief valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with 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. ( ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the
- objectives outlined in (1) the ODCM and-(2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
- A single submittal may be made for a multiple unit station.
LIMERICK - UNIT 1 6-16
i ADMINISTRATIVE CONTROLS l SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT
- 6.9.1.8 The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix ! to 10 CFR Part 50.
- 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.
LIMERICK - UNIT 1 6-17 c
ADMINISTRATIVE CONTROLS l I
)
, EECIALREPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. LIMERICK -. UNIT 1- 6-18
F ADMINISTRATIVE CONTROL 5 ! L , t RECORDRETENTION'(Continued)
- d. Records of gaseous and liquid radioactive material released to the environs. 4
- e. Records of transient or operational cycles for those unit components identified in Table 5.6.1-1.
- f. Records of reactor tests and experiments. t
- g. Records of training and qualification for current members of the unit staff. :
q
- h. Records of inservice inspections performed pursuant to these Technical Specifications.
l
- 1. Records of quality assurance activities required by the Operational Quality '
Assurance Manual not listed in Section 6.10.2.
- 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 PORC and the NRB. i
- 1. Records of the service lives of all snubbers including the date at which the ;
service life commences and associated installation and maintenance records,
- m. Records of analysis required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date.
- n. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
6.11 RADIATION PROTECTION PROGRAM 6.11.1 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 Part 20, each high radiation area in which the intensity of [ i radiation is greater than 100 mrem /h but less than 1000 mrem /h shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual l or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- I
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- Health physics personnel or personnel escorted by health physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.
l LIMERICK - UNIT 1 6-20
m. L ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alams when a preset integrated dose is received. Entry into such areas wi'ch this monitoring device may be made after the dose rate levels in the brea have been established and personnel have been made knowledgeable of them,
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perfom periojic radiation surveillance at the frequency specified by the Healti Physicist in the RWP.
6.12.2 In add *. tion to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that 4 major portion of the body could receive in 1 hour a dose greater than 1000 areas shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of Shift Supervision on duty and/or the health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the inmediate 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 1 hour a dose in excess of 1000 mreas* that are located within large areas, such as the 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 RWP continuous surveillance direct or remote (such as use of closed circuit TV cameras), may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. 6.13 PROCESS CONTROL PROGRAM (PCP) 013.1 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3n. This documentation shall contain:
- 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- Measurement made at 18 inches from source of radioactivity.
LIMERICK - UNIT 1 6-21
9 MARK-UPS 9 8
. . . .. .z .. <
7 .- . . 1 1 DEFINITIONS SECTION 1.0 DEFINITIONS PAGE . ACTI0N....................................................... 1.1 1-1 EXP050RE......................................
- 1. 2 AVERAGE PLANAR 1-1
- 1. 3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 i 1.4 . CHANNEL CALIBRATION.......................................... 1-1 1.5', CHANNEL CHECK................................................ 1-1 1.6 CHANNEL FUNCTIONAL TE5T...................................... 1-1 ;
- 1. 7 CORE ALTERATION.............................................. 1-2
- 1. 8 C RI T I C A L POWE R RATI 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2
- 1. 9 DOS E EQUI VA LENT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 l 1.10 E-AYERAGE DISINTEGRATION ENERGY.............................. 1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........... 1-2 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME....
1-2
'1 13 FRACTION OF LIMITING POWER 0ENSITY...................'........ 1-3 1.14 FRACTION OF RATED THERMAL P0WER.............................. 1-3 1.15 FREQUENCY N0TATION........................................... 13 1.16 I D E NTI FI ED LE A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 ISO LATION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3 1.18 LIMITING CONTROL ROD PATTERN................................. 1-3 1.19 LINEAR HEAT GENE RATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.20 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 1.21 MAXIMUM FRACTION OF LIMITING POWER DENSITY................... 1-4 LIMERICK - UNIT 1 i (;;5 l Dr.i t g 6 .
. . . . . . . . _ . . . 3 i
L I 1 I DEFINITIONS i 1 SECTION . I DEFINITIONS (Continued) PAGE
)
1.22 MEMBER (S) 0F THE PUBLIC...................................... 1-4 ! 1.23 MINIMUM CRITICAL POVER RATI0...............'.........i........ 1-4, j
' 1. 24 0FFSITE DOSE C'ALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 I
- 1. 2 5 OP E RAB LE - O P E RAB I LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4'
- 1. 26 OPERATIONAL CONDITION - CONDITION. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 J 1.27 PHYSICS TESTS................................................
1-4 ( 1.28 PRESSURE BOUNDARY LEAKAGE.................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY................................ 1-5 )
- 1.30 PROCESS CONTROL PR0 GRAM................. . .................... 1-5 1.31 PURGE - PURC7NG.............................................. 1-5.
1,32 RATED THERMAL P0WER.......................................... 1-6 i 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY............ 1-6" ! 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME...................... 1-6' 5 r 1. 35 REFUELING FLOOR SECONDARY CCRTAINMENT INTEGRITY. . . . . . . . . . . . . . 1-6 1 EVENT............................................. 1.36 REPORTABLE 1-7 1 ' 1.37 ROD DENSITY.................................................. 1-7 1,38 SHUTDOWN MARGIN.............................................. 1-7 l t-1.39 SITE B00NDARY................................................ 1-7 i
- 1. 40s 50,LIDIFICATI0h. . . .he.w. w cevh ... y r. M .~. > -i M 1,41 SOURCE CHECK................................................. 1-7 l
1.42 STAGGERED TEST BASIS......................................... 1-3 1,43 THERMAL P0WER................................................ 1-8 l l 1. 4 4 UN I D E NTI F I E D L E A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 1 LIMERICK - UNIT I ii '%$ B E03 , l
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i i; 3875101020 i f 6 i ! r :
.V :
INDEX { OEFIN1110N$ : f SECTION i PAGE- j O!FINIT10NS (Continued) 1.45 UNRESTR1;TED AREA............................................ 1-8 , ir. 1.46. VENTILATION EXHAUST TREATMENT SYSTEM.........................- 1-8 1,47 VENTING..................................................... - 18 i Table 1.1, Surveillance Frequency Notation........................ 19 ; Table 1.2. Operational Conditions................................. 1 10 : c .- ! i r I
?
L 6 I .. l l t , 1 i e i t b s
- .... c, d 3...
N3 iii in LIMERICK - UNIT.1' .
-- - - - - , n -, . . -, - -, - - - - - . - , ,, . - - ~ - - - - - ~, , - +
.e # F _. . . . . . ._ -- INDEX LIMITING CONDITIONS FOR OPERATION ANO SURVEILLAN ! REOUIREMENTS SECTION ' PA3E INSTRUMENTATION (Centinued) Table 3.3.4.2-1 End-of-Cycle Recirculation Pump Trip System Instrumentation....... 3/4 3-48
, Table 3.3.4.2-2 End-of-Cycle Re:irculation Pump Trip Setpoints.................... 3/4 3-49 Table 3.3.4.2-3 End-Of-Cycle Recirculation Pump Trip System Response Time. . . . . . . . . 3/4 3-50 Table 4.3.4.2.1-1 End-Of-Cycle Recirculation Pump Trip System Surveillance Requirements.................... 3/4 3 51 3/4.3.5 REACTOR CCRE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION......................................... 3/4 3-52 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumenta-ti0n................................ 3/4 3 f Table 3.3.5-2 Reactor Core Isolatien Cooling System Actuation Instrumentation ,
Setpoints........................... 3/4 3-55 Table 4.3.5'.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements.......... 3/4 3-56 . 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION....................... 3/4 3-57 Table 3.3.6-1 Control Rod Block Instrumenta-tion................................. 3/4 3-58 Table 3.3.6-2 Control Rod Block Instrumenta-tion Setpoints....................... 3/4 5-60 Table 4.3.6-1 Control Rod Block Instrumenta-tion Surveillance Requirements....... 3/4 3-61 3/4.3.7 MONITORING INSTP.UMENTATION l Radiation Monitifrin~g' Instrumentation me m 3/4 3-63 -
--Tatie-3. 3r7.1-L_ Radiation Monitoring Instruhntation-. .. . .. _Q/4 3-JL L!r.ERICK - UNIT 1 viii ,,,
0, ,... l ~;' i I
. . . . . .. . . . . . . . .. . . . - . .. .- I 1
i
; , - . . . . . . . _ . . . . ...w... 4' . . . . .
p . . . . . r i 1NDEK l LIMITING CCND!TIONS FOR OPERATION ANC $URVEIL'4NCE REDUIREMENTS SECTICN PAGE INSTRUMENTATION (Continued) , i Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance ' i Requirements...................... 3/4 3-66 i Seismic Monitoring Instrumentation...................... 3/4 3-68 Table 3.3.7.2-1 Seismic Monitoring Instrumentation.................... 3/4 3-59 t Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance ' Requirements....................... 3/4 3-71
,_Mbteorelegicei- Monitoring-Instr mentation. u .............. 3 /4'3:7Ss,
~.
- ---Tahie.J t 7. 3-1 Meteor 516gicaHie@, cr)ing . .
M m umentat.icne . . .. m .... 'i/4 P 74 - Table ~ 4. 5- 7. 3-1--Meteorological-Menifiring
. Inst-tenenta: Ten SENHITdhY$
' ' Rrqutremtnts.'.....................
774~3575-Remote Shutdown System Instrumentation and Controls..... 3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System' Instrumentation and Centrols...... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System
- Instrumentation Surveillance Requirements...................... 3/4 3-83 Accident Monitoring Instrumentation..................... 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation............................ 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements.... 3/4 3-87 Source Range Monitors................................... 3/4 3-88 Traversing In-Core Probe System......................... 3/4 3-89 Chlorine Detection Sycten............................... 3/4 3-90 T oxi c G as De tec-i on Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2/4 3-91 '
Fire Detection Instrumentatien.......................... 3/4 3-92 LIMERICK - UNIT 1 ix
- T5 ;
. ~ , .
n 5.......--
- e_
4 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE INSTRUMENTATION (Continued) Table 3.3.7.9-1 Fire Detection Instrumentation. . . . 3/4 3-93 Loos e- Pa rt Detecti on Sys tem. . . . . . .' . . . . . . . . . . . . . . . . . . . . . . 3/4 3-97 RWicactwe-t44dd4ftluant Monitoring Instrumen-tation......................T...... ~~ ~574'3-98 ! N N ' Tab e 3c3.7.11-1 Radioactive Liquid Effluent Monitoriqq Instrumentation. . . . . . . 3/4 3 99
\
, Table 4.3.h 1-1 Radioactive liquid Effluent Monitoring Instrumentation -
i Surveillance Requifeeents........ 3/4 3-1 1 Is,i \ I 6 Rad i@ceet N Gaseews iffh;nt Monitorin Instrumen- s tation................................g.................. 3/4 3-103 u - lm a Table 3.3.7.12-1 Retfhectiv. Ga:::n -Effl u e Monitoring Instrumentation....... 3/4 3-104 C c/w Table 4.3.7.12-1 P Mi:::tive Gas = E?E.,.ni. .. Monitoring Instrumentation Surveillance Requirements........ '3/4 3-107 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...................... 3/4 3-110 3/4.3.9 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................... 3/4 3-112
- Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.... 3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints.................... 3/4 3-114
. Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillanca Require-ments............................. 3/4 3-115
,3/4.4 REACTOR COOLANT SYSTEM .
3/4.4.1 RECIRCULATION SYSTEM Recirculatien Loops..................................... 3/4 4-1
-LIMERICK - UNIT 1 x 'yg $ @ .
.. .. . . . 8.. .. ,
...... . . . u .. .~ ., . .- . :;. . . . .
. . ...............---,.-----m.--------. . . . . . - - . . - . - ~ ~ ~ - - - - - - - - - .
i t. s INDEX L'MITING CONDITICNS FOR ODERATION AND SURVEILLANCE REOUIREMENT5 i
$!CTICN
{AGJ 3/4.9.10 CONTROL R00 REMOVAL Single Control Rod Remova1.............................. 3/4 9-13 Multiple Control Rod Remova1............................ 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1........................................ 3/4 9-17 Leve1......................................... Low Water 3/4 9-18 1 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY........................... 3/4 10-1 3/4.10.2 ROD SEQUEN E CONTROL SYSTEM............................. 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.......................... 3/4.10-3 3/4.10.4 RE C I RCU LATI ON L00PS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION.................................... 3/4 10-5 3/4.10.6 TRAINING STARTUPS....................................... 3/4 10-6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFf.UENTS
- - C o n c e n t r a ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4-II~-T Table 4.11.1.1.1-1 RadioactiveTHf0fd Waste-SampHng-and- Analysis Program........................ 3/4 11-2
- D o s a . .. y . u . . . . . . . . . . . . . . w.. ..,, c. . . . . . . . . . . . . . . . . . . . .
_ ?/411-5s
-tTourd'tidwaste--T rutmant._Sy.s1cL . . . . . . . . . . . . . . . . . . . . . . . 3 4 11 , Liquid Holdup Tanks..................................... 3/4 11-7
-3/4rlh2--GASE005-EFFLUENTS-
^
" ~175Te-Ret 4 ~ ,..... . . .... rrrr..................... 3/4 11-!
Xi 8 db LIMERICK - UNIT 1 xvi i i 6.*. . .em e emameno , y e . . ., eee en. a * *e. ammew eseme eme. .- + e e . . .* .. ********L
3875101020 INDEX j LIMITIN CON !T:0NS FOR OPERATIDW aND SURVE7LLANCE REOUIPEw!N75 l SECTION PA0! RAD 10Aci!vE EFFLUENTS (Continvec) , s Table- Ar114 2rl.2-1 Racioactive Gasacus Waste
- :Sae.nling_and Analysis Program; m...................... 3/4-11-9
- DrsE v Notle-Case 6 . .. .............. .. 3/a12CO
_ l DU U M CIns 1212.IOdff*_1U d E11iumrand. _, Racionu:Hees - in- Pa rticulate-F or% . . . . . . . . . . . . . . . . . . . 3/4 11 13
'"W6tiI8ttotr-Emhouit_ItaAtme.n' _iys t4% . ................ 3%4-11*1a Explosive Gas Mixture................................... 3/4 11-15 Main Concenser.......................................... 3/4 11-15
- Vent i ng o r FliFif MF; . . . . . . . . . ..... ......... 7f4 11-17
- 3f M T 2- 5&L12_R6p1ASTE T MAIME.C .r ....................... 17MM&~
314-It a -TOT A L-Et5 f. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~'T/ M i-E M 3h .17 RAD 10 LOO! CAL Euv!RONwEiTAL M3NITORING \ N N \ 3/4 12-1 3/.4.12,1 MO N I T O R I N G P R O G RM . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . \
\
\
'N Tatie 3.12.1-1 Radiologk;al Environmental
\N Monitoring' Program................. 3'/( 12-2 T'acQe 3.12.1-2 Reporting Leve g For Rtdio- \,
N activity Concentrations In N Nx Envi ronmental Samples. . . . . . . . . . . . . .
. N 3/4 12 d x
\s Taele 4.12.'h 1 Cate: tion Capatilities'Ror N E s 3/4 12-10 \
\nvironmentalSampleAnalysis......
3/4.12.2 LANDUSECENSL's.........i.......................N..x..... 3/4 1E O N N' N 3/4.12. 3 ' I NT E RL ABO RATORY COMPARI SON PR00RM. . . . . . . . . . . . . . . . . . . .g 3/4 12-18 l iI l i 1 , LIMERIX - UNIT 1 xvii g.g gg
L 0875101020 1 e BASES SECTION PAGE 3/4.0 APPLICAEILITY............................................ B 3/4 0-1 3 /4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 S nut OOwN nARG I N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1- 1 3/4.1.2 REACTIVITY AN0MALIES.................................. B 3/4 1-1 3/4.1.3 CONTROL R005.......................................... B 3/4 1-2 . 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS.......................... B 3/4 1-3 .r 3/4.1.5 STANOBY LIQUIO CONTROL SYSTEM......................... B 3/4 1-4
,3 /4. 2 POWER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR NEAT GENERATION RATE.................................................. B 3/4 2 1 LEFT INTENTIONALLY BLANK........................................ B 3/4 2 3 3/4.2.2 APRM SETP0lNTS........................................ B 3/4 2-2 l 3/4.2.3 MINIMUM CRITICAL POWER RAT 10.......................... B 3/4 2-4 3/4.2.4 LINE AR HE AT GENE RATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-5 3 /4. 3 INSTRUMENTATION '
l l- 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............. B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION. . . . . . . . . . . . . . . . . . . E 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION I NST RUMENT ATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 3- 2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION. . . . . B 3/4 3 3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCX INSTRUMENTATION..................... B 3/4 3 4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................. B 3/4 S 4 LIMERICK - UNIT 1 xviii Amencment No. 7 AUG i 4 237
3,. , . . .. lr .' . '. ...a P > ;.' ' l INDEX BASE 3 S!CTION PAGE INSTRUMENTATION (Continued) Seismic Monitoring Instrumentation...................... B 3/4 3-4
~MftTCMTUstCEi MDTNTD?1 ng Instrumentation. . . . . . . . . . . . . . . IF-r;T y.4 Remote Shutcown System Instrumentation and Controls..... B 3/4 3 5 Accident Monitoring Instrumentation..................... B 3/4 3-5 Source Range Monitors.................................... B 3/4 3-5 Traversing In-Core Probe System.....'.................... B 3/4 3-5 t
Chlorine and Toxic Gas Detection Systems................ B 3/4 3-6 Fire Detection Instrumentation.......................... B 3/4 3-6 Loose-Part Detection System............................. B 3/4 3-6 "Riif Ea cti ve-i.i quit E f fl uent- Moni to ri ng
~~~~15 5 t rume n t ati o n . . . . . . . . . . . . . . . . 3 . . . . . , , . . . . . . . . . . . . . . .
.. . P!/4 .9 - -
! , , / c; , U l Aaa ~ " Gaseewt. S nt Monitorin Instrumentation........................g.................. B 3/4 3-7 . 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM..................... B 3/4 3-7 i 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION I NST RUMENTATI ON. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . B 3/4 3-7 - L Eases Figure B 3/4.3-1 Reactor Yessel Water I Leve1...................... B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.................................... B 3/4 4-1 3/4.4.2- SAFETY / RELIEF VALVES.................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE l Leakage Detection Systems............................... B 3/4 4-3 Operational Leakage..................................... B 3/4 4-3 3/4.4.4 CHEMISTRY............................................... B 3/4 4-3 LIMERICK - UNIT 1 xix M EW *
.. s. _ _ _ . . _.. .. . . -.
- 3892300520 INDEX BASES .
SECTION
.P,,Ag{
REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATF0RM................................ B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUELP00L............ STORAGE B 3/4 9-2 ; 3/4.9.B and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L......... 3/4.9.10 B 3/4 9-2 CONTROL RCD REM 0 VAL............................... B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT 3/4.10 SPECIAL TEST EXCEPTIONS B 3/4 CIRC9-2 ' 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY..................... B 3/4 10-1 3/4.10.2 ROD WORTH MINIMIZER............................... B 3/4 10 t 3/4.10.3 , SHUTDOWN MARGIN DEMONSTRATIONS............... B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS............................... B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.............................. 2 3/4 10 1 3/4.10.6 324.11kADICACTIVEEFFLUENTSTRAINING N S 3/4 10-1 STA 3/4.11.1 LIQUID EFFl.UENTS x\N Concentration'....................................% 0 B 3/4 11-1 B 3/4 11-1
\ (o s e . . . . . . . . . . . . . . . . . . . . . . . . . . . .
t quid Radwaste Trea tment Sys tem. . . . . . . . . . . . . . . . . .
'B 3/4 11 2 Liq'Oid Holdup Tanks............................... B 3/4 11-2 3/4.11.2 GASEOUSsEFFLUENTS Dose Rate........
...............'................. B 3/4 11 1 Dose - Noble Gases. ................'.1,...........
s B 3/4 11 3 Cose - Iodine-131. locine-133, Tritium and Radionuclides in Particulate Form......,.....'...... 0 3/4 11-3 , Ventilation Exhaust Treatment System..............s B3/4LS s LIMERICK - UNIT 1 xxii Amendment Mo. i,7 h!AR 22 GEU
. - . , . ......._..... ... . ...:... . . _ . . . . _ . . , _ . : . , c. . , , , , , , , , , _ , . , , , , , , ,,
i INDEX BASES - SECTION PAGE RADIOACTIVE EFFLUENTS (Continued) , Explosive-Gas Mixture............................ , g 3/4 11 4 Main Condenser..................................... B 3/4 11-5 i
%nt4ng-4&Pweg4ng. .. .,,,,,,,,,,,, g 3f4 1;.,e, 37CIr3~~T6EID UDWK5TFTRir.TMtNT. . . . . . . . ..... . .. c e yf4 ,1. g . ..
p 37Cir4- 70TAk-DOSE. . . . . . ............ o .u~, s1 s 3/4.12 RADIOLOGICAL ENVIRONMENTAL MDNITORING N ' N N s\, \ *
!~
MONITORING PR0 GRAM q.....................'.h q...... 3/4.12.1, N B N(12-1 3/4.12.2 LA E USE CENSUS..... N' N 's . N - .............................w'- N B 3/4 12-l x
.N 3/4.12.3-INTERLABORAT(RYCOMPARISONPROGRAM.................\s'Bs3/4 12-2
. N. 'N. 'N -
t
. LIMERICK - UNIT 1 xxiii A.",' Slii3
..+ u . ,. f ., . ,. , . ,, . , -,. ',.,. c. . .;; y . . q .., , ,, . ,
INDEX DESIGN FEATURES .- SECTION PAGE_ 5.1 SITE j Exclusion Area............................................... 5-l' Figure 5.1.1-1 Exclusion Area.......................... 5-2 i Low Population Zone.......................................... 5-1 Figure 5'.1.2-1 Low Population Zone..................... 5-3 s : Map'ss Defining UNRESTRICTED AREAS and SITE BOUNDARY for R'adioactive Gaseous and Liquid Effluents..........;, 5-1 N x Figure 5.i}3-la Map Defining UNRESTRICTED AREAS for
\s s Radioactive Gaseous an'dsLiquid NsEffluents................>,............
's N N 5-4,,\
Figure 5.1.3-lb' Map' Defining UNRESTRICTED AREAS s for 's, Radioactive Gaseous and Liquid N. Eff1uents.% .......................' K ., 5 'N, s. x m s.s Meteor'ological-Tower Locatien...........x..a.................. 's'S-1
~s, x x Fioure5.1'4-1s,'HeteorologicalTowerLoca'.'lon.1.........
s 5-d'x s m_ ~~s_. s s
- 5. 2 CONTAINMENT
' Configuration................................................ 5-1 Design Temperature and Pressure.............................. 5-1 S e c o nd a ry C o nt a i nme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.3 REACTOR CORE Fuel Assemblies.............................................. 5-7 Control Rod Assemblies....................................... 5-7
- 5. 4 ' REACTOR COOLANT SYSTEM Design Pressure and Temperature.............................. 5-7 Volume....................................................... 5-8
- 5. 5 FUEL STORAGE Criticality................................ ................ 5-8 LIMERICK - UNIT 1 xxiv ' 06 S D'd i
~
; y g ; J *./.~.'... ..il. .4 .f
.. -;; . . . . . . . .~ .. ..: .;... . . :-
_~.-
. . . , . , . . . . . . . . - . . , . . . -. a ...--. - . ....... ..- _ .
m: , f L F INDEX- . ADMINISTRATIVE CDNTROLS
$ECTION: ,
PAGE 6.5.2 NUCLEAR REVIEW BOARD (NRB) Function .............................................. 6-9 Composition ...........................-................ 6-9 Al t e r n a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 Consultants............................................' 6-10 Meeting Frequency.................................,.... 6-10 Quorum................................................. 6-10 i 8 Review................................................. 6-10 , 1 Audits................................................. 6-11 Records................................................ 6-12 6.6 REPORTABLE EVENT ACTI0N...........,......................... 6-1? ; 4 6.7: S A ET/ L I M IT VI O LAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 12 J 6.8 PROCEDURES AND PROGRAM 5..........................;.......... 6-13: 6.9 REPORTING REOUIREMENTS 6.9.1 ' ROUTINE REPORTS ....................................... 6-15 i i Startup Report......................................... 15 , Annual Reports ....................... ................ 6-15 m Monthly Operating Reports......................... .... 6-16 Annual Radiological Environmental Operating Report..... 6 15 Semiannual Radioactive Ef fluent Release Report. . . . . . . . . 6-17 6.9.2 SPECIAL REP 0RTS........................................ 3-18 6.10 RECORD RETENTION........................................... 6-15 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-2C 6.12 HIGH RADIATION AREA........................................ 6 20
*u? 8 66 LIMERICK - UNIT 1 xxvii "'
. . .;; , ; . . . . . . . . . < . . . . . . .....- ... - .. , ; w< .,
+
a = , 4 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE-6.13 PROCESS CONTROL PROGRAh! (PCP)..............................
, 6-21 6.34 0FFSITE DOSE CALCULATION MANUAL (00CM)..................... 6-22 64 5- MA30R CRAN 3t5 IO KADTUAGIiVt E 5T2 T EAT"EF MS b 5,... ,, e. 27-t i
4 LIMERICK --UNIT 1 xxviii M 6 iiH ____ _ _ _ _ _ .________.___.__.__-._____._.___m__ __m. _ _ _ _ _ _____. ____ _ _ . _ _ __ _ _ . _ _ _ . _ _ -
.m__m._:.__ _ m-s
lll ' OEF7NRT10NSL E
' LOGIC SYSTEM FUNCT20NAL TEST' -
1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components,. i.e., all relays and contacts, all trip units, solid state logic elements, etc.- of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested. MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.21 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest value of the FLP0 which exists in the core.
~
MEMBER (S) OF THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who'are not occupationally associated with the plant. This category does not include employees of
- the utility, its contractors, or vendors. Also excluded,from this category are. persons who enter the site to service equipment or to make deliveries. '
This category does include persons who use portions of the site for recrea-tional, occupa+1onal, or other purposes not associated with the plant. MINIMUM CRITICAL POWER RATIO ' 1.23 The MINIMUM ~ CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). OFFSITE 00SE CALCULATION MANUAL - '
= .w jirI1' The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology ..
and parameters used in the calculetion 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-grams required by Section 6.8.4 and (2) descriptions of the information that should be -included in the Annual Radiological Environmental Operating and Scan-annual Radioactive Effluent Release Reports required by Specifications 6.S.1.37 and 6.9.1.4LS OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE 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 equipmen.t that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s)., OP'ERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position ana average reactor coolant tempera-ture as specified in Table 1.2. PHYSICS TESTS 1.27 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 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. LIMERICK - UNIT 1 1-4 8II E Mb t ( 6 St $g49W=6n$g g 9 'O O
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.. m'1 ..s. ' . .. s. . . .
. . . _ . . _ . . . . . . . . . . . . . - . _ . . . . . . . . - ..-.+..-- .-- ... ;
. i
.y V
r
.- DErINITIONS ?.
V PRESSURE BOUNDARY LEAKAGE k 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall. 1 PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when: ,
- a. All primary containment penetrations required 'to be closed during
. accident conditions are either:
.1. Capable of being closed by an OPERABLE primary containment. -
i automatic isolation system, or , 2.- Closed by at least one manual valve, blind flange, or- :
-deactivated automatic valve secured in its closed position, '
except as provided in Table 3.6.3-1 of Specification 3.6.3. .
- b. All primary containment equipment hatches are closed and sealed. .
- c. The primary containment air lock is in compliance with the
- requirements of Specification 3.6.1.3. ,
- d. The primary containment leakage rates are within the limits of i
Spe:ification 3.6.1.2. i [ o
- e. The suppression ch' amber'is in compliance with the recolirements of Specification 3.5.2.1.
- f. The sealing mechanism associated with each primary containment penetration; e.g., welds,. bellows, or 0-rings, is OPERABLE, PROCESS CONTROL PROGRAM . i
.w125 TilelROCE'Sli C'0'NTRdLlRUGRAT'(PCPf'shall cont'ain the' currert termulas, sampling, analyses, test, and determinations to be made to ensure that prec:u ing'and packaging of solid radioacti wastes based on demonstrated processic' '
of actual or simulated wet solid w tes will be accomplished in such a way 7 - to assure compliance with 10 CFR arts 20, 61, and 71, State regulations, burial ground requirements, and ther requirements governing the disposal n ' solid radioactive-waste. 4&
,0( : M .
em uovu un iaoorar.ory ' l: scale and full scale testing or experience. .W4* h dawd a"-ing,_.thaJCP-shel l heitRTe~an 1mftWteeMemoLeenditic,n 1:trat-must4e-sat 4s.fied, based on L fu4 scale testing,-.tcL assure- that -dew &ter4D9-Obbe44-Pe$ ins , POW 44Eed h resins, and filter sludge & uilt rac"H ~ 4n "^1' - e nf +r:: meter, :t-the - time-of-dtsposel , w -; tH r. tr.: limits vi 10 CTR *ert 51 :nd d th_e low-level r-adleactM :::t= dienul 14.e . PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or L gas from a confinement to maintain temperature, pressure, humidity, L concent. ration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. h ! -LIMERICK - UNIT 1 1-5 G 615 g l L l 0
t
+ w .O .. . . . . . - .. ., .w '^...
. .n . * .* - ' ' ' ^
, ... ; . ..,m..c.. ..s a.. - '
3 DEFINIT!ONS * '
- s. '
REFUELING. FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued) , i
- b. All refueling floor secondary containment hatches and blowout panels are closed'and sealed.
- c. -The standby gas treattient system is in compliance with the requirements '
of Specification 3.6.L.3l
- d. .At least one door in tach access to the refueling floor secondary-containment is closed. '
L e.
' The se'aling mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
- f. The pressure within t'he refueling floor secondary containment is
, less than-or equal to the value required by Specification 4.6.5.1.2a.
REPORTABLE EVENT L 1.36 A REPORTABLE EVENT shall be any of those~ conditions specified in L Section 50.73 to 10 CFR Part 50. ' ROD DENSITY L 1.37 ROD DENSITY shall be the number of control rod notches inserted as a o fraction of the total number of control . rod notches. All rods fully
. inserted is equivalent to 100% ROD DENSITY. '
1 SHUTDOWN MARGIN '
- i. 1,38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is-i L
suberitical or would be suberitical assuming all_ control rods are fully inserted except for the single control rod of highest reactivity worth
-t which is assumed to'be fully withdrawn and the reactor is in the shutdown 9
condition; cold, i.e. 68'F; and xenon free. SITE BOUNDARY y 'i. :* M u , , i L 1.39 The SITE BOUNDARY. shall be that line as defined in3 Figure Sni.:.. SOLIDIFICATION. m 1.40 SOLIDIFICATION _shall be the immobilizati n et radioactive wastes such as evaporator bbttoms, spent resins, sludges, an everse- sis concen-
- tratesxas a result'of a process 6f thoroughly mixing he wast ype with a i
l solidification agent (s) to form'a free standing monolith with c emical and l physical characteristics specffied in the PROCESS CONTROL PROGRAM'(P.CP). ' i SOURCE-CHECK I 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. ' 1.
~
l L ! l LIMERICK - UNIT 1 1-7
.g* g q* . g p ... . . . . . . . . . ..a ...... . . . - . --- ---
a.: INSTRUMENTATION 3891002560 METEOROLOGICA\xMONITORINGINSTRUMENTATION N. LIMITING CONDITION FOR OPERATION '
\
-3.3.7.3 The meteorolo'icalg monitoring instrumentation channels shown in Tabla '
3.3.7.3 1 shall be OPERAELE.
\
APPLICABILITY: At all timei, ACTION:
- a. With one or more meteorological monitoring instrumentation channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status,
- b. The provisions of Specification 3.0.3 are not applicable. ]
SURVEILLANCE RE0VIREMENTS 4.3.7.3 Each of the above required meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the.CHANtEL CHECK4 and CHANNEL CALIBRATION operations at the frequencies snown in Table 4.1.7.1 1.
+
5 LIMERICK - UNIT 1 3/4 2 73 Amendment No.11 N9'i 7 g.;
f' 389.1002560 TABLE 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION s N MINIMUM N. Tower 1 Tower 2 INSTRUMENTS INSTRUMENT (Primary) (Backup) OPERABLE
- 1. Wind Speed
- a. Elevation 1 30 feet or 159 feet 1
- b. Elevation 2 175 feet or 304 feet 1 ,
- 2. Wind Otraction
... a. Elevation 1 30 feet or 159 feet 1
- b. Ele'vation 2 175 feet or 304 feet 1
- 3. Air-Temperature Difference
- a. Elevations - 266 feet- 300 feet-26 feet or 26 feet I-LIMERICK - UNIT 1 3/4 3-74 g'; 3 ..;-:.;
i
.. . - . . . , . . - . ~ . - - - - - - - - - - - - -
~ '
TABLE 4.3.7.3-1
~ -
. \ .
+
'METEOROLOGIDAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
\ 'N '
CHANNEL- CHANNEL INSTRUMENT x CHECK CALIBRATION N
- 1. .Wind Speed ,
- a. Elevation 1 (Tower'1 and Tower 2) D SA
_,- b. Elevation 2 (Tower 1 and Tower 2) D SA ',
- 2. Wind Direction. .
- a. Elevation 1 (Tower 1 and Tower 2) D SA
- b. Elevation 2 (Tower 1 and Tower 2) D SA
- 3. Air Temperature Difference
- a. Elevations 266 ,26 ft (Tower 1) D SA
~
- b. Elevations 300 - 26 ft (Tower 2) - D SA Ia .- . .
- LIMERICK - UNIT.1 3/4 3-75 i
)
- 1
. - - . - - . - - - .-. . - - - -- -~
-t i .
e 389 0025s0 i - n. INSTR NTATION-
\ .
.-n _ f RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ~"~
"~p ~
. LIMITING-CONSITION FOR OPERATION ,
N 3.3.7.11 The ra(ioactive liquid effluent monitoring instrumentation channels shown in-Table 3.3.7.11-1 shall be OPERABLE with'their alarm / trip setpoints- g set to ensure that\the limits of Specification 3.11.1.1 are not exceeded. The Y. alarm / trip setpointsi of these channels shall be determir.ed and adjusted in accordance with the mathodology and parameters in the OFFSITE DOSE CALCULATION MANUAL.(00CM). APPLICABILIITY: 'At all t mes. ACTION: a ,
- a. With a radioactive quid effluent monitoring instrumentation channel alarm / trip se point less conservative than required by the
', , above specification, i ediately suspend the release of radioactivo ,
liquid effluents monitored by the affected channel, or declare the channel inoperable,
- b. With less than the minimum umber of radioactive liquid effluent monitoring instrumentation c annels OPERABLE, take the ACTION shown in Table 3.3.7.11-1. Restore the inoperable instrumentation to OPERABLE status within the tim specified in the ACTION or explain in the next Semiannual Radioact ye Effluent Release Report why this inoperability was-not corrected e thin the time specified. 3
- c. The provisions of Specification 3.0) are not applicable.
l SURVEILLANCE REQUIREMENTS r q 4.3.7.11- Each radioactive liquid effluent monitori' instrumentation channel shall be demonstrated OPERABLE by performance of the HANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL'T,EST perations at the frequencies shown in Table 4.3.7.11-1.
" Excluding the flow rate measuring devices which are not determined and djusud in accordance with the ODCM.
LIMERICK - UNIT 1 3/4 3-98 Amendment No.11 WQV 'l 5% I
3.._. e .. u. . . . . ; - .= _ .. . . __ a .:. . . . x ,h
\ .
N q TABLE 3.3.7.'11-1
' N !
RADI0' ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION - ' INSTRUMENT N ' MINIMUM CHANNELS' OPERABLE ACTION
- 1. GROSS RADIOACTIVITY MONITORS PROVIDING .
AUTOMATIC TERMINATION OF RE,LEASF
, a. Liquid Radwaste Effluent ine 1 100 s
- b. RHR Service Water System Eff uent Line 1/ loop '
10,1
- 2. GROSS RADIOACTIVITY MONITORS NOT -
PROVIDING' AUTOMATIC TERMINATION
... OF RELEASE
- a. Service Water System Effluent Line .'.
, 1 101
- 3. FLOWRATEMEASOREMENTDEVICES ,
- a. Liquid Ra'dwaste Effluent Line l ' 10 !
- b. Discharge Line 1 102 s
i i LIMERICK - UNIT 1 3/4 3-99 ~ t l'
. . . . . . . . . . - . . . . . . . . - . . . - - - - - - - - - ' ~
e
J.x : : c .-. c. . .
.. _- J .
~ '
. . .t
^
. -. .i ...c.. _
4 ( TABLE 3.3.7.11+1 (Continued) t
, ACTION STATEMENTS N- .
ACTION 100~- Wi he number of channels'0PERABLE less than required by the ! Minim (um Channels' OPERABLE requirement, effluent releases ' continue or up to 14 days provided that prior to initiating. a release: ,1 -
~
- a. At least\two independent samples are analyzed in' accordance - i with Specification 4.11.1.1.1, and- ,.- r l
Atleasttw\otechnicallyqualifiedmembersofthefactlity b. staff independantly verify 'the release rate calculat. ions and discharge ltqe valving; Otherwise, suspend. release of radioactive effluents via thi's pathway. -
. ACTION 101 - WiththenumberofchannelsSPERABLElessthanrequiredbythe Minimus Channels OPERABLE re irement, effluent releases via :
this pathway may' continue for- to 30 days provided that; at
. least once per 8 tyours, grab s les are collected and analyzed for gross radioactivity (beta or\ gamma) at a limit of detection i .
of at least 10 7 microcurie /mL. ACTION 102 - With the number of channels OPERAB 'less than required by-the ->-.a l' Minimum Channels OPERABLE requiremen effluent' releases via- L-this pathway may continue for up to 3 days provided the flow i L J-rate is estimated at least once per 4 h urs during actual I releases. Pump-curves generated in. sit may be used to estimate
. flow. .
L - l, E l :' ' i 1 I' l-i 1.. p' s- .
\ .
r \. L 65 LIMERICX - UNIT 1 3/4 3'-100 _.i
'i; T .. :.
,r
. 1 r- -TABLE 4.3.7.11-1 ,-
4
-t
.i5 !
R RADI0 ACTIVE: LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENIS R -
- CHANNEL j c- CHANNEL SOURCE CHANNEL TUNCTIONAL-
.! $ INSTRUMENT CHECK CHECK
-i CALIBRATION TEST =
j
- 1. '
' GROSS RADI0 ACTIVITY. MONITORS PROVIDING. ,/
l ! AUTOMATIC TERMINATION OF RELEASE- -'
/
a.' Liquid:Radwaste' Effluent Line / P P, ,/ R(3) Q(1) I
- b. RHR Service Water System Effluent Line D M R(3) Q(1) -- !'
.t
- 2. GROSS RADI0 ACTIVITY MONITORS NOT PROVIDING I' l '
AUTOMATIC TERMINATION OF RELEASE *
- ^ s,/
$ a. Service Water System Effluent Line ' 0/ M R(3) Q(2).
- 3. FLOW RATE MEASUREMENT DEVICES g
l 8 '1 g
- a. D(4) N.A. R !
l Liquid Radwaste Effluent Line l Q
- b. Discharge Line D(4) N.A.. R
, Q j:
- (.
/ -
\ l _
*j-l l , /
l 's / i S/ 1-
/. .
.i.
I l , u* g' 4 I 4
..u_4 _
- a. ~.a. ..
.g_ f.-; 5.--+-~~~e-------------~~---~~----- -~~
l; ,: - 0 ?s R- \ ', ~' s -
, TABLE 4.3.7. 1-1 (Continu'ed)
TABLE NOTATIONS - (1) 'The' CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of.the followi'ng conditions exists:
- 1. Instrument indicates measured-levels-obove the alarm / trip setpoint.
2.: Circuitifailure. . s .
,- ' 3. Instrument, indicates a downscale failure; ,
(2)'The CHANNEL F'UNCTIONAL 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.
l 2. Circuitfailure.'\ \ -
- 3. -Instrument indicates a downscale failure.'
\- .
(3) The initial CHANNEL CAlt9 RATION shall be performed using one or more of the reference standards cerstified by the National Bureau of Standards (NBS) or using standards that havt 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 be'en related to the initial calibration shall be used.
(4) CHANNEL' CHECK shall consist of verjfying indication of flow during periods
', of release. CHANNEL CHECK shall be s made at least once per 24 hours on days ',
on which continuous, periodic,;or batch releases are.made. ,
\
\
\ \
\
LIMERICK - UNIT 1 3/4 3-102 , g 1 s _ e 4 mme -nnmE ~ 1 -m
. . , . . . . - ~ ~ . - . . - . ~ ~ ~ - - - - - - - - - - --- - - ~ ~
l INSTRUMENTATION WhAY.m ASEe d CirLZ "i MONITO?.fNG INSTRUMENTATION LIMITING CONDITION FOR OPERATION e 5 3.3.7.12 The =_q.k=:t w eh: gaster O'" k =t monitorin instrumentation channels shown in Table 3.3.7.12-1 shall be OPERABLE with t ir alarm / trip setpoints set to ensure that the limits of Specification 3.11.2. are not exceeded. 'che-
+h / trip =tpint;* ;f th;. s.;;iinth ;b;ar.r.h eheli L. Ce;.ereine; ie 7.;;er;==
.w i th th: =th.~" h;;; =0 ;;r;st:r; in th. O*".
APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:' cm Cv N .
- a. With : r:di =f:th S gas-^r w :'"1=:t monitoring instrumentation channel alara/ trip setroint less conservative than required by the above Spec.,ification.-i t ht:1y n;; n th. ..;. .. .i .edio.a;,e
;. 3.....
- ^ - ' ' - - " b;7 th; ;ff;;;; -h channel inoperableAuf to iv tLt ACCcN c d ,;c.c ; er declare the un A E R 3.3,?,Jt-l
- b. With less than the minimum number of - N: gas;;e: ;ffhat monitoring instrumentation channels OPERABLE, take the ACTION shown-in Table 3.3.7.12-1. Restore the inoperable instrumentation to th:^ . IT!" er :gh'-
OPERABLE
-hy this inaaerali'ity status withintir.: g! nat;;;if
- = =ttdhd f-
': ti--1; nn=r 'n th:
.~ . e a . ... . . i n .a 4... .L. e , < 3 . . .. n a . . n o c.
\
The provisions ~of Specifica ions 3.0.3 and 3.0.4 ore not applicable. . SURVEILLANCE-REQUIREMENTS 30 dA7< 4ad8tusuGe.55bl f [Y* somri L 5 rl % ek % -t La Co"~ ssvo n eu mW c.4 /05 M E- b 4.3.7.12 Each re.di.{ nth; gaseeve-e#%ent monitoring instrumentatfon/tirTu<.i f b h 6.T.c w/ f channem oY.gf,ff'7 w shall be' demonstrated OPERABLE by perfo.mance of the CHANNEL CNECK, Z = 0;
- g **
- g
/
{M CHANNEL CALIBRATION, and CHANNEL FfNCTIONAL- TEST operations at the frequencies shown in Table 4.3.7.12-1. eaAN*'$ g ufe
*The alare/ trip setpointe for the Main Condsnser Of fgas Treatment System Explosive Gas ."onitoring Systes =d the " ia Cenden er 0**;r *ntreet = t Sdicti= "=itrb set in accordance w'th Specification 3.11.2.5 and 3.11.2.6, resp :t! =1y.
1 LIMERICK - UNIT 1 3/4 3-103 ng bk 4 m ui ll
e-e i.- y f/f'Logd6 TABLE 3.3.7.12-1
~
~
w~ { n N'GASEfRIS-EfftBEWT MDNITORING INSTRt#ENTATION c (
~
INSTRUMENT . MINIMt#f CHAf81ELS OPERABLE- APPLICABILITY ACTION
- 1. MAIN CONDENSER OFFGAS TREATE NT SYSTEM ~
- f. EXPLOSIVE GAS MONITORING SYSTEM -
- a. Hydrogen Monitor 1 k.* 110 i
- 2. SOUTH. STACK EFFLUENT --
9 E !' M ITORING SYSTEM 'N ' i
- a. Gas Activity Monitor N
- e %
1 N I1 .
- b. Iodine Sampler
- 1 s 11 w s x h
- c. Particulate Sampler 1
- 112 :
; d. Effluent System Flow Rate Monitor. 1
- 113
- e. Sampler Flow Rat Monitor- 1
- 113
- 3. NOR STACK EFFLUENT ITORING '
SY TEM
- a. le Gas Activity nitor
- 114
- b. I ine Sampler 1
- 12 i -
- c. P rticulate Sampler 1
- 12
- d. ~Ef,1uent System Flow Rate Monitor . ; - --
1 - 1 :-
\
\ l N -
- e. Sampier Flow Rate Monitor s
\x 1 113 l>
x i s
.a u -....--.--.. . -~ ~ ~ ~
TABLE 3.3.7.12-1 (Continued) TABLE NOTATIONS gg *At Y ' W ...D =i 46& D"Duringtreatmentoperation system. of the mai.n condenser steam jet air ejector and offgas
""Ou,17,9 epereiter, of tre f.et e ir,ter.er.ee ;t.ep wer,ti'ett: =' = ;;tr.
ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimus Channels OPERABLE requirement, operation of main condenser
'offgas treatment system may continue e r "p t^ 30 tye provided.
within the following 4 hours. grab samples are collected at least once per 4 ho
' ACTION 111 -
N With the number of chandels OPERABLE less'than required by the Mini'aum Channels OPERABLE requirement, efflue'nt releases via s
'this pathway may continue for s up to 30 days pr M ded' grab samples areJtaken at least onctsper.8 hour:, and t se samples are analyzed'(or gross activity 6tithin 24 hours ,
ACTION.112 - With the number'of charinels OPERABL less than requir d by the Minimum Channels sOPERABLE requiremen , effluent releas s via this pathway say continug for up to 30 day provided sample are con-inuously collected w th auxiliary sambling equipment a required n Table 4.11.2.1.2-1. ' ACTION 113 -- ' th the number of-chann is OPERABLE les than required y the M nimum Channels OPERABL requirement, e luent releases via this pathway may continue for-to 30 days pr vided the flow ate is es inated at least once-pe 4 hours.
. ACTION 114.-
Wit the number of channels PERABLE less t an required by the Mini um Channels OPERA 8LE re irement, affluent releases v this path y say' continue for up t 30 days provi ed grab sample are taken t least once per 8 hourg and these a les are analyz d for gr' s activity within 24 h6urs and provi d the machenic 1 vacuum s are not' operated. ACTION 115 - With the n erofchannelsOPERAKLElessthan quired by the Minimus Cha nels OPERABLE requir nt, releases the environne t may continue or up to 72 hours pro ided that the orth Stack Effluent Noble as Activity Monitor s OPERABLE; ot erwise, be \ in at least HOT HUTDOWN within 12 ho s. s
\
1: !L % LIMERICK - UNIT 1 3/4 3-106
y .
'TA8LE 4.3.7.12-1 c .(sP40 w/E- '
3' 22 0%!!VE GASEG'J ZTTiUEi a g - MONITORING INSTRUNENIATION SURVEILLANCE REQUIREMENTS s.
~
CHANNEL MODES IN WHICH ;--
- .- E INSTRt#EENT CHAf81EL -SotNitE- CHANNEL FUNCTIONAL SURVEILLANCE-
~CHFCK 4!HEElt- CALIBRATION TEST 3
n IS REQUIRED _ .
- - 1. " '
. MAIN CONDENSER OFFGAS TREATfENT
-SYSTEM EXPLOSIVE GAS MONITORING SYSTEM ;:
t0 ;.
- a. :t,d g .. Monitor 0 M
2. c ~ QO) 'M *T ' ' TH STACK EFFLUEltT TORING SYSTEM \
- a. Nob Gas Activitf Moniter-
\ 0
~
R(2)N Q(1) x-
- i
'[
9
- b. Iodine Sampler W (4) N. . M.A. N.A.
N
- l { c. Particu te Sampler W (4) N.A. N.A. N.A. *
- d. Effluent System Flow Rate Monito D N.A. q
.R *
- e. Sampler F ow Rate Monitor D N.A. R Q i-
- 3. NDRTM STACK E FLUDIT NDIIITORING SY TEM
, I' *
- a. Noble Gas twity Monitor D' M R 2)
- 1)
- b. Iodine Sample 5
W (4 N.A. N. N. A *
- c. Particulate 5 ler. W (4) N. A.
N.A. N.A.
- f I d.' Effluent System F Rate Monitor D
! N.A. R 'Q \ *
! y- e. Sampler Flow Rate Monitor D N.A. 'R.
- Q I I :
S,n - o I
~
.~ - - - . . - . - - . , . , - .
- n. . . - - ,_. . - . . , , . - . . - . . ,~,.,, _ . -. ,
l'
-3 . . . . . . . - . ...... . _ . _ . , . . . . ,,
w L
- TABLE 4.3.7.12-1 (Continued)
TABLE NOTATIONS
.m _,. , l 1
29 During operation of'the main condenser steam jet air ejector and offgas
. treatment system.
*** 0 ria; eper;tien ;f th; het :.. int;nenee et.ep wenti'etien . ;.w.i .y.i .
(1) The CHANNE UNCTIONAL ESTshallalsbdomonstratbtcontro1\som , larm annune tion occurs itsany of the-f' llowing condit ons exist *
\
- 1. . Instrument i icatesasa\ sured levels ab ve the ala rip setpo t.
- 2. cult.failur '
l L 3. In rument indica es a downsca s failure.
- 4. Instr nt controls t set in o rate mode. -
(2) 3 e M nitial CHAN cal.ISRATIONDi e perfortiedsusing one 1eore of j the rh erence standards'Cartified by t or usir(ig'rgnent a been obtai ' Natio~nal standards:that hassurance afron'suppli Bu auofStandQds(NBS) s that partTcipate in measu ivities with S. These s dards shall ermit calibratin'Athe system over ts intended r nge of energy nd measur nt range.' For sybsequent CHANNE CALIBRATION, sources:that we been rol ted to the initialN klibration sha ( be used. !. ( D (47 The CHANNEL CALIBRATION shall-include the use of standard gas samples L containing a nominal: , i
- 1. 0.0 volume percent hydrogen, balance nitrogen, and 'i
- c. 2. 4 volume-percent hydrogen, balance nitrogen.
'( 4')' ridged rticulate filt 1 be chan'g LIMERICK - UNIT 1 3/4 3-109 U t In '
. . ;. '. . . . . !: 2i" J := =- -- ' . -. . - -.--. - - i -. -- -- - ---.- - a ' ' ' '
I
, a i.
. '3 11.1 LIOUID EFFLUENTS ~
f CONC TRATION. LIMITING C DITION FOR OPERATION < 3.11.1.1-- The co entration of radioactive mateNa1 released in liquid effluents- .- ! to UNRESTRICTED- ARE45 (see 71gure 5.1.3-1) shall limited to the concentrations
.specifiedin10;CFRhrt70,'AppendixB,TableII, Jumn 2-for radionuclides other than Cissolved og entrained noble gases.
For 'ssolved or entrai-ed noble'
' gases, the concentratioh shall be limited to 2 x 10 4 crocuries/mi total activity. ,
. 1 APPLICABILITY: At all times. , ,
hCTION: W h the concentration of radioacti
~
l '- material released in liquid fluents to UNRESTRICTED AREAS exceeding the abo limits,.'immediately restore e concen- ! ,
- tration to within the above liinits.
s- , SURVEIL CE REQUIREMENTS g(- 4.11'1.1.1 Ra oactive liquid wastes shall bc sa led and analyzed according K .to the sampling. d, analysis program of Table 4.11. ;1.1-1. l 4.11.1.1.2' The res ts of the r'adioactivity snalyses hall be used in' . accordance with the m thodology and parameters in the O M to assure that the L concentrations-at the int of release are maintained wit in the limits of Specification 3.11.1.1. v L l l O LIMERICK UNIT 1 3/4 11-1 D 5 fr/ . . m ._ t is
., . . - . . . . . . . . . . . . . . . . . - - ~ ~
l? v ,r?j, y)sf W ' r, *. **'
*' .O* s py * * s o
n' V"' Y,b T ' ' ~ 4 'A ' '
' ' '"7.'
?'* ~ '
a,. ,....i. -... .. . .. u . .. .. . . . . . . ~ . ~ . . . ~ . . . . . . ' . . .-
. . :.. L,;. .. ;;, b ' ^ "
....~...............-...:- '
)
.. .\
TABLE 4.1111.1.1-1 ' RADIhCTIVELIQUIDWASTESAMP NG AND ANALYSIS PROGRAM
\ \ LOWER LIMIT L MINIMUM TYPE OF
. '; LIQUID RELEASE SAMPLI ANALYSIS ACT! ITY OFDETEC{ ION (LLD)
, STYPE FREQUENCYs FREQUENCY ANAL S (pCi/ml)
(- . A. Batch Waste Release Each Batch P
\/P Each Batch Principa Emitters}G a 5x10 7 b
Tanks - N I-131
\1x10.s
- 1. Floor Drain P M. Dissolved and 110.s
- l. ' Sample Tank One Batch /M Entrained Gases No. 2 (Gamma Emitters) -
- 2. Laundry P .M H-3\. lx10 5 Drain Sample' d \
Tank Each Batch Composite
. Gross Alpha 1x10 7 .
P Q Sr89, Sr90 5x10J . Each Batch Composite d q Fe-55 -1x10 8 , i
\
B. Continuogs W W Prinedpa 5x10 7 . j N 'Rele'ase Grab Sample . Emitters} Gamma I-131 1x10.s
- 1. RHR Service W -W Dissolved nd 1x10 h Water System Grab Sample Entrained Ga.'es
- Effluent Line f (Gamma Emittet5)
- 2. Service Water W M H 1x10 8 System Grab Sample Composite d Eff1 nt Gross Alpha 1x10 7 Line W Q d Sr-89, Sr-90 s 5x10.s Grab Sample Composite Fe-55 1x10 8 AI.'3 8 l30
- LIMERICK - UNIT 1 3/4 11-2 -- t
- e. . # .we . e e . we e-..se.. e a. . ****a.m* * * * * * * * * * * '* * * * * ' ~ " * * "
. . . . . :. .L .u . . . - . . . . . . . . . . . . .. _ . . . . . . . . . . .- . . .
. . . ~ ..
,. = .
.w -
' ~ '
',~. :. . . . . .
r . . N. -- ^ - .; .. -- L '
~~? -- ~ =
-s l
~ ,
TABtf 4.11.1.1.1-1 (Contin',ed) TABLENOTATION[
"The LLD is defined, for' purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only
~5% probability of falsely concluding that a blank observation represents a "real" signal. .
For a peticular ineasurement system, which may include radiochemical separation: - 4.66s'b ' LLD = E V 2.22 x 10e a y . exp (.Aat) Where: LLD is the a priori lower limit of detection as defined above (as microcuries'per unit mass or volume), shis 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 disintegration,
, V is the sample size, in units of mass or volume, ,
= 2.22 x 10s is the number of disintegrations per minute per microcurie, ,
0 is the fractionali radiochemical yield, when applicable,* N A.is the radioactive decay constant for the particular radionuclide, and At,for the plant effluents is the elapsed time between the midpoint of sample collection and time of counting.
~
- Typical values of E, V, Y, and At~ should be used in the calculation.
It should be recognized that the LLD is defined as an a priori (before the ' fact) limit representing the capability of a measuremeiit system and not as an a posteriori (after the fact) limit for a particular measurement. LIMERICK - UNIT 1 3/4 11-3 t I l
. l
, . . . . . . . . , . . . . . . . . . ~ . . ~ . ~ * ~ ^~ ' ' * ' ' ' ' ' " '
4 d . g%, S.*} % $ %.h* W I 8%*l #* '
.. :.... .. . . . . ri .
TABLE 4.11.1.1.1-1 (Continued)' - TABLk-NOTATIONS b A patch release is the discharge of liquid wastes of a discrete volume. Erb r to sampling for analyses, each batch shall be isolated, and then
'thorou,ghly mixed to assure representative sampi.ing.
C N The prinhipal gamma emitters,for which the LLD, specification aoplies include . 1 the-following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134 i Cs-137, Ce-141,.and Ce-144. This list does not mean that only these nuclides', are to be considered. lOther gamma peaks that are identifiable, together with-those of the above nuclides, sh'all'also be analyzed and' reported in the Semi- ' annual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8. composite' sample is, one in which the quantity of liquid sampled is propor- , onaltothequantitfJofliquidwastedischargedandinwhichthemethodof' ' sampling employed results in a specimen that is representative of the liquids released.
'3 .
'A continvous release is the discharge of liquid wastes of a nondiscrete .
volume, e.g. , from a volume of'a' system that has 'aq input' flow during the continuous release.. . f Whenever effluen
\ releases are in e\ xcess of the monitor's setpoint.
N g
\
1 L , l-LIMERICK - UNIT 1 3/4 11 E'3 5 3;; -- t
. . : . . - ,. v . , . . . . . . ,. .. . , .r ->.c i- '
3892300520 RADI0AC'TIVE EFFLUENTS s DOSE s x - l . N . N LIMITING CONDITION-FOR OPERATION \ \ , 3.11.1.2 Thedos\
\ \x e or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid e'ffluents released from the site to UNRESTRICTED AREAS (See l Figure 511.3-1) shall b'e mited:
- a. During any calendhr quarter to less than a equal to 3 mrems to s
.thytotalbodyandtoless.thanorequalt{ox10mremstoanyNrgan,and .
gb.s N N \ i Durin'g any c'alendar year to less than'or equal to 6 mrems to tha
\ otal bo'dy and to less t a r equal to 20 mrems\ o any organ.
At all times. N \ ' APPLICABILITY: - ACTION: s s x \
- a. With the calculated dose from the relgase of radioactive \
materials in liquid e'f submittothe(luentsexceedinganyoftnqabovelimits,preph(eand Commission within 30 days,xpursuant to Specifica- - tion 6.9.2, a 5'pecial Report which identif(es the cause(s) for\ . exceeding the limist(s) and defines the corrective actions that have been takentotoassurksthat to be taken reduct thesubsequent releases and the prbp\will be in complianceosed releases withtheabovelimits.NThisSpecialReportshal]alsoincludethe' - < radiological impact on fi Qished drinking water supplies at the nearest downstream drinking \ water source. ( *
\
The provisions of Specificati'an 3.0.3 are not applicable.
\
.\ ,
SURVEILLAN E RE0VIREMENTS \ _
.'x \ -
4.11.1.2 Cumula ive dose contributions from likvid effluents for the current calendar quarter d the current calendar year shall be determined in accordance with the methodolo and parameters in the 00CM at\least once per 31 days. LIMERICK - UNIT 1 3/4 11-5 Amendment No. 26 M
RADIDACTIVEEFFLUhN'Ts . . 389230052.0 LIQU10 RADWASTE TREATMENT $Y$7EM
'LIMITINGCONDITIONFOROPEEATION The liquid radwaste-tre tment system shall be CPERABLE and ap opri-3.11.1.3 ate ortions of the system shall beNsed to reduce the ra'dioective mate als in-liqu d waste prior to their discharge en the effluent, from the site tc UNRESIRICTE see Figure4EA$
5.1131)(projected would exceed doses l d4 J 0.06 mrem to the otal body or 0.2 mrom tex (ny organ in a 31-day period. APPLICABILITY: At times. \ , ACTION: .
- a. With radioactive li) d waste being dischar d without treatment and 4
- n excess-of the abov imits, prepare and su t to the Comiss} n thin 30 days pursuant Specification 6.9.2 Special Report which includes the followt information: *
- 1. Explanation of what liqu 11 radwaste was being d acharged without #
-treatment, identification y inoperable equipinent or sub-inoperability, \
s'ystems, and the reason for Act M ) taken to restore the operable equipment OPERABLE 2. status ,(etnd l Summary des iption of action (s) tak to prevent a recurrence. L 3.
\
N b. The provisions of Sp ification 3.0.3 are'no applicable.
) J
! $VPVEILLANCE RE0VIREMENTS ; x Ocsesdueto.liquidreleasesfqomthesitetoUNRdSTRICTL$ l 4,11.1.3. be projected at least'once-per 31' days in accordance with the AREAS shal methodology a parameters in the OOCH. \ 4.11.1.3.2 The iquid radweste treatment system shall be demonstrated .- t OPERABLE by metti Specifications 3.11.1.1 and '3.11.1.2. g i
. \
i Amendment No. 26 juh 2 21969
-LIMERICK - UNIT 1 3/4 11-6
- RA010ACT!viEFFLUENTS LIQUID'HOLOUP TANKS 3891002560 LIMITING CONDITION FOR OPERATION 3.11.1.4_ The' quantity of radioactive material contained in any outside temporary. tanks shall be limited to less than or equal to 10 curies, excluding
. tritium and dissolved or entrained noble gases. ,
APPLICABILITY: At all times. ACTION:
- a. With the qu'antity of radioactive r<nterial in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tonk and within 48 hours reduce the ,
tank contents to'within the limit and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release
- Report,
- b. The provisions'of Specification 3.0.3 are not a'pplicable. '
SURVEILLANCE REQUIREMENTS 4.11.1.4- The quantity of radioactive material contained in each of the above j- tanks shall be determined to be within the above limit by analyzing a repre-
- 1. sentative sample of _ the tank's contents at least once per 7 days when radio-active-m3terials are being added to the tank.
. s LIMERICK - UNIT 1
- 3/4 11-7 Amendmer ' - o. 11 g 7 g;
t i 1 RADIOACTIVE EFFLUENTS 3g()i002560
~3/4.11.2 GASEOUS EFFLUENTS' \x ,
DOSE RATE - ' LIMITING CONDITION FOR OPERATION 's N 3.11.2.1 The dose rate due to radioactive materi'als released in gase i effluents from the site to areas at or beyond.the SITE BOUNDARY (see' pus Figure 5.1.3-1) shall be limited to the-following:
- a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less.than or equal to 3000 mrems/yr t'o the skin, and
- b. For iodine-131, for iodine-133, for tritium,.and for all radio-nuclides in particulate form with half-lives greater than 8 days; '
L Less than or equal to 1500 mrems/yr to any organ, pathways only.) (Inhalation j p , APPLICABILITY: At all times.
- ACTION: '
h a.; With the do'se rate (s) exceeding the above limits, immediately restore I the release ratt to within the above lie'ts,
- b. The. provisions of Specification 3.0.3 are not applicable. l-t SURVEILLANCE REQUIREMENTS i fS l
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 methodology , Jand parameters of the ODCH. 4.11.2.1.2 The dose rate due to iodine-131, iodine 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in '
. gaseous affluents shall be determined to be within the above limits in accordance with the methodology and parameters of the 00CM by obtaining representative iamples and performing analyses in accordance with the sampling -
and analysis program specified in Table'4.11.2.1.2-1. ' LIMERICK - UNIT 1 3/4 11-8 Amendment No.11 WDV 713H i
.~.
y
^ TABLE 4.11.2.1.2-1 , t t ~ ~. C :
3 RADIDACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PIIOGRAM ** m , **.;
- -~. 5
/
f) s
/
i
, s Q , LOW [R LIMIT OF i MINIMUM. !
- i TYPE OF DETECTION (LLO)*
'5AMPLING ANALYSIS jf- E ,
ACTIVITY ANALYSIS (pC1/mt) ; GASEOUS RELEASE TYPE FREQUENCY FREQUENCY i G i,.
; y Principal Gamma EmittersDI P P I-
.5 Each Purge 1x10 *
. A. Containment Purge Each Purge i-
^1 (Pretreatment) Grab ,
Sample b . Principal Gamma Emitters' 1x10 *- l B. North Stack and M . i' South Stack Crab 1x10 *
'i ' '~- Sample H-3 :
4 t.s C I-131 1x10.ar . fi A C. Hot Maintenance ' Continuous W . I-133 Ix10 i Shop Ventilation Charcoal M $*"PI ' -
~- -
- ~'
j, s!> h Exhaust and All ,
'8 t
Release Types d C Principal Gamma Emitters' 1x10 Continuous W Listed in 8 Particulate (I-131, Others) - above - Sample 8' d Gross Alpha 1x10 ' I Continuous Q ,
' Composite Par-ticulate Sample d Sr-89, Sr-90 - 1xl"0 88 U
- Continuous q !
Composite Par- , ticulate Sampie I 4 Noble Gases 1x10 ' All Release Types Continuous Noble Gas l
;g 0. (Dased on 5, !
Listed in B above Monitor' Gross Beta or Gamma Xe-133) ', 1 m and the Main -'
, c Condenser Of f Gas 5i'
- Pretreatment Radio- -
activity Monitor > . - .e
~ ~ - ~ ~ -
~'
, , , , , ,, , ,, , . . . ~~.- .
l t i 7 \ , TABLE'4.'11.2d.2-1 (Continued) ( I n i 4
. TABLE NOT'ATIONS .
'N -
"The LLD.is defined, for purposes of these specif(cations, as the smallest concentration of radioactive material in't. sample that will yield a net count, above system background, that will be detected with 95% probability with only 1
5% probability ofs falselyxoncluding that a blank observation represents a I "real" signal. x x
\ \ i For a particular m'asurement e system, (which may include radiochemical separation): 'N N '
3
- \ N 4.'665 s l b
LW =- N , I m E V 2.22 x*1 % Y exp (-Aot) \ ! N xN \' l Where: N. .x ' l l LLD is the a priorblower limit of sietection as defined above (as - microcurie er unit mass or volume) \
\ ,
sDis the standard \ deviation of the back'g.round counting rate or 'of the counting rate of blank sample as appropr\igte (as counts per eute), E is the counting eff iency (as counts per di ntegration), x V is the sample size (in itsofmassorvolume) ' 2.22 x 10s is the number ofs#isintegrations per sin te per microcurie, \
)
Y is the fractional radiochem e 1 yield (when applic ), .
\
*\
A s the radioactive decay' constan fortheparticularrkdionucibe,and At is the elapsed time between midpoiht of sample collecti \ N and time of
- l counti (for plant effluents, not envi nmental samples)
The value fs used b in the calculation of\the Li.0 for a detection system ' shall be ba d on the actal observed variance bf the backgrouh4 counting , rate or of t counting rate of the blank sa es (as appropMate) rather than on an unv ified thecretically predicted -{ rience. , I
, . l Typical values of , V, Y, and at shall be used i tha calculation.
It should be recogni d'that the LLD is deffoed ts hn a priori (bef re the i fact) limit represent g the capability of a measureheiit system and not as I an a posteriori (after he fact) limit for a particul r measurement. N l 1
- r. e lid LIMERICK - UNIT 1 3/4 11-10 t
. .., .... ~., v .. ..>. .
~ ~ ' ~_ C
i
. )
. i
. 3892300520 !
I TAl([ }} 2.1.2 1 (Continued) [
\
- TABL! NOTAffCNS N
b SamplingandanalysesshallalsoDeperformedfollowin\ g etutsown, startue, l N or a THERMAL PowtR change exceeding 15% of the RATED THERWAL POWER within ! a 1 hour perige. This reewirement does not apply if (1) annlysis shows that the DOSE fQtt!YALENT l 131 concentration in the primary coolant has not i increased mohe than asf actor of 3; ans (2) the main consenser 'offgas pre- ; treatmentradQactivity'monitorshowsthateffluentactivityh46,not N locreased more sn_ an a f actor of 3. s c Samples shall be chenged at lea 2t once per 7 day ( and analyses shall te completed within 48 hqurs after cha ing, or af ter removal from sampler. ! Sampling shall also tex performed at east once per (4 hours f or at le'st ( 7 days following each shutdown, startup, or THERMAL YpWER change excee'ing c ,
, 15%,c '
ofch(anging. When RATED THERMAL samples egllected POWEksin for 24 hours 1 are hour and analyses analyted, the complete. corrasponcingsLL0s may be inchased by a f actor of 10. This recuirement ' does' cot apply'if (1) analysis s. hows that the 005E EQU! VAL (NT l-131 , , concentrationintheprimarycoo}enthasnotincreasedmore\thana f actor 'o has not i(ncreasec moresthan a f actor \cf 3.3; and (2hthe noble gasyenitor show , e TheratiocNnesampiefiow rate to tk sampiee stream new ra e shan be known fer tna time perio: covered by hch dose or dose rate calculation made in accorcante with specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.
\ \
e The principal gamma \ emitters for which the LLV specification applies incluce the fo11 ewing Kr 87, Kh-64, Xe 133, Xe-133m, Xe135,Xe-135mandXekracionuclides: 38 for gaseous emission 6 and Mn 54, Fe 59, Co SS, ; Co 60, 2n 65, Mo 99, I-Ill, Cs 134, Cs 137, Ce lhi and Ce la4 for para ticulate emissions. This\]ist does not mean that ynly these nuclides art to te consicered. Other gaisma peaks which are identifiable, together with those of the above nuclides,\ pall also be analy2ed'and reported in the emiannual Radioactive Efflue.. Release Report, pursdent to specifica- < t n 6.9.1.8. \
\
f Unce the provisions of footnote
\above, only noble gases need to te consi(red.
g Required ly when handling or storing irradiated fuel in the seconcary [ containmen h.
\
Required for tge hot maintenance shop v9ntilation exhaust only during crera- , tion of.the hot aintenance snop ventila, tion exhaust system. ,
'\
\ r LIMERICK UNIT 1 3/4 11 11 g g jggt[
RADICACTIVE tFFLtirNTS oost noitt casts 3892300520 LIMITING CONDITION FOR OpfRATION s 3.11.2.2 The air dose due to noble g rele6 sed in gaseo ffluents from the site to areas at and beyond the SITE BNNDARY (see Figure 511.3'l) thall l be11mitedtothefolla
- a. DuringanyblendArguarter: Less thaR or equal to 10 mrats for gamma radiat' n and lede s than or equal t 20 mrads for beta \rediation I b.
and During any calenc ' ear:
\Less' Dan or equa to to areds for \'gtema radiation and less an or equal tos 40 erads hr beta radiation.\
\ \
APPLICABILITY: At all times.
\ \'
ACTION:
\ '
a, 5th the calc'ulated air dose from radioactive rioble gasts in aseous e subs t to th vents exceeding ommission with (30 nydays, of theMbove pursuant limits to $ prepara ands spec a<1 Report which identifies the cause(s)pecification for exceeding \the(.9.2, a limit (H and defines tha,correctiv6 actions that have been taken to reduceteseleasesandtheproposehc'errectiveactionstobetaken to assure hat subsequent releases wit) be in comp)fance with\the
.above limit -
- b. NThe provisions f Specification 3.0.3 are not applicable.
\
SVRV LLLANCE REQUIREMENTS 4.11.2.2\umulative dose contrhutions for the current calender quarter and current cal dar year for noble gises shall be determined in accordance with the feethodolo and parameters in he ODCH at least once per 31 days.
\
i LIMRICK = UNIT 1 3/4 11*12 Amendment NO. 26 j@ 0 2 $$ ,
BAoI0 ACTIVE ErrtVENTS 3892300520 DOSE - 10 DINE-133, TRITIUM, ANDNtADIONUCLIDES IN PART LIMITING CONDITION FOR OPERATION Y ' 3.11.2.3 m 'N
't tritium, and all rhdionuclide'ssin particulate ,
er form than the SITE8BOUNDARY-(se days in gasebys effluents released, from the site eycnd to area Figure 5.1.321) x shall be s limited to the following:
- a. N s During organ and, any ca tndar quarter: 'lessthanNrequalto15 mre
- b. During any calenda year:
organ. Less than 'or equa\to 30 mrems to any ' A APPLICABILITY: At all times. \ \
's ACTION: N \
\
- a. Wi g thecalculateddosefrohthereleaseofiodine-tri ium, and radionuclides in' particulate form with hal -lives, iodin great r than 8 days, in gaseous \ effluents exceeding any{pf the a limits prepare and submit to the Commission within 30'de to Spec fication 6.9.2, a Special Report which identifies'ys, pursuant
' for exce ing the limit and defines thecause(s) been taken to reduce the releases a\the corrective actions that hav to be with thetaken abo to assure that subsequent releases will be in limits.
x N The provisions Specification 3.0.3 arp not applicable.
\
SURVEILLA.'E REOUIREMENTS 4.11.2.3 Cumu Stive dose contrib tions for the current calendar quarter and current calenda ' in particulate fo ear for iodine-131, iodine 133, tritium, and radionuclides - accordance with th with half-lives greater than 8 days.shall be determined in ' l 31 days. methodology and parameters in the 00CM at least once per
\
is LIMERICK - UNIT 1 jn; 2 15D ' 3/4 11 13 Amendment No. 26 l-1
3892300520 RADICACTIVE EFFLUENT $ YENTILATION EXHAUST TREATMENT SYSTEM , l LIMITING CONDITION FOR OPERATION \ . , N . 3.11.2.4 The VENTILATION (XHAV5T TREATM i $YST(M shall be OPERA 8LE and - appropriate por,tions of ht4ystem shall b used to reduce radioactive materials in galeous waste prior t thafA discharge w n the projected dos'es due to i gaseous (see Figure affluent release 5.1.3-1) whenfrom the% veraged ver site 31to argas days et and would beyond exceed 0.6 mkemtha $1TE 80VNDARY to any organ in a 31 day period. APPLICABILITY: At all times. 'N, ACTION:
- a. Wi and in excess of sous wasteprep e limits, beinge and ischarged submit without treatment, ion within to the\Commiss 30 ys, p suant to Speci cation 6.9.2, a 5petini Report which '
inci es the 'fellowing info ation: -
'N
- 1. {x lanation ofswhy gaseou radweste was being discharged without tre tment,. identification any inoperable eqpipment or >
subs tems, and the reason r the ineperability,
- 2. Action (t taken to r'estore th inoperable equipment to OPERABLE status, d
- 3. Summarydesriptionofaction(s)takentopreventarecurrence. -
T X Ihe provisions of ecification 3.0.3 a e not applicable, , i SURVE (LANC( REQUIREMENTS 4.11.2.4 1 Doses due to gaseous leases from the site to areas at and beyond l the SITE biundary shall be project at least once per 31 days in accordance ' with the me odology and parameters in the ODCH. 4.11.2.4.2 Tht VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstreted OPERABLE by mee'ing Specifications 3.1 .2.1, 3.11.2.2, and 3.11.2.3.
\
\
LIMERICK - UNIT 1 3/4 11-14 Amendtnent No. 26D0 2# 1
. . . . ~ . . . . . , , :. . .
. . .>..c~..,..
. v.
.....c.. . . , .....~....,c ...
. . . , u.
.: . .n.:.. .;:.,..
a .a.. .:... ... w e l
.es RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE 38910MBH i
_ LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of hydrogen in the main condenser offgas treatment i system shall be limited to less than or equal to 4% by volume. APPLICABILITY: Whenever the main condenser air ejector system is in' operation. , ACTION: a. With the concentration of hydrogen in the main condenser offgas treatment system excee' ding the limit, restore the concentration to within the limit within 48 hours.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS i 4.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits by continuously ! monitoring the waste gases in the main condenser offgas treatment system with
' the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specification 3.3.7.12. i LIMERICK - UNIT 1 3/4 11-15 Amendment No.11 L'07 7tiii
- . u. . . ; . , . .:. . , ;
1 [ i i l {
- 9 0 RAD 10ACT!vE EFFLUENTS l l MAIN CONDENSER .
LIMITING CONDITION FOR OPERATION i 3,11.2.6 The rate of the sum of the activities of the noble gases Kr-85m, Kr-87, Kr 68, Xe*133, Xe 135, and Xe 138 measured at the recombiner af ter-condenser cischarge shall be limited to less than or equal to 330 millicuries /second. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3*. I ACTION: With the rate of the sum of the activities of the specified noble gases at the recombiner after condenser discharge exceeding 330 millicuries /second, ' restors the gross radioactivity rate to within its limit within 72 hours or be in at least HOT SHUTOOWN within the next 12 hours. . SURVEILLANCE REQUIREMENTS 4.11.2.6.1 The rate of the sum of the activities of noble gases at the recombiner after condenser discharge shall be continously monitored in accor-dance with Specification 3.3.7.12. 4.11.2.6.2 The rate of the sum of the activities of the specified noble gases from the recombiner after condenser discharge shall be determined to be within the limits of Specification 3.11.2.6 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the recombiner after condenser discharge:
- a. At least once per 31 days,
- b. Within 4 hours following an increase, as indicated by the Main Condenser Off Gas Pretreatment Radioactivity Monitor, of greater than 50%, after factoring out increases due to changes in TH(RMAL POWER level or air in-leakage, in the nominal steady-state fission gas release from the primary coolant.
- c. The provisens of Specification 4.0.4 are not applicaole.
k "When the main condenser air ejector is in operation. LIMERICK - UNIT 1 3/4 11-16 ggg g gg
. ..... . ;.- ..:. .. v . .:.. . .~ . ...
RADICACTIVE EFFLUENTS VENTING OR'PURGING rN 389i002560
'\
'\ \.
LIMITING CONDITION FOR OPERATION , \
~ \ \
3.11.2.7 VENTI (G or PURGING of the Markh containment shall be through the standby gas treatment systema s APPLICABILITY: h enever the con {ainment is ve ed or purged." ACTION:
~
\
N
' a. With the reQu (ements of the above specification not satisfie suspend all VENilNG and PURGING of the contal ent.
PN b. The, provisions of ecification 3.0.3 are not an licable. s
\
SURVE!L ANCE REQUIREMENTS . . - 4.11.2.7.1 containment shall be det mined to be aligned for VENTING or
\
PURGING throug the standby gas treatment ystem within 4 hours prior to i tart of and at ast once per 12 hours durt g VENTIN3 or PURGING of the i
' tainment.
'N 7.2 Prior to 4.11. e of the purge system thraugh the standby gas treatment system ssure that:
a, th standby gas reatment system trains a OPERABLE whenever the p e system 1s in use, and l l b. Whene er the purge sy tem is in use during OPERATIONAL CONDITION 1 l or 2 o 3, only one of he standby gas treatment system trains may l be used. i s i' x
's
*Except for the one inch /two inch vent valves to the Reactor Enclosure i Equipment Compartment Exhaust Filters when used for containment pressure ~
control and nitrogen make-up operations.
\
LIMERICK - UNIT 1 3/4 11-17 Amendment No.11 py 7 tyg
\
\ -
- J Ga . c.;G G :.b.r.w..L.?.G:?.. :h. ~ 1
. ' .? :.. . ... . . . . . ,
i RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RA0 WASTE TREATMENT , 3891002560 s' LIMITING CONDITION FOR @> ERAT 10N s
'\, 'N, ,
3.11.3 Radioactive' wastes'shall be SOLIDIFIED or dewatered in'accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requi,rements during trans t, and disposal site requirements when received at the' disposal site. ' APPLICABILITY: t all times. , ACTION:
* * \
s S ' x 'N
- 4. With SOLIDIFlCATION or dewatering not meeting disposal site and
-hipping and transportation requirecents, suspend shipment of the nadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the' procedures and/or the solid waste' system as necessary to prevent recurrence. N ,
- b. With the SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) test the improperly processed waste in each container to ensb qquirementsand(2)take(ethatitmeetsburialgroundandshipping appropriate administrative action to prevent recurrence. N s
- c. The
\
ovisions of Specificatiom 3.0.3 are not applicable. SURVEILLANCE REQUIR ENTS
\
4.11.3.1 The PROCESS NTROL PROGRAM shall be used to verify that the properties of the packaged waste me t the minimum stability requirements of 10 CFR Part 61 and other h quirements fo transportation to the disposal site and receipt at the disposal \ site.
\ .
l LIMERICK - UNIT 3 3/4 11-18 Amendment No. 11 NOV ? Illi L
~.
.c.
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, . . . . e ,.3 g:.. n.g ;w: n .
. .~ ... . ..:. ..e.
3891002560 i RADICACTIVE EFFLUENTS
\
SURVEILLANCEREQUIREMENTS(Cohtinuedl 4.11.3.2 li t SOLIDI ICATION method is used, the Pf0 CESS CONTROL PROGRAM
\ \\
shall be used *. verify the SOLIDIf1 s CATION of at least che representative test specimen from at least every tenth b (e.g., filter slu 'as, spent resins,%ch of each type of wqt radio, active waste evaporator bottoms, an sodfum sulfate solutions). ' the SOLIDIFICA-If any test speciesen N of the tch under test fails shalto\beve ify SOLIDIFICATION, suspended unti1 \such time ad Monal test specimens can be obtbined, alte native SOLIDIFICATION parambers can ,
-determined in accor(ance with the PROCESS CONTROL PROGRAM,'a.nd a s sequent test verifie SOLIDIFICATION \ SOLIDIFICATION f the batchmay t en be resumed using(the alternative SOLIDIFICATIO p rameters det4rmin by the PROCESS CONT L PROGRAM. .
- b. If t initial st sp cimen from a batch o waste fails to verify SOLIO ! CATION thesPROB(S$ CONTROL PROGRAM s 11 provide for the s
collect h n anc, testing o representative test 6pecimens from each consecutih batch of the s type of wet waste'yntil at least three consecutivesinitial test \ p imens 4 demonstrate SOLIDIFICATION. The PROCESS CONThQL PROGRAM sha{ be modified as required, as provided in Specificatfan '6.13, to ass re SOLIDIFICATION of subsequent batches of waste.
'\,
i l' i L . r (. b AUS 8 1385 LIMERICK - UNIT 1 3/4 11-19 o
s , .. ,.v.. _ , , . t. ...:. .u . . m .o r v . ~. w :v.a. . w-. u n a .%. .. ~ ~ ~
.. 'a>' - '
389100360 RADIDACTIVE EFFLUENTS i
^
3/4.11.4 TOTAL DOSE _LIMITINPCONDITION FOR OPERATION \ ' 3.11.4 The annual (calendar ye,. i dose k tdose commitment t any MEMBER OF THE PUBLIC, due to reresses of radioactivity and radiation, fr uranium fuel cycle. Sources shall be limited to less than or equal to 25 mrem to the total i body equal to or 75 any organ, except"the thyroid, which shall be limited to ess,than or mrems.
\
APPLICABILITY: Atalltimes\ ACTION: a. With the calculated doses from the releut o'f ra .oactive materials inliquidorgaseouseffluentsexceedingtWicethe\limitsofSpecifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.24., 3.11.2.2b\ 3.11.2.3a., or 3.11.2.3b., calculations shall be made including d contributionsfromthereactorunitsandfromoutside(rectrediation ttorage tanks to determine whether the above limits of Specification A 11.4 havr
- been exceeded. If\such is the case, prepare and submit t's the Commission within 30 days, pursuantuto Specification 6.9.2' a Special i Report that defines \the corrective attion y to be taken to re uce subsequent releases jto prevent recurrertte of exceeding the a ve limits and includes,the schedule for achttving conformance vi the above limits. This 'Special Report, as def tged in 10 CFR 20.40 shall include an analysis that estimates theNradiation exposure ,
(dose) to a HEMBER OF THE PUBLIC from uranium' fuel cycle sources, including all effluent pathways and direct radiktion for the cal dar , year that includes ther,elease(s) covered by this\ rep, ort. It shal also describe levels of 'Mdiation and concentrationf of radioactive material involved, and the'gause of the exposure le9els or concentra-tions. If the estimated dost(s) exceeds the above li'inits, and if the release condition resultirtg in violation of 40 CFRyart 190 has not already been corrected, the'Special Report shall inc ude a i \ request for a variance in accordancef 40with CFR the provi l Part 190. Submittal of the report \is considered a timely equest, ahdavarianceisgranteduntilstaf(actionontherequestis complete. N
\
- b. The\revisions cf Specification 3.0.3 are not applicable.
S0RVEILLANCE RE UIREMENTS i s 4.11.4.1 Cumulat shallbedetermine(vedosecontributionsfromliquidan5\gaseouseffluents j 4.11.2.3, and in ac(cordance with the methodology and parameters in the ODCM 4.11.4.2 If the cum ative dose contributions exceed the l'imits defined in 3.11.4, ACTION a, Cum ative dose contributions from direct Yadiation from unit operation shall b determined in accordance with the methodology and parameters in the ODCM. J LIMERICK - UNIT 1 \ 3/4 11-20 Amendment No. 11 A?/ 7[m l i 1 - -
- . : v _ ~ : - . . -
. y..y- . . . . .
p .., j . y ;;;;,~.*-l,';. ;.;.n;.,. 3 .? . . ,:. , , : . ' , ' l 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ' 1 3/4.12.1 NNITORING PROGRAM ' l LIMITING ':0NDITION FOR OPERATION \ i 3.12.1 Tieradioloicalenvironmentalmonhtcringprogramshallbeconducted as specified in Tab e D 2.1-1. \ ' APPLICABILITY: At all tjmes. ' ACTION: , , ., ,\ ,
- a. With the radiologi Q environmental monitorihg program not being conducted as specified in Table 3.12.1-1, prepere and submit to the l Commission, in the Annual Radiological Environmental Operating i Report per Specification'6.9.1.7, a description 'fo the reasons for ,
l not conducting-the program'as required and the platis for preventing once. b. a rt K Wiih the levekof radioactivity \as the result of plant \ effluents in an k(ivironmentalNampling medium'at a specified location exceeding the reporting levels'of Table 3.12>1-0 when averaged over any calendar quarter prepare and s'ubmit to the Carnission within 30 days, pursuant toSpeckication6.9.2,a~SpecialRep%tthatidentifiesthecause(s) j for exceed (ng the limit (s) and definen vie corrective actions to be takentore69ceradioactiveeffluentsto%hatthepotentia1 annual \ dose to a MEMEER OF THE PUBLIC is ess than the calendar yeat limits of Specificati6qs 3.11.1.2, 3.11.2. and .'.) .2.3. When more\than one of the radiohuclides in Table 3.12.1-2 0 detected in this sampling medium, this report shall be submitt if: N - w concentration (i') reporting level (1(reporting
- concentration level (2) (2 b + .. d 1.0 When radionuclides other than those in Table 3.12.1-2 are detected and are the result of plants effluents, this report shall be submitted-if the potential annual dose,to a MEMBER.0F THE PUBLIC is equal to l'
or greater than the calendar year liraits of Specificnions 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 such an event, the condition shall be reported and described in the Annual Radiological Er.vironmental Operating Report. l . LIMERICK - UNIT 1 3/4 12-1 M 6
- i RADIOLOGICAL ENVIRONMENTAL MONITORING 3891002560 i j
i LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) 'N s
's_
- c. With milk or fre leafy vegetable samples 1.iavailable from one or more of the sampi locations required by Table 3.12.1-1, identify teettions for obta Ring replacement samples ahd add them to.the i
radio 1'o'gical environ %ntal monitoring program sithin 30 days. The ! spe:ific 16 cations fr g which samplesswere unava{1able may then be deleted fr6msthe monitoring program. Pursuant \to Specifica- l ion 6.9.1.8,idsntifytQcauseoftheunavailabilityofsamplesand l entify the new lo'ation(s) e for obtaining replacemhnt samples in the ) Tie -Semiannual Radioactive \ Effluent Release Report knd also include l intereportarevised'figuFgs)andtablefortheOCMreflecting the n location (s). N s
- d. The prov ions of Specification 3.0.3 are not applicable. :
a 1 SURVEILLANCE REQUIREME S 4.12.1.1 The radiological nvironmental monitoring samples shall be collected pursuant to Table 3.12.1-1 om the specific locations given in the table and figure in the ODCM and shall e analyzed pursuant to the icq.'ieeme.it of-Tables 3.12.1-1, the detectio capabilities required by Table 4.12.1-1. LIMERICK - UNIT 1 3/4 12-2 Amendment No.11 ,62 7 7 lid
...e..,.....~ .. . . . . . . . . - .
. ... .g.. . . . . , . . . . .. .
- a. ... .-
s . n\ .
' i. .
{. L -- TABLE 3.12.1-1
, - tC C '
x RADIOLOGICAL ENVIROMENTAL MONITORING PN0 GRAM 4 g [ '. '.
. E NUpBER OF REPRESENTATIVE f'.
. EXPOSURE PATHWAY SAMPLES AND c AND/OR SAMPLE SAMPLELOCA,110N5$
SAMPLING AMO TYPE AND FREQUENCY "Y COLLECTION FREQUENCY OF ANALYSIS , b-
- 1. DIRECT RADIATIONI) 40 routi iteriqi s tions At least Quarterly [ Gamma dose at least
.- th two or more simeters- [li i-i either or w one instrument f , meanuring quarterly. [ ., ..
i ?! i a recording dose rate o d inoously ^ li ~' aced as follows: (1? inner ' I ring of stations, one. In ach , , ! .
; meteorological sector, i the general 5
area of the SITE F,0UNDA ; (2) An outer ring of str.t.fons, one in each j [ t r ?
'e
- / l
, meteorological sector, n the 3-to '/ 8-R 9-alle range from the ite; (3) The , / ! j., .:.
balance of the stati ' s placed in '- . . . . special interest a
/ { ~U s such as U population centers,, arby resi' /' . [ ..*.
J. / - dences, schools, and in 1 or,4 l ' T
- ['
, . . 4
^ ~
areastoserveas/ontrol tations. [. q
- 2. AIRB0kME Radiciodine and
[. Samples from 5 loca fons: Continuous sampler ~ Radiofodine Cannister: , Particulates operation with sample I-131 analysis weekly. : . /, ! I a.3samplesf[r close to the collection weekly /or
! 3 SITE BOUNDARY locations (in more frequently if -
{ i:] i different sectors) of the requiredbydusI Particulate Sampler: [
} highest calculated annual loading. Gross beta radio i .,.
average groundlevel x/q. activity analysis
~
- b. I sample from the vicinity -
following filter ' . community.having one of the I.~
. highest calculated annual g* , d highest groundlevel X/Q. g,,,,jsgtopic m y sgs og y f -j composite (by
?'
# location) at least *I-- ,,' i .
quarterly. j' [I
*E 5 *.
- 5.S.
. 4 . -';
i . c.
}
's
. - -- y w'w -
---wrm .a.,-e. , eey-,---q,
. i' i -
y I r- TABLE 3.12.1-1 (Continued) k, f*
. = RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3,
l U
- x. -
NUPEER OF REPRESENTATIVE c- EXPOSURE PATHWAY SAMP J Cit,ND SAMPLING AND 5 AND/OR SAMPLE SAMPLE 40 CATION}S S) TYPE AND FREQUENCY COLLECTION FREQUENCY Of ANALYSIS . y 2.
- AIR 80RNE (Continued)' / k I shie from a control loca 2 '. t n, as for example 15-30 km istant and in the least pre !
- l. ',
valent wind direction
-- ~- ~
} ;-
,j *
- 3. WATERBORNE
. *y
- a. Surface I 1 sample upstream. omposite sampi
~
i I sample downstreas. Gamma isotopic analysis'
- 1-monthperiod.gover monthly. Composite for g'
m
/ tritieme analysis quarterly.. i ,*
b l ' g b. ound Samples from 1 or 2 sources Je 7 Quarterly. h a isotonic and tr W um only if likely!.tobeaffectedh
; ~ n t' l
/
- c. Drinking 1 sample of each of 1 /3 Monthly. (Composite) analysis /usrterly.
I-131 analysis on each 1
,j of the nearest water upplies [ J,
- composite when the skese
- that could .ed by its I
discharge.,be aff ca}culated for the con
~
sumption of the water . ( is greater than I mrem 3 1 -
; per year. Composite ;
I sample from a control , location. for gross beta and gamma ( ,
, Isotopic analyses monthly. t' Composite for tritium , j~
analysis quarterly. i -)
. d. Sediment I sample from downstream area Semianovelly.
from Gamma isotopic analysis' with existing or potential semiannually. s shoreline recreational value. , I. 3
. .h
$ i ,. 9 I
g .
- s_ --+_m _ _ _ _ _ _ _ _ _ _ - . ,_ -___..a
i TABLE 3.12.1-1 (Continued) , . . t * . E ' RADIOLOGICAL ENVIROf91 ENTAL MONITORING PROGPAM ~
. .-
- h[
O NUPEER OF REPRESENTATIVE
- EXPOSURE PATHWAY SAMPLES AND
. +
., SAM 7 LING AND TYPE AND FREQUENCY i I g ANU/OR SAMPLE SJnMPEG. LOCATIONS (,)
COLLEC'. ION FREQUENCY OF ANALYSIS -
).
4 4. INGESTION \
' i~ h
- a. Milk k
- a. Samples frost milking animals Il i,
- Semimonthly when ani Gamma isotopic and I-131 .- 7 In 3 locations within 5 km mais.are on pasture,. analysis semimonthly when '
I distancehapingthe. highest monthly at other times. animals are on pasture; a dose potential. If there are monthly at other time,s. . . I none; then,/1 sample from allk- '$'; l . inganimals/ineachof3 areas * ^; '
- between 5 ;to 8-km distance
' -I-. .
I - where doses are calculated to I e q greater then 1 mrem per y l . g I sample f m all animals j - ' e at a contr61 1 ation (15-30 kna / / km distant and in the least
/ .;
I prevalent wind direction). / - 7' .
/
-! b. Fish and a. I sample of two recreationally Sample in' season, or ^ Gamma, sotopic e -;
~
Invertebrates important species ire vicinity sentannually if they a-,1ysis on edible .
, of plant discharge area. are not seasonal. ,rtions. :. ,;
1 i I sample of same species in . l,.' a, areas not influenced by platit , 4 discharge. l
.. J
. s e-
. t . . , .
- - - . . i .
3 . .
- ~;
- \
i
, y 7 , , ..m c., .a w - , , + . . _ . _ - - , _ -ow.-- . _.___.__.___s m.s
i :
.
- 1:
; ?.
- i. -
l jy, TABLE 3.12.1-1 (Continued) . A-l 5 . - i- ^ 'd . p -RADIOLOGICAL ENVIRONMENTAL MONITORING PRobRAM ; ; e,
- sy
) - c . ! z ER OF RE SENTATIVE - ' l U l- - EXPOSURE PATHWAY SAMPLE; AND SAMPLING A.10 I- l ( w AND/OR SAMPLE . SAMPLE LOCATIONS I *) COLLECTION FRE(,'IENCY TYPE AND FREQUENCY OF ANALYSIS i 4. ' l ~7 #' INGESTION l .c1 (Contin \ f', - I .c. ood Products Sample of 3 different Monthly when available, Gamma isotopic
- arwt !i l l kinds of broad leaf if milk sampling is I-131 analysis. I:
( :j vegetatfori grown nearest not perf . h} ", 1 j each of t different ; . l._ offsite 1 tio,i:: of highest
- w predicted anual average - - l .
1 groundlefelD/q1faiIk I-
-l sampling'is not performe V - ,- I samp19 of each the siellar - Monthly when available. Gas q isotopic' a..d -
l'.
; broad leaf v tion grown if alik sampling is I-1 1 analysis. 4.
! 15-30 km d ant in the least not performed. '
I '.! prevalent wind direction if ! ; allk sampilng is not performed. ,;
/ 3 !
/ - .! h.
i
/
/ 5 ; ii:
- / .
~
l
. . .,/. . - -
t c.-
./ s 3 o !
c'? . 8 s. i
- l. '
M - ; e .
*\
8
. _ e, . 7 .. . - - - -
.. -t w#*. de .24t*-***=
,, .- .a m , # .-- -- - . . , . . . , . - . - e- , . - . ..,,-<.r, - - - , , e---_,-,-w.#- . .~.r .yr. . -
. . -.:. ~. :.: ::
^
'.T. . . . . . . . . ... . a ..
(
' RADIOLOGICAL ENVIRONMENTAL MONITORING -
[ TABLE 3.12.1-1 (Continued) -
+
r , TABLE NOTATIONS
" Specific parameters of distance and direction sector from the centerline of the two reactors and additional description where p,ertinent, shall be provided for each and every sample location in Table 3.12.1-1 in a table and figure (s) in the ODCM. Deviations are permitted from the requir sampling schedule if r specimens are unobtainable due to hazardous conditions, seasonal unavailability,
'mOfunction of automatic sampling equipnient and other le'0 timate reasons. If
' specimens are unobtainable due to sampling equipment malfu(getion, 'every effort
- shall be made to complete corrective action prior to the en4 of the next sampling '
period. All. deviations from the sampling schedule shall be documented in the Annual Radiological Environmental OpCat 4'$eport pursuant to\ Specification 6.9.1.7.
~ ;
Itisrecognizedthat,attimes,itmaynotbep'ossibleorpract4cabletocon- . tinue to cbtain samples of the medi of choice at tht~most desitetl location ; or time. In these instances suitabl alternative media'and locatihns may be '
' chosen for the particular pathway in question and appropriate subst) utions
, made within 30 days in the radiologic 1 environmental monitoring g pro ((am.
Pursuant to Specification 6.9.1.8, ide tify the cause of the unavailabKlity . of samples for that pathway and identif the new location (s) for 'obtaint g
' replacement samples in the next Semiannu 1 Radioactive Effluent Release ort
' and also include in the report a revised igure(s) and table for the ODCM reflecting the new locationfs). ,
D
- l. one or more instruments, such as a pressurtz ion chamber,'for m'easuring and l
[ recording dose rate continuously may be used place of, or in addition to,
- integrating dosimeters. For the purposes of t is table, a thermoluminescent l dosimeter (TLD) is considered to be one phospho ; two or more phosphors in a
' packet are consideJs4-ai.4wo or more dosimeters. Film badges shall not be used as dosimet s for measuM ng direct radiatio .
x C Methodology to recovery of radiciodi e'shall-be' iescribed in the 00CM. d Airborne particu te sample filters shall be analy2'ed for gross beta radio-
. activity 24 hours .r more after sampling to allow for radon and thoron daughter decay. I ross beta activity in air particulate samples is greater than 10 times the ye ly mean of control samples, gamma isotopic analysis shall be performed on he individual samples.
' Gamma isotopic analysis e ans the identification and quantification of gamma-emitting radionuclides tha 'may be attributable to the effluents from the facility..
I The " upstream sample" shall be take'n at a distance beyond significant influence of the din:harge. The "downstr am" sample shall be taken in an area beyond but near the mixing zone. t LIMERICK - UNIT 1 3/4 12-7 E 5 '.T -
.g 4
=
': . . .. , e . . . . . , - . , . . .
.. . . . . . : . . - . .: . .. + .... .
]
. 1, ,
RAD *0 LOGICAL ' ENVIRONMENTAL MONITORING I TABLE 3.12.1-1 (Continued) 1 _ TABLE NOTATIONS i 9 A composite samp1, 1 e quantit (al uot) of liquid sampled is proportional to the quantity M ing liq id an N ni which the method of sampling employed results in a specimen tytt is representative of the liquid e l flow. 'In this pro am composite sample alitruots shall b'es collected at time intervcis that are r +a tb compositing < 1 period (e.g. , month 19)y short (e.g. in order to assure obtaining , hourly) thati"*pret ntative sample. I h Groundwater samples sha 1 be taken when this source is. tipp'ed to drinking or , irrigation purposes in a eas where the hydraulic gradient oNrech ge properties are suitable for containi ation.
, / -
I The dose shall be caleclated for the maximum organ and age group, using the methodology and parameters in the ODCH. [ . 6 l l 4 l%f g"...." LIMERICK - UNIT 1 3/4 12-8 -- g
s ,
, .~
.l r
TABLE 3.12.1-2 ;'-
- ~] .
Cr .
.e m REPORTING LEVELS FOR RADI0 ACTIVITY C'4ICENTRATIONS IN ENVIRONMENTAL SAMPLES 5-a 7 Reporting Levels e .
.N. ,^,
c . z n
-~ G Al '2TICULAE FISH - MILK F000 PRODUCTS ,-
WATER-i g (pCf/kg, wet) (pCl/L) (pCl/kg, wet) ,; ANALYSIS . (pCf /8 ) or pCf/u5) \ ; 7,
~
^
H-3 20,000* >e
~
30,000 -I Mn-54 1,000 : l 10,000 _
'1 .
j Fe-59 400 i
. ! 30, .'[
s.,') Co-58 1,000 -
, r
' ~
.t } 300 0,000 Co-60 i
, g m. ,N -
. 1 .20,000 - .. . . '
- Zn-65 300 . .
i- '. Zr-Nb-95 400** . 3 100 ..," I-131 2 0.9 .j
~
1,000 60 1,000 , Cs-134 30 10
! Cs-137 50 20 2,000 70 /
/ 2,000 I 300 / .
Ba-La-140 200** .- -
, 1 ~
./
I g er pathway exists, a value of
*For drinking water samples. This is 40 CFR Part 141 value. If no drinkt '
-- 30,000 pC1/L may be used. ..
t.:
** Total for parent and daughter. ~
lr g
- t i .
. ,L
. . 9 y - -
. _ . - - ~ ~ ...
w _ . . _ _ _ ..
-i- . .
- .I -
, i
. g.
i~ M TABLE 4.12.1-1
- o
~ '.
.l Q DETECTION CAPA8ILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS I *)
h z LOWER LIMIT OF DETECTION (LLD)(b)(c) ' M
- [T ~
WATER IRBORNE PAR'TICULATE FISH MILK F000 PRODUCTS SEDIMENTS-
; ANALYSIS (pCf/L) OR GASES (pCif/m8 ) (pCi/kg, wet) .. (pC f /L) (pCi/kg, wet) (pCl/kg, dry) t Gross Beta 0.01 e l
, a i *
- I* 11- 3 2000 / .E, Mn-54, 15 - ! ~
130 -
/ i I w. -re-59 30 / 260 7
Co-58, 60 15 / 130
. ; 7 Zn 30 60 h -
; t; Zr-95 30 .. . . . .
. Nb-95 15
$ I-131 Id)
. Cs-134 I
15 O. 0.05 130 1F 1 60
~60 150
? . Cs-137 18 0.06 150 18- 80 180
>l Ba-140 60 60 t
.La-140 15 15 i 1i . ' '
- 3. '
ki i, ! <- . .
*h
- i;e 4 I .
.? , I t
. i-
, , , , . . . _ . . . , _ , . - ~ _ , -. .,mi- . . _ . . _ . - _ . . ,
__. . ~ _ . _.
. . . ... s .. :.. .u..._...-....-. . . . . . . _ _ . . . , . . . . _ _
~
t TABLE 4.12.1-1 (Continued) .
. TABLE NOTATIONS ,
.(a)This list does not mean that only these nuclides to'be considered. Other peaks that are identifiable at 95% confidence level, ogether with those of theabovenuclides,shallalso.
logical Environmental Operating bear.alyzedandreporte}intheAnnualRadio-report pursuant to Spec ication 6.9.1.7.
,(b) Required detection capabilities for thermoluminescent dosime ers used for environmental measurements are gi - . in Eagulatory Guide 4.13.
(C)The LLD is defined, for purpos ofthesespecifications,asthes liest con-
\
centration of radioactive mater al in a sample that will' yield a ne count (above s only 5% ystem background) that w 11 be detected with 95% probability ith a "real" probability signal. of falsely co cluding that a blank observation-teg, sents
' For a particular measurement system hich may include radiochemical
, separation): - '
4.66 s b ' LLD = E V 2.22 Y exp(-Aat ' I where LLD is the "a priori" lower limit of detec ion as defined above (as or; vol ume), I picoc0 ries p [er unit mas } s s is the sgandard deviation Cf t6i background counting rate or of the c5unting rate of'a blank sample as appropriate (as counts per minute), l E is the ce nting efficiency (as counts per disintegration), V is the sam e size (in units of mass or volume), l 2.22 is the nu er of disintegratior.s per minute per picocurie, l Y is the fraction 1 radiochemical yield (when applicable), A is the radioactiv decay constant for the particular radionuclide, and At for environmental mples is the i.apsed time between sample collection (or end of the sar.ple 11ection period) and time of counting. Typical values of E, V, Y, and t should be used in the calculation. , j'" p r
"~
LIMERICK - UNIT 1 3/4 1 -11 't
v... ..s. . . . , ,, e . . . . . . . . . . .. . . , . ..
- .. ..e.. ..,.2..,
- ......
. .. ...-4 .
, Table'4.12.1-1 (Con )
TABL _ATIONS It should be recognized th,,t the LLD is defined as a prio.i (before the fact) limit representing trie capability of a measuremtht syst and not as an a posteriori (after.the fatt) limit for a particular measuremen . Analyses shall De pe # ....u in sIIch a manner that the stated LLDs w'ill be chieved under routine e itions. Decasionally background fluctuations, unav.oi bly small sample siJes, the presence of interfering nucliots, or other uncoqt ollable tircumstat ees may render these LLDs unachievable. .In such cases, t contrib-uting fact rs shall be identified and described in the Annual Radiological Environment 1 Operating Report pursuant to Specification 6.9.1.7. (d)LLD for dr king water samples. - l l 1
- i l
l 1 LIMERICK - UNIT 1 3/4 12-12 '%
.. : ^ ~" ~~ ~ .
~ ~ ~ " - ' ' ~ " * '
.~.,.~T* -
RADIOLOGICAL ENVIRONMENTAL MONITORING s891002560 ! 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 Alandusecensusshallbeconductedands\ha identify within a ' distance of 8 km (5 miles) the location in each of theN 6 meteorological sectors of the nearest mi M 1 he nearest residencNand the nearest , garden
- of greater than,50 m2 (500 ft q ucing broad leatf vegetat. ion.
APPLICABILITY: At all imes. \. x etCTION: ,
- a. With a land use ensus identifying a location (s) which yi ids a , l calculated dose o dose commitment greater than the values currently being calculated i Specification 4.11.2.3, identify the new loca- ,
' tion (s) in the next Semiannual Radioactive Effluent Release Report, t pursuant te Specifi tion 6.9.1.8. >
- b. With a land use censu identifying a location (s) that yields a calculated dose or do e commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordancel with Specification 3.12.1, add the new loca-tion (s)totheradiolfgicalenvironmentalmonitoringprogramwithin 30 days. The sampling location (s), excluding the control station location, h ing4he lowest calculated dose or dose commitment (s)
( (vi same exposure pathway) may be deleted from this monitoring gram after October 31 of the year in which this land use census as conducted. Pursuant to Specification 6.9.1.8, identify the.new Tocation(s)inthenextSemiannualRadioactiveEffluent and also include in the report a revised figure (s) and table for the DCM reflecting the new location (s),
- c. e provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE EQUIREMENTS 4.12.2 The lan use census shall be conducted during the growing season at least once per 1 months using that information that will provide the best results, such as a door-to-door survey, aerial survey, or by consulting local agriculture thorities. The results of the land use census shall be '
included in the Ann 1 Radfological Environmental Operating Report pursuant to Specificatio,n 6.9. 7.
- Broad leaf vegetation s .pling of at least three different kinds of vegetation may be performed at the ITE BOUNDARY in each of two different direction sectors with the highest predicte D/Os in lieu of the garden census. Specifications for broad leaf vegetation ampling in Table 3.12.1-1 item 4.c. shall be followed, including analysis of contrh1 samples.
LIMERICK - UNIT 1 3/4 12-13 Amendment No. 11 NOV 7 IS!!
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l i' I RADIOLOGICAL ENVIRONMENTAL MONITORIN l 3891002560 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM j l LIMITING CONDITION FOR OPERATION \ 3.10.3 Analyses shall eperformedonradio\ tivt materials supplied as i part of an Interlaborat y Comparison Program c Q as been approved *by the Commission. APPLICABILITY: At all time ,. 1 ACTION: , , ,
- a. St Q not bein erformed as required above, report the (corrective actione- tak to prevent a recurrence to the Commission I in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
- b. e provisions of Specification 3.0.3 are not applicable g ,
SURVEILLANCE QUIREMENTS 4.12.3 The Inter boratory Comparison Program shall be described in the ODCM. A summary of the re uits obtained as part of the above required Interlaboratory Comparison Program s all be included in the Annual Radiological Environmental Operating Report purs ant to Specification 6.9.1.7. [ LIMERICK - UNIT 1 3/4 12-14 Amendment No. 11 p
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9 INSTRUMENTATION . BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolatio.n cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment. Operation with a trip. set less consefvative than its Trip Setooint but within its specified Allowable Value is acceptable on the basis that the dif-
- ference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls
.and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation.
The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trio Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION
- The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation livel, trip setpoint is exceeded; and (3) sufficient information
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is available on selected plant parteu rs to monitor and assess these variables following an accident. This capabi. 6y is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64. . 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in'the design basis for the unit, m 3/4.3.7.3 METEOROLODICAL MONITORING INSTRUMENTATION ine OPERABILITY of 'tne meteorological monn.oring instrumentation, ensures
\
that sufficient meteorolo'gical data is available'for estimating potential radir tion dos'es,to the puolic as 'asqesult of routine or' accidental release ofxradio
- active materials to the atmosphete. This capability'is required to evaluate thethe of need for initiating public. protective' measures This instrumentation'is to protect consistent with th'e the recommendations health s
and safet)\oJ Regulatory Guice 1.'21 s ' "Onsite Meteorolodcal Programs," Febr0ary,1972.
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INSTRUMENTATION BASES , METEOROLOGICAL MONITORMG INSTRUMENTATI (Continued) -- N y Site data compiled since January 1972 p vide correlation tion 1 (T'bwer 1) and Elevatio , (Tower 2), an etween Elevation tween Eleva-- (Tower 1) and Elevatio N2 (Tower 2). This errelation serv as justification or the
'use of the app riate Tower 2 ins strument as shown,in Table 3.7.3-1. \u.nentasaback \ptotheTower1 ~
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,3f4.3.7.4 REMOTE SHUTOOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and mainte-nance,of HOT SHUTDOWN of the unit from locations outside of the control room. 1 This' capability is required in the event control room habitability is lost and is consistent with General Design Criterion'19 of 10 CFR Part 50, Appendix A. '
3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that - sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is con-sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation i for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. 3/al3.7.6 SOURCE 13ANGE MONITOR *i The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup - and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the inter- ; mediate range. monitors are on scale, adequate information is available without the SRMs and they can be retracted. 3/4.3.7.7~ TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the= traversing in-core probe system with the specified. minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e., detectors) prior to performing an LPRM' calibration function. Monitoring core thermal . limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demon-strated by comparing the detector (s) output' in the resultant heat balance calcu-lation (P-1) with data obtained during a previous heat balance calculation (P-1). LIMERICK - UNIT 1 B 3/4 3-5 C^ - t.g c
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INSTRUMENTATION - BASES . 3/4.3.7.8 CHLORINE'AND T0XIC GAS DETECTION SYSTEMS . The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly ' cnd the necessary protective actions will be automatically initiated for chlo- l rine and manually initiated for t.;xic gas to provide protection for control room personnel. Upon detection;of a high concentration of chlorine, the control . room emergency ventilation system will automatically be placed in the chlorine - . isolation mode of operation to provide the required protection. Upon detection -
.,of a high' concentration of toxic gas, the control room emergency ventilation i system will manually be placed in the chlorine isolation mode of. operation to .
provide the required protectionc The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95 " Pro- . tection-of Nuclear Power Plant' Control Room Operators against an Accidental Chlorine Release," february 1975. . 3/4.3c7.9 FIRE DETECTION INSTRUMENTATIO'N OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extin-guishing agent.in a timely manner. Prompt detection and suppression of fires L will reduce the potential for damage to safety-related equipment,and is an integral element in the overall facility fire protection program. Fire c'etectors that are used to actuate fire suppression systemt reprc ,cnt a more critically important component of a plant's fire protection program thaa detectors that are installed solely for early fire warning and notificcion. . Consequently, the m.inimum number of OPERABLE fire detectors must be gttater. The loss of detection capability for fire suppression ' systems, actuated = by fire detectors,. represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must h initi-ated at an earlier stage' than would be warranted for the loss of detect. ors that provide only early fire warning. The establishment of frequent fire patrols - in the affacted areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.7.10 LOOSE-PART DE'/ECTION SYSTEM The OPERABILITY of the loose part dctaction system ensures that suffictent capability is available to detect loose metallic parts in the primary systein and avoid or mitigate damage to primary system components. The allovacle out-
'of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.
3/4;3.7.11 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid ' effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / 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 instru-mentation is consistent with the requirements of General Design Criteria 60, - 63,- and 64 of Appendix A to 10 CFR Part 50. LIMERICX - UNIT 1 8 3/4 3-6 t"? -
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&Pt.anv6 3/4.3.7.12 4A"I0 ACTIVE GASE40G EEEDWiPR MONITORING INSTRUMENTATION The rh(ioactivAgaseous\ effluent\jnstrum'gntati'an is p'srovide to moni or an control, as appitcable, the releases of radioact(ve materials n gaseo s ef fibgnts durigg actual or potehtial releases o gaseous effsluents Thealarm/
tripsa$pointsforthestinstrumntsshallbeciculatedinkccord4ncewith the procedures On the ODCM to ens re that\ the al rm/tr,ip wi.11t occud prior 2
- e xW i ng -the -14 m Rs-o f--10 -Cf R--Pa rt--06. This instrumentation-the includes provisions for monitoring the concentrations of potentially explosive gas mix-tures in the.off gas 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.' ~ . 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect.the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.
3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to' initiate action of the feedwater system / main turbine trip system
, in the event of failure of feedwater controller under maximum demand.
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3892300520 , 3/4.11. 3 Lt0VIDRMATMREATNENTShTEH The quirement that the a3projriate portions of this systes be used when specified rovides assurance that t Ateleasts of radioactive matarials in liquid effluents w 1 be kept "as low as is reMonably achievable". This specification implements th requirements of 10 CFR 50. Gaoeral Design Criterion 60 of ' Appendix A to- CFA Part 50 and the desian pctivegiveninsection!!.0of Appendix ! to 10 R Part 50. The specified 1 ' governing.the use of appro- ' priate portions of a liquid redwaste treatment ta were specified as a ! suitable fraction of the dose design ob actives set fore in Section !!.A of Appendix I,10 CFR Part M, for liquid ffluents. i
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3/.4/11.1.4 LIOUID HOLOUP TANKS The tanks listed in this s ecification include all those outdoor radweste tanks that are not surrounded b liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding aree l drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material c'ontained in the specified
~
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.i L t nks 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 !!, Column 2 at the nearest potable water supply and the nearest s0rface water supply in,an UNRESTRICTED AREA. 3/4.11.2 GASE0US EFFLUENTS , 3/4 11.2.1 00$E RAT'E This specification is provided to ensu e that the dose at any time at and-beyond the SITE BOUNDARY from gaseous efflu a from the site will be within l E .the annual dose M t! M 10 CFR Part 20 to UNR(STRICTED AREAS, The annual dos 6 limits a the dose ass with the contantrations of 10 CFR Part 20, L ' Appendix B, M le !!, Column 1. limits proh de reasonable assurance that
. radioactive natorial discharged in gaseo ffluentk will not result in the L
exposure of i MEMBER OF'THE PUBLIC in an UNR ICTE AREA, either within or L TE BOUNDARY,.to annual average con ra ens-exceeding the limits i outside the L specified in pendix 8. Table II of 10 CFR Part 20 (1 CFR 20.106(b)(1)). For ! HEMBERS OF THE UBLIC who may at times be within the SITE BOUNDARY, the occupancy E of that MEMBER THE PUBLIC will usually be sufficienti low to compensate for any increase in t e atmospheric diffusion factor above t at for the SITE BOUNDARY. q Examples of calcul ions for such MEMBERS OF THE PUBLIC, with the appropriate ' occupancy factors all be given in the ODCH. The specified release rate limits restrict, at all time the corresponding gamma and beta dose rates above background to a HEMBER OF THE PUBL C at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the otal body or to less than or equat to 3000 mrams/ year to the skin. These releasA rate limits also restrict, at all times, the corresponding thyroid" dose rate above background-to a child via the inhalation s pathway to less then or equal %'o 1500 mesms/ year. This specification applies to the release of radioactive materials in gaseous effluents from all reacto'es at the site. N N 88 LIMERICK UNIT 1 9 3/4N11-E Juh 2 2 mf="= t *- ;
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s, RADIOACTIVE EFFLUENTS . 7 BASES DOSE RATE (Continued) . The required datEtion cap 3TTT r radioactive materials in gaseous waste samples are t bulated in terms of.the lower li ts of detection (LLDs). Detailed discussion f the LLD, and other detecthn li its can be found in. ' HASL Procedures Manua , HASL-300 (revised annually)Mrfe, L. A. , " Limits i for Qualitative Detect n and Quantitative Determjnation - Application to
- Radiochemistry," Anal. C
. 40, 586-93 (1968), and Hartwell, J. K., "Detec-tion Limits for Racioanalyt Hanford' Company Report SA-W ARH Kal S (June Counting 1975). Techniques," Atlantic Richfield 3 -
3/4dttt-00SG- -N08LE_ GASES '
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This specification is provided to implement the requirements of . Sections II.B. III.A, and IV.A of Appendix I,M O CFR Part 50. The Limiting -
. Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statementis ' provide the rkquired operating flexibility and at the sa.gedTiiie'Tmphaent the. guides set fort in Section'IV. A of Appendix I to assure that the releases k n dioactive materia (l (n gaseous effluents will be kept "as 1 w as is reasonably adWvable". The Sukveillance Requirements implement the equirements in Section I R A of Appendix with the guides of Appendix I be shown by talculationa\I that based l p ocedures conformance on 7
models and data ch that the actual exposure'o(a MEMBER (B{ THE PUBLIC through appropriate pathwa ( is unlikely to be substantialiysundere imated. The dose calculation established in the ODCM for calculating t N doses ue-to the actual release rates o radioactive noble gases in gaseou y ff ents are consistent.with the meth ology provided in Regulatory GuideT109, "Calcul'a- i
-tion ~of Annual Doses to Ma from Routine Releases of Reactor Effluents for.the Purpo'se of Evaluating Complia e with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Gui 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseo Effluents in Routine Releases from Light- .
Water Cooled Reactors," Revision 1, ly 1977 with site specific dispersiol curves.and deposition methodology. .The ODCM equations provided for deteraining the: air doses at and beyond the SITE BOUNDARY are based upon the historical , average atmospheric conditions. I 3/4.11.2.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This specification plement't h requirements of Sections II.C, III.A. IV.A of Append 10 CFR Part 50. The Limiting Conditions for Operati n are the guides set f inhection II.C of Appendix I. The ACTION statements rovide the required operat g ftaxibility and at the same time implement th guidessetforthinSectionIV4%fAppendixIto assure tnat the release of radioactive materials in gaseMs effluents will be kept "as low as is reaso bly achievable." The ODCM calculational methods specified in the Surveilla e Requirements implement the requirements in Section III.A of Appendix I at conformance with the guides of Appendix I be
- shown by calculational proced es based on models and data, such that the actual exposure of a MEMBER OF E PUBLIC through appropriate pathways is LIMERICK - UNIT 1 B 4 11-3 A .J S MS .,
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DOSE-10 DINE-131,10 DINE-133[TiRIU EANfMLADIONUCLIOES IN PARTICULATE FORM (Continued) i unlikely to be substantially nderestimated. The DCM calculational methods ; for. calculating the doses due o the actual release rates of the subject materials are consistent with t methodology provided in Regulatory Guide 1.109, i
" Calculation of' Annual Doses to from Routine Releases of. Reactor Effluents
- for the Purpose of Evaluating Compi nce with 10 CFR Part 50, Appendix I,"
Revision 1, October 1977 and Regulato Guide 1.111. " Methods for Estimating ' Atmospheric Transport and Dispersion of aseous Effluents in Routine Releasesi from Light-WaptC:: h d E actors," Revisio 1, July 1977 with site specific ! dispersion carrves and depos methodology. These equations also provide for determining /the actual-doses base n the hi orical average atmospheric ' conditions. The' release rate specific s for dine-131; iodine-133,- tritium,.an radionuclides in particulate
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ith half-lives greater than
- 8. days are d endent on the existing radionucli e pathways to man in areas at.,
and beyond the ITE BOUNDARY. The pathways which were examined in the develop-ment of these ca ulations were: (1) hdividual inhalation of airborne radio-nuclides, (2) depo tion of radionuclides onto green leafy vegetation with subsequent consumpti 'by man, (3) deposition onto grassy areas where milk animals and meat produ ng animals graze with consumption of the milk and meat-by man, and (4) depositio on the ground with subsequent exposure of man. 3/4.11.2.4 VENTILATIOiEXHAUSTTiiEITMDtT@ STEM The requireme that the appropriate portions of this' system be used, when
$pecified, provides r onable assurance that the releases of radioactive materials in gaseous eff ts will be kept "as low as is reasonably achievable."
This specification implements requiremen.ts of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 rt 50, and the design objectives given - in Section 11.0 of Appendix Lto 10 CF 50. The specified limits governing the use of appropriate portions 6 the syste re specified as a suitable fraction of the dose design objectives' set-fort ctions 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 poten-tially explosive gas mixtures contained in the main condenser offgas treat-ment system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will b.e controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. LIMERICK - UNIT 1 B 3/4 11-4 g up . x. g
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RADIOACTIVE EFFLUENTS BASES ' 3/4'.11.2.6 ' MAIN CONDENSER l Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable a.ssurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small. fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification *
. implements the requirements of General Design Criteria 60.and 64 of Appendix A to 10 CFR Part 50.
- i 3
l 3/4.11.2.7 VNM nD DilDnf NG ' i: This s'pec ation provide i ssurance that releases from l drywell purging op tions will not exceed (nnual dose limits of 10 CFR
- l. Part 20 for UNRESTRIC 'A8EAS.
3/4.11.3 0 RADWAS NT '
- j. The OPERABIL of the solid yttem ensures that the system will be available for use w ever solid radwastes r'equ Qe processing and packaging prior to being shipped of e, This specification implements the requirements ,
of 10 CFR 50.36a and General n Criterion 60 of Appendix A to 10 CFR.Part 50. The R meters included in tablishing the PROCESS CONTROL PROGRAM may include, but are ' ted to wa type, waste pH, waste / liquid /solidifi-i cation agent / catalyst ratios, was c tent, waste principal chemical constituents, mixing and curing times. . 3/4.11.4 TSTAL DOSE This spec cation is provided to meet the dose limitations of 40 CFR Part 190 that have been incorp ted into 10 CFR Part 20 by 46 CFR 18525. The specification .
-requires the preparation submittal of a Special Report whenever the calcu- '
lated doses from plant radio ive effluents exceed twice the design objective-doses of Appendix I. For sites ntaining up to four reactors, it is highly unlikely that the resultant dose to MEMBER OF THE PUBLIC will exceed the
.doselimitsof40CFRPart190ifthe%dividualreactorsremainwithinthe reporting requirement level. The.SpeciaNeportwilldescribeacourseof '
action that should result in the 'limitationNof the annual dose to a member of ? the public to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from,other uranium fuel cycle sources is negligib} , with the exception that - doce dontributions from et ae M a e fuel cycle f tilities at the same site or within a radius of 5 iles must d'iio7ntidered. I he dose to any MEMBER ,n ' OF THE PUBLIC is estim ted to exceed the requireme tts f 40 CFR Part 190, the Special Report with a equest for a variance (provid'ed the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in
.accordancti with the pro siens of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request an fulfills the requirements of 40 CFR Part 190 until
-NRC staff action.is complet . An individual is not considered a MEMBER OF THE PUBLIC during any period in w 'ch he/she is engaged in carrying out any operation that is part of the nuclear fuei cle.
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4 3/4.12 RADIOLOGICA,L ENVIRONMENTAL MONITORING BASES 3/4.12.'1 MONITORING PROGRAM x The radioTogical environmental monitoring program required by this specifica-
. tion provides repr ntative measurements of radiation and of radioactive materials i,n those exp ure pathways and for those radionuclides that lead to the highest potential ra tion exposures of MEMBERS OF THE PUBLIC .resulting Trom the station operation, his monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part and thereby supplements the radiological effluent monitoring program by verifying hat the measurable concentrations of radioactive materials and levelt of radia a e not higher than expected on the basis of the effluent measurements and m n of the environmental exposure pathways.
The initially'specified monitoring am will be effective for at least the first'3 years of commercial operation. Following this period, program changes may be initiated based on operational expe fence. The required dete capabilities for env Momental sample analyses are tabulated in terms of the lowe its of detection (1.'tDs The LLDs required by Table 4.12.1-1 ar considered op for routine enviro ental measurements in industrial laborat ries. It should be cognized that th LLD is defined as an a priori (t,efore Me fact) limit repres ing the capab 11ty of a measure-ment system and not % an a posteriori (after the t) limit or a particular
- measurement
( Detailed dis ion of the LLD, and ot'e Nd etection limits, can be found I. in NASL Procedures Man -HASL-300(revisedanneaQy); Currie,'L.A.," Limits for Qualitative Detection a uantitative Determinati n - Application to . . .- Radiochemistry" Anal.' Chem. 40, (1968); and Hartw J. K., " Detection Limits for Radioanalytical Counting Te ues," Atlantic Ri ield Hanford Company Report ARH-SA-215 (June 1975). s . 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond'the SITE' BOUNDARY are Adentified and that modifications to the monitoring prograq are made if requ, ired by the results of this census. The , best information f%m the door-to-door survey, 'aarial survey or consulting < l with local agricultural authorities or any combination of these methods shall l be used. Thiscensussa(isfiestherequirementsofSectionIV.B.3ofAppendixI l to 10 CFR Part 50. Restricting the census to gardens of' greater.than 500 l squarefeetprovidesassuraKethatsignificantexposurepathwaysvialeafy vegetables will be identified end monitored since a garden of .this size is the minimum required to produce the\ quantity (26 kg/ year) of leafy' vegetables i assumed in Regulatory Guide 1.109Nfor consumption by a child. To'oetermine this minimum garden size, the follo' wing 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 yieldsof 2 kg/ square meter. LIMERICK - UNIT 1 B 3/4 12-1 IE5 '9 - - g
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l RADIOLOGICAL ENVIRONMENTAL MONITORING - BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM Th req 'e ment'for participation i laboratory C Phon Program l is provided to e re that independent c ecks on t precision and accuracy of . the meas 6rements of (dioactive material l in environmeRal sample matric'es are performed \as part of th
.in order to demonstratetha %a'lity assuran are he.results e program reasonably for1 % ironmental vall monitoring or the purpose of Section\IV.B.2 of Appendix 10 C Part 50.
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, lg LIMERICK - UNIT 1 B 3/4 12-2 t
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eneric Letter. 89- 01 - 2.- Enclosure-3 I t- ..
.- 6.0 " ADMINISTRATIVE' CONTROLS 1 1 3
3 56/ [ PROCEDURES AND PROGRAMS ( Q
, = 6d!r 4JedoRtwinggo,grShatNutAhabtRKey/Tihqlp e Radioactive Effluent Controls Program z A program shall.be provided conforiqing 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, -f ' (21 shall be implemented by operating procedures, and (3) shall in-- ~ clude remedial actions to be taken whenever the program limits are t exceeded.~ The program shall include the following elements: ; 1)- Limitations on the operability of radioactive liquid and gaseo n mon'itoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the , ODCM,- 4 l
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, .
L 3) Monitoring, sampling, and analysis of radioactive liquid and l _ gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, _
- 4) Limitations on the annual and quarterly doses or. dose commitment i to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS confera-ing-to Appendix I to 10 CFR Part 50,
- 5) Determination of cumulative and projected dose contribu- h .p..
from radioactive effluents for the current calendar quai current calendar year in ac:ordance with the methodolog) 90 parameters in the 00CM at least every 31 days, [, 6 .
'l Limitations on the operability and use of the liquid anc t 6) effluent treatment systems to ensure _that the appropriat [/6) k
/
i portions of these systems are t, sed to raduce releases of activity when the projected doses in a 31-day period wot exceed 2 percent of the guidelines for the annual dose ( commitment conforming to Appendix I to 10 CFR Part 50,
- 7) Limitations on the dose rate resulting from radioactive material r released in gaseous effluents to areas beyond the SITE BOUNDARY L
conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, l} I (! l
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- [j ADMINISTRATIVE CONTROLS
,y , 6.8.4 , _ LRadioactive Effluent Controls Program (Cont.-) ,
- 8) Limitations on the annual and quarterly air doses resulting from noble gases released in. gaseous effluents from'each unit-to- ,
areas beyond the SITE 80VHDARY conforming:to Appendix I to-10 CFR Part 50, 9). Limitations on the annual and quart'erly 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 un.it to areas beyond the-SITE BOUNDARY conforming to Appendix I to 10 CFR~Part 50,-
- 10) Limitations on venting and purging of the Mark II= containment through the-Standby Gas Treatment System.to maintain releases.
as low as reasonably achievable w .. . . ,....m .. - . . . _ - and
- 11) Limitations on the annual dose or dose. commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
p R_adiological Environmental Monitoring Program i 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) verific'ation of the accuracy of , the effluent monitoring program and'modeling of environmental expo-sure pathways. The program shall'(1) be contained in the 00CM,
.(2) conform to the guidance of Appendix I to 10 CFR Part 50, and 4 (3) include the'following:
- 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the envirer.ar,t in accordance with the metho6 ology and parameters in the 00CM,
.A_ Land Use Census to ensure that changes in the use of areas at 2) and'beyond the SITE B0UNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and
- 3) ParticipationiiiafInterlaboratoryComparisonProgramtoensura that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sar '- g.
matrices are performed as part of the quality assurance pr p gram for environmental monitoring,
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r o . 3892300520 3/4.11 'RADI0 ACTIVE EFFLUENTS , BASES . 3/4.11.1-L10VIDEFFLUENTS
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3/4.11.1.1- CONCENTRATION This specificahqn is provided to ensure that the concentration of radio- ! active materials releaktd in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentPation levels specified in 10 CFR Part 20. Appendix B,
._ Table II, Column 2. This Heitation provides additional assurance that the levels of radioactive materials in b'odies of water in UNRESTRICTED AREAS will result in
- exposures withirrtibtbe Sectio'rt4I. A design objectives of Appendix I,10'CFR Part 50, to a M MBER OF'7HE PUBLI population. The concentrattoglimi and (2) the limits of 10 CFR 20.106(e) to the for dissolved or entrained noble gases are based upon th Xe 25 is the controlling radioisotope and L .its MPC in air was assumption enverted to antha(equiv'tlent concentration in water using thl L methods described in the Internationa b omhi' sion on Radiological Protection
[ (ICRP) Publication 2. . The required detect'on capabilities for rad active materials in liquid waste samples are tabulated inN terms of the' lower limitt detection (LL0s). Detailed discussion of the. LLD, and hher detection limits can be found in HASL Procedures Manual, HASL-300 (revised annhally), Currie, L. A. , " Limits for Qualitative Detec-tion and Quantitative Determinktion - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968),- and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting ,- /- Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.1.2 DOSE This specification is provided to impleme t the requirements of Sections II.A,
'III.A, and IV.A of '
10 CFR Part 50. he Limiting Condition for Opera-tion implements ) e guides set h in Section I A of Appendix I. The ACTION statements provf ng flexibili and at the same time imple-ment the guidesl,de the required set- forth oper in Section IV, f Appendix to assure that the releases , of radioactive laterial in liquid effluents 11 be kep "as low as is reasonably achievable." A o, for fresh water sites with inking ter' supplies which can be potentially affe ted by plant operations, there is (easona e assurance.that the operation of the- acility will not result in radionucy de co centrations in the finished drinking hater that are in excess of the requi'rement of 40 CFR Part 141. Thedosecalculatiohmethodologyandparametersinthe00CMimplementtherequire- s . ments in Section III. of Appendix I that conformance with the, guides of Appendix I ,. be shown by calculati al procedures based on models and data,'su that the actual l' exposure of a MEMBER 0 THE PUBLIC through appropriate pathways f niikely to be
-.substantially underestim ted. The equations specified in the 00CM for calculating the doses due to the actu release rates of radioactive materials in liquid efflu-ents are consistent with th methodology provided in Regulatory Guide 1.109,
,' " Calculation of Annual Doses o Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Comp iance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Gui 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reac r Releases for the Purpose of Implementing
-Appendix I," April'1977.
This specification applies to t4 release of radioactive materials in liquid ef fluents from the site. LIMERICK - UNIT 1 8 3/4 11-1 Amendment No. 26 M 2 2 O
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.f-; 5.0' DESIGN FEATURES
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5.1 SITE-EXCLUSION AREA ' 5.1.1:.The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE - 5.1. 2 The low population zone shall be as shown in Fig'ure~ 5'.1.2-1. ' MPS DEFINING UNRESTRICTED AREAS AND' SITE BOUNDAhFOR RADI0 ACTIVE LI0tr10 EFFLUENTS
- 5.1. 3 In ation regarding radi nctive gaseous and liqu effluents, which F
will. allow i kntification of structu'ess and release points a(well as defini- - tion of UNRESTRIC D AREAS within the SIT BOUNDARY that are ac<;essible to t MEMBER OF THE PUBLI , hall.be as shown in ures'5.1.3-la and .3-lb. ! 1 NETEOROLOGICAL ' TOWEPt.0 CAT 16M O s ! 5.1.4 twneteorological t 11 be located as n on Figure 5.1 M-1. 5.2 CONTAINMENT- ' t CONFIGURATION . ~ 5.2.1 The primary containment is a steel lined reinforced concrete structure > consisting of a.drywell and suppression chamber. The drywell is a steel-lined reinforced concrete vessel in the shape of a. truncated cone on top of a water- , filled suppression chamber'and is separated by a diaphragm slab and connected
.to the suppression chamber through a series of downcomer vents. The drywell has a maximum free air volume of 243,580 cubic feet at a minimum suppression pool level of 22 feet. 'The suppression chamber has a maximum. air region of
- 159,540 cubic feet and a minimum water region of 122,120 cubic feet. -
DESIGN TEMPERATURE AND PRESSURE 5.2.2. The primary containment is designed and shall be maintained for: a.- Maximum internal pressure 55 psig. L b. Naximum internal temperature: drywell 340'F.
, suppression pool 220'F.
- c. Maximum external pressure 5 psig,
- d. Maximum floor dif.ferential pressure: 30 psid, downward. 1 20 psid, upward.
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'FOR RADI0 ACTIVE G EOUS AND LIQUID E FLUENTS l LIMERICK - UNIT 1 5-4
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. 3891062$60-ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
'6.8.4. The following programs shall be established,, implemented, and maintainW:
- a. Primarv Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious trensient or accident to as low as practical levels. The systems include.the core spray, high pressure coolant injection, reactor core isolation cooling, residual heat removal, post-accident sampling system, safeguard piping fill system, control rod drive scra.i discharge' system, and containment air' monitor systems.. The program shall include the following: .
- 1. Preventive maintenance and periodic visual inspection requirements, and-
- 2. Integrated leak test requirements' Tor each system at refueling cycle intervals or less. , .
- b. In-Plant Radiation Monitorino A program which will ensure the capabi.lity to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
- 1. Trainingofpersonnel, ,
- 2. Procedures for monitoring, and
- 3. Provis' ions for maintenance of sampling-and analysis equipment.
- c. Post-accident Samolinc*
a A program which will ensure the capability to obtain and analyze resctor
- coolant, radioactive iodines and particulates in plant gaseous afflu ents, and containment atmosphere samples under accident conditions.
The program shall include the following:
- 1. Training of personnel,
- 2. Procedures for sampling and analysis, and
- 3. Provisions for maintenance of sampling and analysis equipment.
' "Not required until prior to exceeding 5.4 of RATED THERMAL POWER.
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4 L L, 3893224f50 ADMINISTRATIVE CONTROLS 6.9 REPORTING RE0VIREMENTS-ROUTINE RE00RTS ' 6.9.1 In addition to the applicable recorting requirements of Title 10, Coce
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e of Federal Regulations, tne following eports snail be submitted to the Regional Administrator of the Regional Office of tne NRC unless otnerwise notec. STARTUP REPORT
- 6. 9.1.1 A summary report of plant startup and power escalaticn testing shall be suomitted following (1) receipt of an Operating License, (2) amencment to the license involving a planned increase in power level, (3) installation of fuel that has a dif ferent design or has been manufactured by a dif ferent fuel supplier, and (4) modifications that may have significantly altered the nuclear, tnermal, or hydraulic performance of the unit.
6.9.1.2 The startup report shall address each of the tests identified in sub-section 14.2.12 of the Final Safety Analysis Report and shall include a descrip- - tion of the measured values of the operating concitions or characteristics . obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions tnat were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
- 6.9.1.3 .Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following initial criticality, whichever-is earliest. If the startup report does not cover all three events (i.e. , initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.
ANNUAL REPORTS
- 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted-prior to March 1 of each year.
.The initial report shall be submitted prior to March 1 of the year following initial criticality.
- 6. 9.1. 5 Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility,-
and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated-man-rem exposure according to work and job functions"" (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket "A single submittal may be made for a multiple unit station.
"This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.
. \
LIMERICX - UNIT 1 6-15 g gtg
ADMINf5TRATIVE CONTROLS
~~
ANNUAL REPORTS (Continued)- J !
' dosimeter, thermoluminescent dosimeter (TLD), or film badge measure- .!
ments. Small exposures. totalling less than 20% of the individual ! total oose need not be accounted for. In the aggregate,'at least 1 (p 80% of the total wnole-body dose received from external sources:should be assigned to specific major work functions; b 1 ,.
- b. Documentation of all challenges to safety / relief valves; and
-l ~ c. An'y other unit unique reports raquired on an annual basis.
U d. The results of specific activity analysis in which the primary . l coolant exceeded the limits of Specificati m 3.4.5. The following ( '"< information shall be included: (1) Reactor power history starting ! 48 hours prior to the first sample in which the limit was exceeded; .{ )f#' (2) Results of the last isotopic analysis for radiciodine performed i f ! prior to exceeding the limit, results of analysis while limit was. hlR l !4 exceeded and results of one analysis after the radiciodine activity ; was reduced to less than~ limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup , 1 41 system flow history starting 48 hours prior to the first sample in i which the limit was exceeded; (4) Graph of the I-131 concentration , , i and one other radiciodine isotope concentration 'in micrucuries per i cram as a function of time for the duration of the specific activity R above the steady-state level; and (5) The time duration when the t specific activity of the primary coolant exceeded the radiciodine 7 limit. .i k MONTHLY OPERATING REPORTS b 6.9.1.6 Routine reports of ' operating statistics and shutdown experience, includ-ing documentation of all challenges to the the main steam system safety / relief : J- . valves, shall be submitted on'a monthly basis to the Director? Office of Resource !
# Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a C 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.
'i ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1'.M)TheAnnualRadiologicalEnvironmentalOperatingReportcoveringthe 5 operation of the unit during the previous calendar year shall be submitted ,
before May 1 of each year. The report shall include summaries, interpreta-l
--tions, and analysis of trends of the results of the Radiological Environmental i
Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and /?) Sections <
,. IV.B.2, IV.B.3, and IV.C of Appendix 'I to 10 CFR Part 50.
*A single submittai may be made for a multiple unit station.
( - i f b lCK UNIT 1 6-16 Amendmen t No. 20 MM i 9 O
/--
s-. 1
4 ADMINf5TRATIVE CONTROLS .
. ,. m ANNUAL REPORTS (Continued) N .
environment surv llance activities for'the report period, nelucing a com-par'ison (as app s riate), with preoperationat studies,'operatiohal controls-and previous enviro ental surveillance report'ssand an assessment'of the ob-j served impacts of the nt operation on the environment. The-reports shall also include the results o land use censuses requi' red by Specification. 3.12.2.
\
The ual. Radiological Environm 'al Operating Reports 'shall include the'results i of all r 'ological environmental sa me'asurements ken during the report pclesoc and of all environmental radiation'N pursuant to the locations specified in the tables a figures in the OFFSITE O F CALCULATION MANUAL'has well as , summarized and tabu ted results of these ana es and measurements'%n the format of the-table in he Radiological-Branch Tec ical Position, Revi4ion 1, ovember 1979. .In the eve i usion with the report, t thatport some individual r ults are not available for shall be submitte'd oting and explaining
-the ?easons for the missing resu ts. The missing data sh 11 be submitted as 1 soon as pos ible in a supplementary N ort. .
The reports sha o include the follow 1 - a summary descriptioNf the radiological enviro ental monitoring progra asttwolegiblema)R*
'l l
3
)
q
. r
. l t
s
^^0n'e' map sii i caver stations near the SITE BOUNDAR % e p il include ,
e the m6[s/di n t 's ta t i ons . - L.. V w 8/ LIMERICK - UNIT 1 6-16a Amendmen t No. 20 MAY i 91983
" z 3892300520-t
.f 1
* .' I ADMINISTRATIVE CONTROLS
[" ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 1 h ng all samplinblocations keyed h table giv)hg distances Artd directions-from he centerline of th ' in the Interlaboratorym Co(parjson Program, r'equired by Specification 3.'12 discussion of'all deviations front the Sampling'5chedule of Table 4.12.1-lhand discussion of alisanalyses in whihNthe LLO require (by Table A12.1-1 was not 1 achievable. N } SEMIANNUALRADIOACTIVEEFFLUENTRELEASEREPORT* 6.9.1. W 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-1 clude a summary of the quantities of radioactive liquid and gaseous effluents i and solid waste released-from the unit. The material provided'shall be 'l
.(1) consistent with the objectives outlined in the ODCM and PCP and (2) in con- -
formance with 10 CFR 50.36a and Section IV.B.1 of. Appendix I to 10 CFR Part 50. j . e, a roo e n,
.............,....m. .m ., ., ..
no..y,=. v, novi. a c t i v e M s.~,.. a'y ,
,~ . . . ..,terials in ; quid and Gaseous Effluents from Light-Water-Cooled s
Nucita L Power Plan ," Revision 1, Jun'e 1974, with data' summarized on a q0arterly basis following th format of Appendi s B thereof.
"The Seniiannual Radioactive Ef fluen Release Report t'a be submitted 6 days after '
N-V Jsnuary l'of ea h year shall includ an annual summar ofhourlymetebrological datascollected o er the previous yea . This annual su mary may be either in i
~
the thrm of'en h ur-cy-hour listing n magnetic tape o wind speed, wi nd dire - 1 tion ahd atm sphe ic stability, and- recipitation (if m asured), or ir, the fo . 1 of joint freq enc' distributions of ind speed, wind di,ection, atmossheric stabil ty,** This same report shall nclude an assessment of the radiation f doses ue to the r icactive liquid a d gaseous effluents released from the unit stat on duri g the previous ca endar- year. This same report shall also i incio an assessmen of the radiation oses from radicaitive' liquid and gase- i l ous effluents S OF THE PUBLIC ue to their acti ities inside the SITE- \ -j
- i. BOUNDARYares (Fig 5.1.3
\ to MEMSla and 5.1.3-lb) during the repor period. Al assump-tions and 'lo$ sed in making the assessments (i.e. , specific act'vity, exposu e time 1
i I c4 tion) sh'all be in dedintheserhorts. The asse sment of ra 'ation doses shall be pehformed in accordance with'the methodology nd parameter of . theOFFSITE'00SECALCULATIONMA(UAL'(00CM). N 3 l N N
$*A single submittal may be made for a multi unit station. The submittal should combine those sections that are common to all units at the station; l
however, for units with separate radwaste systems, the submittal shall specify I the releases of radioactive material from each unit. required'#Leteorological' data on } site in a f Qej that shall'be_provided to the NRC upon request. LIMERICK - UNIT 1 6-17 Amendment No. 26 JUII E lb _---------------------------------Y
3892300520 4 - ADMINISTRATIVE CONTROLS- . SEMIANNUAL RA'Oh0 ACTIVE EFFLUEift RELEASE REPORT '(C$ntinued) s ~The S isn'nual Ra tactive Effluenhltelease Report be N January hof each ye'h( shall also indlude an assessmeh c, submitted f radiation d ses60todays after . -
\the likely\eost expose 4 MEMBER.0F THE fyBLIC from react'cr releases and her -
nearby uranium fuel cych sources (incloding doses from risary effluen path-
' hys and direct radiatioh) for the providus calendar yea. to show confo nce with 40 CFR Pa 190, Env naental Radia on Protection standards for Nu lear Power Operatio Acceptab methods for c iculating the se contribution from, ligdid and gase' us effluent are given in R ulatory Guicie 1.109, Rev.1
-October 19'n. . .
The S iannual Radion:tive Ef luent Release oports shalt. i clude the foil np inforniation for nach type of lid waste (as befined in
- D R Part 61) shipped
. offsitle during the report pari d: -
\
. Container volume. -
_j
. Total curie quantity {specify whethe deterv.inued b) measurement c -
esti te, - - {
- c. Prinqipal radionuclidi s (specify whether detersrinet by measuremert or es timate),
d.. ' Source of' waste and processing employ d (e.g. , dewatered spent ri sin, compacted dry waste, aporator bott s),
- e. Type of container (e.g , LSA, Type A, Type B, Large Quantity), a.
- f. SOLI IFICATION agent o absorbent (e.g , cement ; ur a formaldehyc ).
. -The $ miannual adioactive Efflue t Release Repo s shal! inci de a list an descr tior) of planned releases rom the site UNRES7.ICTE AREAS,of rad o ' .
active terials in gaseous and liq id. effluents - de dur-ing t reporting period.' The Semia nual Radio etive Effluent Re ease Reports s li include any chan;es made durin the repor 'ng period to the ROCESS CONTROL F5ROGRAM ( P) and to the ODCM, a well as a ting of new loc tions for dose =alculations and/or environmenta monitoring entified by the\ land use censes pursuant t.c Specification .12.2.
\
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regiona7 Administrator of the ' Regional Office of the NRC within the time period specifiund for each report. LIMERICK - UNIT 1 6-18
M S W6 ,
,,y ega , eWW 6 64 4'4989 4-
y,_7 _ . -s. s . ADIONISTRATIVE CONTROLS ._ 6.10 RECORD RETENTION - 6.10.1. In addition to the applicable rc ord. retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum ~ period indicated. 6i.10.2 The following_ records shall be retained for at.least 5 years:
- .a. Records and logs of unit operation covering time intervalJ at each power level.
- b. Records and logs of principal maintenance activities,_ inspections, repair, and replacement of principal items of equipment related to >
nuclear safety.
- c. All REPORTABLE-EVENTS,
- d. Records of surveillance activities, inspections, and calibr~ations-required by.these Technical Specifications.
. e .- Records of changes made to the procedures required by Specification
'6.8.1. ,
- f. Records of rad'ioactive shipments.
-g. Records of sealed source and fission detector leak tests and results.
- h. Records of annual phy'sical inventory of all sealed source material ofJrecord.
j
- 6.10I3 The'following records shall be retained for the duration of the unit Operating License: i
- a. Records and drawing changes reflecting unit' design modifications made to_ systems and equipment described in the Final Safety Analysis Report. t
- b. Records of new and irradiated fuel inventory, fuel . transfers,' and assembly burnup histories. . i L 'c. Records of radiation exposure for all individuals entering radiatloa control areas, L
l~ I l'
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i j' t: 1
- LIMERICK - UNIT 1 6 p3 ][~igJ ~ - _._
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ADMINISTRATIVE CONTROLS
-RECORO RETENTION (' Continued) 4 ,
- d. Records of gaseous and liquid radioactive material released to the 4 environs. _
e. Records of transient or operational cycles for those unit components identified in Table 5.6.1-1. f.- "F.ecords of-reactor tests and experiments.
- g. t Records of training and qualification. for curr.ent members of the unit
. .s taf f. !
d
' I
- h. Records of inservice inspections performed pursuant to these Technical
; Specifications. .
- i. Records of quality assurance activities required.by the' Operational Quality Assurance Manual not listed in Section.6.10.2. ]
'j. Records of reviews. performed for changes made to procedures or equip ^
i ment or reviews of tests and experiments pursuant to 10 CFR 50.59. d
- k. ' Records of meetings of the PORC and the NRB. -]
, j
- 1. Records of the service lives of all snubbers including the date a't which the service life commences and associated installation and 1 maintenance records. i.
- m. -
Recorcs of- analysis required by the Radiological Environmental - Monitoring Program that would permit evaluation of the accuracy of the analysis at a later data, m
/7p. Records of. reviews performed for changes made to the OFFSITE DOSE 7
CALCULATION . MANUAL ar ~ -'.1 PROCESS CONTROL PROGRAM. 6.11 RADIATION 9ROTECTION PROGRAM ' 6.11.1 Procedures for personnel radiation protection shall be prepared can-sistent with the requirements of 10 CFR Part 20-and.shall be approved, main-tained, 'and adhered to .for all operations involving personnel radiation exposure. -
- 6.1_2 HIGH RADIATION AREA '
1
-6.12.1 . In lieu of the " control oevice" or " alarm signal" required by paragrapn L
' 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which tne intensity of radiation is-greater than 100 mrem /h but less than 1000 mrem /h shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radicticn Verk Permit (RWP)*. Any individual or group of individuals permitted to enter such are %
shall be provided with or accompanied by one or more of the following: i
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
V ,I i
' Health physics personnei or personnel escorted by health physics personnel ~
l shall ce exempt from the RWP issuance requirement curing the perfarma.6ce or
.their assigned raciatica protection duties, provided they are otherwise following plant radiation protection procedures for antry into hign radicion areas. 3 t
' LIMERICK - UNIT 1 6-20 G >; g ,
p es . .*
$,Ni$.Y$10. , SS.., .....k . n1 2. C i h U h t. L $ 5 % s h . .'.!.i S w . . w'd, . . a -.':0.. % . U..d.Y<.:E . . W ' [' *
- E
. . . . . , . _ ~ ._ . _ 1. *
. \
ADMINIWRATIVE CONTROLS -. 1 HIGH RADIATION. AREA (Continued) !
- b. A radiation monitoring device which' continuously integrates the radiation dose rate in the area and alams when a preset integrated '
dose is received. Entry into such . areas with this monitoring device may be made after the dose rate lev,els in the area have been estab-11shed and personnel have been made. knowledgeable of them.
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing pdsitive control over the activi-ties within the area and shall perform periodic radiatien surveil-
.. lance at the frequency specified by the Health Physicist in the RWP. 6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible *
, to personnel with radiation levels such that,a major portion of the body could receive in I hour a dose greater than 1000 mrems shall be provided with locked doors to prevent unauthoHzed entry, and the keys shall b'e maintained under the administrative control of Shift Supervision on duty ahd/or the health '
physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the inmediate 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 boQ could receive in 1 hour a dose in excess of 1000 mrems" that are located withia large areas, such as the 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 RWP, continuous surveillance direct or remote (such as use of closed circuit TV cameras), may ,
be made by personnel qualified in radiation protection procedures to provide . positive expcsure control over the activities within the area. 6.13' PROCESS CONTROL PROGRAM (pCP) Changes to the PCP: .
- a. Shall be documented and records of reviews performed sha;l be retain- -
ed as required by Specification 6.10.30. This document w on shall &g contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and s &
" Measurement made at 18 inches from sour:e of radioactivity.
- LIMERICK.- UNIT 1 6-21 I- A > - -- J 4 m .Of5 4D C26
@ J .canN m NNNru-L & 2 ^m aA m h h #*x".=.m m = 'b
. h / '
'}
On ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (Continued)
- 2. A determination that the change djd not reduce the overall conformance of the solidified waste product to existing erh:rk a p i18414. Sr ::1M r;t:;; 5, It. M4s.bs cr.d rep ;rueJ, A Fede<<.) SV4 - e 61 e , '
3 :' ' cumentation he facts haththangelas beoorthewed~ and found acc3p able% y th ORC. '
- b. Shall become effegtive upon review'and acceptance by the PORC,r - L M agn ud 4 %eTh+Mce e.
. _6.24 0FFSITE DOSE CALCULATION MANUAL (OD )
r5anges to the 00CM; ' a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.3#. This decementation shall contain: 0
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- 2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
l b. Shall become effective after review and acceptance by the and the approval of the Plant Manager. c. Shall be submitted to the' Commission in the form of a complete, . , l legible copy of the entire 00CM as a part of or concurrent with the L 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 - incicate the date (e.g., month / year) the change was implemented. l w. w. A L awsyass en e wie beu ruaJus wuouwsm was a sw e veww e vs waa w ey e wsma l (liquid, gaseb(s and solid): \wu
- a. Shall be y ported to the Commission the Semiannual Radioactive Effluent R sase Report for the pcrio n which the change was\made effective. ThK discussion of ear.h chang shall contain:
- 1. A summary of h evaluation that led to the change could e determination thatx -
made in accordance wit 10 CFR 50.59;
\
s LIMERICK - UNIT 1 6-22 A'.' 8W i
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'[ . ;
. ss KAJOR CHANGES TO RADIOACTIVE WASTE TREATNENT SYSTEMS (Continued 1 N
- 2. Sufficient detailed inform \ i i
f6r.thechangewithoutbenef{1at on of to totally support additional or suppthe reason emental information; l 1
- 3. A detailed description of the eq pment, components, and processes ;
involved and the interfaces with o her plant.systents; ^
- 4. An evalua fon of the change which sh ws the predicted te pases
- s i of radioactive ma+,erials in liquid an gaseous effluents And/or !
quantity of solid waste that differ fr a those previously re- i dicted in the license application and endments thereto;
- 5. An evaluation of\the change which shows t v expected maximum exposures to indikidual in the UNRESTRICT [
AREA and to the ' general population \that differ from those eviously estimate in the license application and amendments t ereto; .
- 6. A comparison of ther,edicted p\ releases of ra osctive material ,
in liquid and gaseous affluents and in solid waste, to the actusi releases for the period \riortowhenthechanesaretobemadj; ;
- 7. An estimate of the exposure to plant opcilting rsonnel as a !
result of the change; and - k 8. Documentation of the fact th't the change was rev .wed and found acceptable by the PORC. I bi Shall be reviewed and accepted by th PORC prir. to implapentation,
, j i
1 LIMERICK - UNIT 1 6-23 , -s,.s LDb blili-
s,5 t i 8 k r i e t I ATTACHMENT 3 I L 0 I I' l . i
,h
-)
l b Y l l l l i l 1 b
DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.22 MEMBER (S) 0F THE PUBLIC.......................................... 1-4 1.23 MINIMUM CRITICAL POWER RATI0..................................... 1-4 ; 1.24 0FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 j 1.25 OPERABLE - OPERABILITY........................................... 1-4 1.26 OPERATIONAL CONDITION - CONDIT10N................................ 1-4 1.27 PHYSICS TESTS.................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE........................................ 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY.................................... 1-5 1.30 PROCESS CONTROL PR0 GRAM.......................................... 1-5 1.31 PURGE - PURGING.................................................. 1-5 1,32 RATED THERMAL P0WER.............................................. 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY................ 1-6 l-1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME.......................... 1-6 1.35 REFUELING FLOOR SECONDARY. CONTAINMENT INTEGRITY...................!-6 l l i 1.36 REPORTABLE EVENT................................................. 1-7 l 1.37 ROD 0ENSITY...................................................... 1-7 l l 1.38 SHUTDOWN MARGIN.................................................. 1-7 l 1.39 SITE B0VNDARY.................................................... 1-7 , 1.40 (0eleted)........................................................ 1-7 1,41 SOURCE CHECK..................................................... 1-7 1.42 STAGGERED TEST BASIS............................................. 1-8 1.43 THERMAL P0WER.................................................... 1-8 1.44 UNIDENTIFIED LEAKAGE............................................. 1-8 l LIMERICK - UNIT 2 ii
i
-INDEX 3
LIMITING CON 0!T!0N5 FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements ........................ 3/4 3-66 Seismic Monitoring Instrumentation ........................ 3/4 3-68 Table 3.3.7.2-1 Seismic Monitoring Instrumentation ..................... 3/4 3-69 Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirements ........................ 3/4 3-71 The information from pages 3/4 3-73 through 3/4 3-75 has been intentionally omitted. Refer to note on page 3/4 3-73 .......... 3/4 3-73 Remote Shutdown System Instrumentation and Controls ....... 3/4 3-76 Table 3.3.7.4-1 Remote Shutdowr System Instrumentation and Controls......... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance Requirements ........................ 3/4 3-83 Accident Monitoring Instrumentation ....................... 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-i tation .............................. 3/4 3-85 I Table 4.3.7.5-1 Accident Monitoring Instrumentation L Surveillance Requirements ........... 3/4 3-87 Source Range Monitors ..................................... 3/4 3-88 l Traversing in-Core Probe System ........................... 3/4 3-89 l Chlorine Detection System ................................. 3/4 3-90 Toxic Gas Detection System ................................ 3/4 3-91
~
Fire Detection Instrumentation ............................ 3/4 3-92 LIMERICK - UNIT 2 ix
INDEX 'l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)- ! Table 3.3.7.9-1 Fire Detection Instrumentation ...... 3/4 3-93 Loose-Part Detection System ............................... 3/4 3-97 The information from pages 3/4 3-98 through 5 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98....................... 3/4 3-98 , Offgas Monitoring Instrumentation ......................... 3/4 3-103 Table 3.3.7.12-1 Offgas , Monitoring Instrumentation ......... 3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements .......... 3/4 3-107 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM .................. 3/4 3-110 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTAITON ...................................... 3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation ...... 3/4 3-113 ' Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumentation Setpoints ............................. 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip L System Actuation Instrumentation Surveillance Requirements ........... 3/4 3-115 . 3/4.4 REACTOR COOLANT SYSTEM i l 3/4.4.1 RECIRCULATION SYSTEM Recirculation loops ................................... 3/4 4-1 LIMERICK - UNIT 2 x
t-INDEX LIMITING CCWDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REFUELING OPERATING (Continued) 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Remova1.................................. 3/4 9-13 Multiple Control Rod Remova1................................ 3/4 9-15 t 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1............................................ 3/4 9-17 Low Water Leve1............................................. 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY............................... 3/4 10-1 3/4.10.2 R0D WORTH MINIM 12ER.......................................... 3/4 10-2 3/4.10.3 SHUTDOWN MARCIN DEMONSTRATIONS............................. 3/4 10-3 3/4.10.4 RECIRCULATION L00PS......................................... 3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION........................................ 3/4 10-5 l 3/4.10.6 TRAINING STARTUPS........................................... 3/4 10-6 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.............. 3/4 10-7 3/4.11 RADI0 ACTIVE EFFLUENTS , 3/4.11.1 LIQUID EFFLUENTS
- l The information from pages 3/4 11-1 through 3/4 11-6 has been intentionally omitted. Refer to note on page 3/4 11-1..................... 3/4 11-1
' Liquid Holdup Tanks.......................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS The information from pages 3/4 11-8 through 3/4 11-14 has been intentionally omitted. Refer to note on page 3/4 11-8..................... 3/4 11-8 LIMERICK.- UNIT 2 xvi
.!NDEX
' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
'SECTION PAGE RADIOACTIVE EFFLUENTS (Continued)
Explosive Gas Mixture....................................... 3/4 11-15
. Main Condenser.............................................. 3/4 11-16 The informa.;an on page 3/4 11-17 has br.u.
intentionally omitted. Refer to note on this page.................................................... 3/4 11-17 3/4.11.3 (Deleted) The information on pages 3/4 11-18 through 3/4 11-20 has been intentionally omitted. Refer to note on page 3/4 11-18. 3/4.11.4 (Deleted).................................................. 3/4 11-18 3/4.12 (Deleted) The information on pages 3/4 12-1 through 3/4 12-14 has been intentional'y omitted. Refer to note on page 3/4 12-1...................... 3/4 12-1 1 LIMERICK - UNIT 2 xvil
INDEX : BASES ! SECTION PAGE INSTRUMENTATION (Continued) Seismic Monitoring Instrumentation.......................... E, 3/< 3-4 (Deleted) .................................................. B 3/4 3-4 l l Remote Shutdown System Instrumentation and Controls......... B 3/4 3-5 Accident Monitoring Instrumentation......................... B 3/4 3-5 Source Range Monitors....................................... B 3/4 3-5 Traversing In-Core Probe System............................. B 3/4 3-5 Chlorirse and Toxic Gas Detection Systems.................... B 3/4 3-6 Fire Detection Instrumentation.............................. B 3/4 3-6 Loose-Part Detection System................................. B 3/4 3-6 (Deleted) .................................................. B 3/4 3-6 Offgas Monitoring Instrumentation........................... B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM.........................- B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION P INSTRUMENTATION............................................. B 3/4 3-7 Bases Figure B 3/4.3-1 Reactor Vessel Water Level........................... B 3/4 3-8 3j42 4, REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..................... ........ ........ B 3/4 4-1 , 3/4.4.2 SAFETY / RELIEF VALVES........................................ B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................... B 3/4 4-3 Operational Leakage......................................... B 3/4 4-3 3/4.4.4 CHEMISTRY................................................... B 3/4 4-3 l LIMERICK - UNIT 2 xix
INDEX BASES SECTION PAGE REFUELING OPERATIONS (Continued) ' 3/4.9.6 REFUELING PLATF0RM.................................... B .'/4 9-2 t 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L................ B 3/4 2-2 -r 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L............. B 3/4.9-2 , 3/4.9.10 CONTROL R00 REM 0 VAL................................... B 3/4 9-E 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION......... B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY......................... B 3/4 10-1 3/4.10.2 R00 WORTH MINIMlZER.................................... B 3/4 10-1 3/4.10.3 SHUTOOWN MARGIN DEMONSTRATIONS........................ B 3/4 10-1 , 3/4.10.4 RECIRCULATION L00PS................................... B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.................................. B 3/4 10-1 3/4.10.6 TRAINING STARTUPS..................................... B 3/4 10 ; 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING........ B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information on page B 3/4 11-1 has been intentionally omitted. Refer to note on this , page.................................................. B 3/4 11-1 (0eleted)............................................. B 3/4 11-2 Liquid Holdup Tanks................................... B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS - (0eleted)............................................. B 3/4 11-2 The information on page B 3/4 11-3 has been intentionally omitted. Refer to note on this page.................................................. B 3/4 11-3 (0eleted)............................................. B 3/4 11-4 LIMERICK - UNIT 2 xxii
INDEX BASES SECT!ON PAGE RADIDACTIVE EFFLUENTS (Continued) Explosive Gas Mixture................................. B 3/4 11-4 Main Condenser........................................ B 3/4 11-5 (Deleted) ........ ................................... B 3/4 11-5 3/4.11.3 (Deleted)............................................ B 3/4 11-5 3/4.11.4 (Deleted)..........................-..,............... B 3/4 11-5 3/4.12 (Deleted) The information from pages B 3/4 12-1 through B 3/4 12-2 has been intentionally omitted. Refer to note on page B 3/4 12-1...... B 3/4 12-1 i l l 1 i i l 1 l l l l 1 LIMERICK - UNIT 2 >.xiii
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area.................................................. 5-1 Figure 5.1.1-1 Exclusion Area........................ 5-2 Low Population Zone............................................. 5-1 Figure 5.1.2-1 Low f>pulation Zone................... 5-3 (Deleted)....................................................... 5-1 The figures on pages 5-4 through 5-6 have been intentionally omitted. Refer to note on page 5-4................................... 5-4 (Deleted)...,................................................... 5-1 5.2 CONTAINMENT Configuration................................................... 5-1 Design Temperature and Pressure................................. 5-1 Secondary Containment........................................... 5-7 5.3 REACTOR CORE Fuel Assemblies................................................ 5-7 Contro'1 Rod Asu:ablies......................................... 5-7 5.4 REACTOR COOLANT SYSTEM { Design Pressure and Temperature................................ 5-7 Volume......................................................... 5-8 5.5 FUEL STCRAGE Criticality.............. ..................................... 5-8 LIMERICK - UNIT 2 xxiv
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-21 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................... 6-22 , 6.15'(Deleted)................... ................................. 6-22 I i I l 1^ l
' LIMERICK - UNIT 2 xxviii 4
DEFINITIONS i i 1 LOGIC SYSTEM FUNCTIONAL itST ' l.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, soli'l state logic elements, etc. of a ' logic circuit, from sensor through and including the actuated device, to i verify OPERABILITY, The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any 1 series of sequential, overlapping or total system steps such that the entire i logic system is tested. MAXIMUM FRACTION Of LIMITING POWER DENSITY t 1.21 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest I value of the !LPD which exists in the core. MEMGER(S) 0F THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plar,t. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant. MINIMUM CRITICAL POWER RATIO 1.23 ~ The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists inthecore(foreachclassoffuel). OFFSITE DOSE CALCULATION MANUAL 1,24 The 0FFSITE 00SE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program, lhe ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Progr ws 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 6.9.1.7 and 6.9.1.8. OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE 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 required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related supm rt function (s). OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2. PHYSICS TESTS 1.27 PHYSICS lESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. LIMERICK - UNIT 2 1-4
DEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, p 9e wall or vessel wall. Pa! MARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:
- a. All primary containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3. '
- b. All primary containment equipment hatches are closed and sealed.
- c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- d. The primary containment leakage rates are within the limits of Specification 3.6.1.2. '
- e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1. -
- f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or 0-rings, is OPERABLE.
PROCESS CONTROL PROGRAM [ 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the SOLIDIFICATION or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for transportation to the disposal site and receipt at the disposal site. With SOLIDIFICATION or dewatering, the PCP shall identify the process parameters influencing SOLIDIFICATION or dewatering based on laboratory scale and full scale testing or experience. PURGE - PURGING 1.31 PURGE or PURGING shall be the contrclled 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. LIMERICK - UNIT 2 1-5
DEFINITIONS REFUEllNG FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)
- b. All refueling floor secondary containment hatches and blowout panels are
- . closed and sealed.
,E c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
- d. At least one door in each access to the refueling floor secondary containment is closed.
- e. The sealing mechanism associated with each refueling floor secondary e containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
- f. The pressure witMn the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.
REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. ROD DENSITY 1.37 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total nuinber of control rod notches. All rods fully inserted is equivalent to 100% R0D DENSITY, SHUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subtritical or would be subcritical assuming all control rods are fully inserted excen for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in tne shutdown condition; cold, i.e. 68 F; 0 and xenon free. SITE BOUNDARY 1.39 The SITE BOUNDARY shall be that line as defined in the ODCM, Figure 1.2.2-la. 1.40 (Deleted) SOURCE CHECK 1,41 A SOURCE CHECK shall be the qualitative assessment )f channel response when the channel sensor is exposed to a radioactive source.
'llMERICK - UNIT 2 1-7
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Section3.3.7.3-(Deleted) f THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-74 THROUGH 3/4 3-75 0F THIS SECTION HAVE BEEN INTENTIONALLY OMITTED. ; I I 1
=LIMFRICK - UNIT 2 -3/4 3-73
i:: Section 3.3.7.11 (Deleted). i THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED I TO THE 00CM. TECHNICAL SPECIFICATIONS FAGES 3/4 3-99 THROUGH 3/4 3-102 0F THIS 1 SECTION HAVE BEEN INTENTIONALLY OMITTED. , i l i-l i i i 1 l i I I l
- l "'ERICK - UN!T 2 3/4 3-98 l-
INSTRUMENTATION OFFGAS GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.12 The offgas monitoring instrumentation channels shown in Table
-3.3.7.12-1 shall be OPERABLE with their alarm / trip setpoints_ set to ensure that the limits of Specifications 3.11.2.5 and 3.11.2.6 respectively, are not exceeded.
APPLICABILITY: As shown in Tabic 3.3.7.12-1 _ ACTION:
- a. With an offgas monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.12-1.
- b. With less than the minimum number of offgas monitoring instrumentation Channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1. Restore l the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely mhnner in the next Semiannual Radioactive Effluent Release Report.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVf!LLANCE REQUIREMENTS 4.3.7.12. Each offgas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1. LIMERICK - UNIT 2 3/4 3-103
{ TABLE 3.3.7.12-1 b R OFFGAS MONITORING INSTRUMENTATION f c 25 MINIMUM CHANNELS [,' INSTRUMENT OPERABLE APPLICABILITY ACTION
'I. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Hydrogen Monitor 1 **
110
- 2. (Deleted)
, 3. (Deleted) 10
- 4. MAIN CONDENSER OFFGAS PRE-TREATMENT ^
RADI0 ACTIVITY MON! TOR Y jj a. Noble Gas Activity Monitor 1 ** 115
- 5. (Deleted)
. O. ! TABLE 3.3.7.12-l'(Continued) TABLE NOTATIONS j l (Deleted) 1 During. operation of the main condenser steam jet air ejector and offgas I treatment system. (Deleted) f ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the ; Minimum Channels OPERABLE reouirement, operation of main condense-offgas treatment system may continue for up to 30 days provide grab samples are collected at least once per 4 hours and analyzes within the following 4 hours. ACTION 111-114(Deleted) ACTION 115 - With the number of channels OPERABLE less than required by the i Minimum Channels OPERABLE requirement, releases to the environment may continue for up to 72 hours provided that the North Stack Effluent Noble Gas Activity Monitor is OPERABLE; otherwise, be in at least HOT SHUTDOWN within 12 hours. t
+
LIMERICK - UNIT 2 3/4 3-105
l i 4 i w INTENTIONALLY LEFT BLANK , l ll l t 1 l I 1 l t LIMERICK - UNil 2 3/4 3-106
w. g TABLE 4.3.7.12-1 0FFGlo MONITORING INSTRUMENTATION SURVEILLANCE Rf00lREMENTS E-CHANNFL MODES IN WICH E . CHANNEL . SOURCE CHANNEL
' FUNCTIONAL SURVEILLANCE
' NSTRUMENT-CHECK CHECK CALIBRATION TEST IS REQUIRED
- 1. MAIN CONDENSER OFFGAS TREATME!iT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Hydrogen Monitor D N.A. Q(3) M **
g 2. (Deleted) s. y 3. (Deleted)
- 54. MAIN CONDENSER OFFGAS PRE-TREATMENT RADI0 ACTIVITY MONITOR (STEAM JET AIR EJECTOR)
- a. -Noble gas activity monitor D M R(2) **
Q(1)
- 5. (Deleted)
~ ~. . _ . . . . _ _
. _ . _ _ . .- __~ __
TABLE 4.3,7.12-1 (Continued) , I TABLE NOTATIONS l
* (Deleted) l ** During operation of the main condenser steam jet air ejector and offgas )
treatment system, l
- (Deleted) 1 (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate thtt control room alarm annunciation occurs if any of the following conditions 6xists: ,
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), previously National Bureau of Standards, or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement , range. For subsequent CHANNEL CALIBRATION, sources that have been related ' to the initial calibration shall be used. (3) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
-1, 0.0 volume percent hydrogen, balance nitrogen, and
- 2. 4 volume percent hydrogen, balance nitrogen, f (4) (Deleted) t LIMERICK - UNIT 2 3/4 3-108
\ r 4 l INTENTIONALLY LEFT BLANK ! LIMERICK - UNIT 2 3/4 3-109
i 1 Section 3/4 11.1.1'through 3/4 11.1.4 (Deleted) ) j i l-i I l THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE 00CM. TECHNICAL SPECIFICATIONS ' PAGES 3/4 11-2 THROUGH 3/4 11-6 0F THIS } SECTION HAVE BEEN INTENTIONALLY OMITTED. l l t I 4 l
. t U4EO ' - UNIT 2 3/4 11-1
i: i i l Secton3/411e201throughSection3/411c2.4(Deleted) 4 i 1 l i l I i THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECFICATIONS ! PAGES 3/4 11-9 THROUGH 3/4 11-14 0F THESE > SECTIONS HAVE BEEN INTENTIONALLY OMITTED. t I
'i n 1CV T '
3/4 11 8
e: ,,_ , g :g 9 [' , oSection 3/411'-2.7)(Deleted) f, <l ., P
.W),
y i L
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THE lNFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS RELOCATED TO THE 00CM.- 1 i
'k'
, . LIMERICK.'- UNIT 2 3/4 11-17
- L c:a. , ,
s,. [9 Sect 6n-3/411.3-tbrough--3/411.4i(Deleted)L 1 , THE.lHFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED-
~
TO-THE PCP. TECHNICAL SPECIFICATIONS PAGES ;- 3/4 19 THROUGH 3/4 11-20 07 THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
.c
.y k
4 4
. LIMERICK' JJHIT 2 3/* 11-18
-' f , $ _
q r
, . Sectiont3/4.12'(Deleted);
1 .
- 4 \;> l' -
s 4 i
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^i,..<
i
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THE INFORMATION FROM-THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE 00CM. TECHNICAL SPECIFICATIONS . . PAGES 3/4 12-2 THROUGH 3/4 12-14 0F THIS : SECTION HtWE BEEN INTENT 10NALLY'0MITTED,
\.
I LIMERICK -' UNIT 2 3/4 12-1
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INSTRUMENTATION'- 1 BASES ; l
- 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION l s
The reactor core isolation cooling system actuation instrumentation is provided to initiate. actions to assure adequate core cooling in the event of reactor! isolation from its primary heat sink and the loss of feedwater flow to the reactor. vessel. This-instrumentation does not provide actuation of any of the emergency core. cooling equipment. Operation with a trip set less conservative than its Trip Setpoint but j 'within its specified Aliowable Value is-acceptable on the basis that the- 3 difference between each Trip Setpoint and the Allowable Value is an allowance a for instrument drift specifically allocated for each trip in the safety enalyses.- 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block funtcions are provided consistent with the : requirements of the specifications in Section 3/4.1.4, Control Rod Frogram Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block, f Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the' basis that the difference between each Trip Setpoint and the Allowable Value is an allowance - for instrument drif t specifically allocated for each trip 'in the safety analyses. , 3/4.3.7 MONITORING INSTRUMENTATION '3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION
-The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action-is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables.
following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64. 3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic _ monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the unit. 3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTivh RELOCATED TO THE 00CM. I LIMERICK - UNIT 2 B 3/4 3-4
'INSTRUNFHTATION J ,i BASES: -
).
~3/4.3.7.4 REMOTE SHUTDOWN' SYSTEM INSTRUMENTATION AND CONTROLS
' The OPERABILITY of the: remote shutdown system instrumentation and; _
controls-ensures ~that sufficient capability is available to permit shutdown
-and maintenance of-HOT SHUTDOWN of the unit from= locations outside of the control room. This capability is required in the event control room habitab*i?ty is lost and;is consistent with General Design Criterion 19 of.
T l10 CFR Part 50, Appendix A. The Unit 1 RHR transfer switches are included fonly-due to their potential impact on the RHRSW system, which is common to both units.
-3/4.3.7.5- ACCIDENT MONITORING INSTRUMENTATI0h i
. The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendatior! 4 Regulatory Guide 1.97
" Instrumentation for Light Water Cooted Nuclear _ Power Plants to Assess
- Plant Conditions During and Following an Accident." December 1975 and' NUREG-0737, " Clarification of TMI Action Plan r.g irements," November 1980, 3/4.3.7.6 SOURCE RANGE MONITORS The_ source range monitors provide the oper.ste with information of the
? status of the neutron level in the core at very low power levels during-startup and shutdown. At these power levels, reactivity additions shall not be made without.this flux level information available to the operator.
When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted. 3/4.3.7.7 -TRAVERSING IN-CORE PROBE SYSTEM
.The'0PERABILITY of the traversing in-core probe system with the
.specified minimum' complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial
' neutron: flux distribution of the reactor core.
The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,' detectors) prior to performing an LPRM calibration function.- LMonitoring core thermal limits may involve utilizing individual detectors
'to monitor selected areas of the reactor core, thus all detectors may not
-be required _to be OPERABLE. The OPERABILITY of individual detectors to be
- used'for monitoring is -demonstrated by comparing the detector (s) output in the. resultant. heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).
LIMERICK - UNIT 2 B 3/4 3-5
c -INSTROMENTATION
.i BASES-j l .
1
'3/4.3.7.8 CHLORINE AND T0XIC GAS OETECTION SYSTEMS !
The OPERABILITY of-the chlorine and toxic gas detection systems ensures that an - accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlorine and manually _ initiated for i toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically c be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection. The detection sy;tems required by this specification are consistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control l Room Operators against an Accidental Chlorine Release," February 1975. i 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION 0PERABILITY of the detection instrumentation ensures that both adequite warning capabilitv is available for nrompt detection of fires and that fire suppression systems, that are dctuated by fire .atectors, will discharge extinguishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for.early fire warning and notification. Consequently, the minimum number of OPERABLE fire detectors must be greater. The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage'than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
,3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.
3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE 00CM. LIMERICK - UNIT 2 8 3/4 3-6
S'
. , INSTRUMENTATION 2
BASES 3/4.3~.7.12'0FFGAS MONITORING INSTRUMENTATION' This instrumentation includes provisions for monitoring the concentrations of ' 1potentially explosive gas mixtures and noble gases in the off-gas system.
- 3/413.8 TURBINE OVER'$ PEED PROTECTION SYSTEM
This, specification isLprovided'to' ensure that the t'irbine overspeed protection system. instrumentation and the turbine speed control valves are OPERABLE and will Ei 1 protect the turbine'from excessive overspeed. Protection from-turbine excessive: L overspeed is required since excessive overspeed of the turbine could generate
-potentially damaging missiles which could impact and damage safety related
- components, equipment or structures.- !
I 13/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION
.The ;feedwater/ main turbine trip system actuation instruw:ntation.is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand.
L L l i m 4 LIMERICK - UNIT 2 8 3/4 3-7 l t L _--2____ - , ,
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' 3/4.11.1.4 LIQU10 HOLOUP TANKS The! tanks listed n this specifit.ation include all those outdoor radwaste tanks that are not surrounded-by liners, d me;, or walls ~ capable of holding the tank
- contents and.that do not have tank overflows and surrounding. area drains connected to the' liquid radwaste treatment system.
Restricting 1the quantity of radioactive materiai-contained in the specified tanks .
= provides assurance:that1in-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 1II Column 2, at the nearest portable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
3. o A r g i iLIMERICK' - UNIT 2 B 3/4 11-2 I r
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,1RADI0 ACTIVE EFFLUENTS '
BASES.
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.I THE~INFORMATION FROM THESE SECTIONS HAS BEEN RELOCATED TO THE ODCM.
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RADIOACTIVE EFFLUENTS-' . <
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I I e i 3/4.ll.2.5' EXPLOSIVE GAS MIXTURE This specification-is~provided to ensure that the concentration of potentially , explosive gas mixtures contained in~the' main condenser offgas. treatment. system is maintained below the flammability limits of hydrogen and c tygen. Maintaining the
= concentration of hydrogen below'its flammability limit provides assurance that.the;- '
releases of radioa'tiva materials will be controlled-in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. , 4 LLIMERICK - UNIT:2 B 3/4 11-4 ,
!J - RADIDACTIVE EFFLUENTS
' BASES 3/4.11.2.6 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides recsonable assurance that the total body exposure to an individual at the exclusion _ area boundary will.not' exceed a small fraction of the limits of.10 CFR Port 100
, in the-event this-effluent is inadvertently discharged directly to the environment without
' treatment. This specification-implements'the requirements of General'0esign Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
~
3/4.11.2.7. 3/4.11.3, and 3/4.11.4 (Deleted) - INFORMATION FROM THESE.;f.CTIONS RELOCATED TO THE 00CM. i 4 i: t i-i s LIMERICK - LINIT 2 B 3/4 11-5 'I 0. f-
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- %, -3/4.12 R DIOLOCICAL A ENVIRONMENTAL' MONITORING ~
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1 Section 3/4.12 (Deleted); i THE-INFORMATION FROM THIS SECTION HAS BEEN RELOCATED.T0 THE ODCM. BASES PAGE B 3/4=12-2 HAS BEEN INTENTIONALLY 0MITTED.
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2 1 LIMERICK'- UNIT 2 B 3/4 12-1
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t 5.1-- SITE ' i EXCLUSION AREA
'5.1.1' The exclusion: area shall be as shown in Figure 5.1~.1-1.
'LOWPOPULATIONZbHE I s 5.1.2 :The low population zone shall be as shown in Figure 5.1.2-1.-
-5.1.3 _(Deleted)'
5.1.4'(Deleted)-- 5.2 CONTAIN!1ENT
' CONFIGURATION 5.2.1
~
The primary containment is a stael lined reinforced concrete structure consisting of a drywell and suppressier chamber. The drywell is a steel-lined reinforced concrete: vessel in a shape-of a truncated cone on top of'a water filled-suppression chamber.and is separated by a diaphragm slab and connected to the- l ( suppression-chamber through a series of downcomer vents. The drywell nas a maximum free air volume of 243,580 cubic feet at a minimum suppression pool level
'of 22 feet. The suppression chamber has a maximum air region of 159,540 cubic :
feet =and.a minimum water region of 122.120 cubic feet. 3 DESI_GN-TEMPERATURE AND PRESSURE
- 5.2.2' The primary containment is designed and shall be maintained for:
- a. Maximum internal pressure 55 psig.
-b. Maximum internal temperature: drywell-3400F.
suppression pool 2200 f. <
- c. Maximum external to internal differenttal pressure ti psid, >
- d. -Maximum floor differential pressure: 30 psia, downerd.
20 psido upw.ty d . h t LIMERICK --UNIT 2 5-1 _ _ = _ _ -. _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - - _ _ - - _ _ _ - _ - - _ _ -
ih-.,, . . . , , ti s s I i . 7-THE FIGURES ON PAGES 5-4 fHROUGH 5-6 .: NAVE bEEN RELOCATED TO THE ODCM. , PAGfS 5-5 AND 5-6 ARE INTEN'iONA!.iY l OMIT 1EO, l k '. y
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;ADMINISTRATIVECONTROLS PROCEDURES AND PROGRAMS (Coatinued)
- d. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control
-of radioactive effluents.and for maintainitig-the doses to MEMBERS OF THE PUBLIC-from radioactive efficents as low as reasonably achievable. The program (1) shall be contained in the 00CMr (2) shall:be implemented by operating procedures, ar.d (3) shall . include remedial actions to be taken whenever the program limits are exceeded. The program shall includ2 the!
following elements:
- 1) Liteitations on the operability of radioactive liquid and gaseous
.a,Laitoring instrumentation including sarveillatte tests and setpoint s
s determinetioh in accordance with the aethodo!cgy in the OCCM,- 3
's k) . Limitations on itse concentrations? of radio.activtretertal relea"w in 1!(uid effluent.s to UNRESTRICTlID AREAS conforming to.10 CFR b >
- 1[h i .
q 20. AppenMa B. Tabic II, Colume 2, Manitoring, smplito and malysis of esdioact'ive liquid and gaseous 3) effluents in Accordmce Wth 10 CFR 20,106 end with the methodology @ and parac ters in the CDCW,, j; o
- 4) Liettetion$ en the amy,:a13rvi quarterly doses or dose commitwnt to a MOtBER OF THE PUBl.lC from radicactive Naterlah in lfquid
; f h
ef flusti celcased from each unit to UNRESTMCYE0 AWG confmsing. ) k w to Appendix I cto 10 CM Part 50, g
'l S): Getermiration of cumulative and pnjected (ose ctmtributhus frw <
j rMiogetive effluents for the current ca;en6ar quarter med currant : ' R 0 l t.y' calendsr year in accurdance with the' methodology aci paramettr*: in. .< 4 f, the 00CM at'lesst every 31 days - a.
'6) Limstations on the operability and use.of the iiqUid. ara 3aseous -
l l' effluent treatment systeus to enture 'that the uppr6priate portiont of these systest are used to redvce M1dese:. of radioactivity vfaen ]- J
')1 l the projected doses in a 31-day persed w>uld exceed 2 percent of the -
gui@. lines, for the annua'i aum or dose committent conforicio) to l
; i
^ ppendtx l'to 10 'JFR Part 50e '
; , N
] 7) Limitat$ons en tr.c dosa rate resulting from rsdioactive material ( i I
H released in vaseous et f!uents to areas teuond the SITE BOUNDAP1 h [ conforming to the doses assoclated 9ith 13 CFR Cart '20., Appendix 8, , j
! - Table 11 Colum 1, 1 Y
_ 1 B_ 2 J - _ _N_ _s {
- ADMINISTRATIVE CONTROLS NRitT50RES AND - PRf4 RAMS _ (Continued) 8)f Limitations on the annual _and quarterly air doses resulting from noble gases released in~ gaseous effluents from each unit to areas beycad the SITE.
B0UNDARY conforming to Appendix I to_10 CFR Part 50,
- 9) _ - Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from lodine-131. Iodine-133.,tritian, and all radionuclides in particulate form-with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR-Part 50, t
- 10) Limitations on venting and purging of the Mark II containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable, and
-11) Limitations on the annual dose or dose commi hent 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.
[ e. &cteorological Monitoring Program g A orogram shall be provided to provide meteorological information in the-
- environs of the plant. The program shcIl provide sufficient meteorological 7 datG fCr estimating potential radiation doses to the public.
c 1 'H. . The program shall (1) be contained in the 00CM, (2) confore to the guidance c7 Regulatory G:uide 1.23. " Safety Guide 23 - Onsite Meteorlogical Program", 3 and (3) include limitations on the operability of meteorological monitoring g insixuoentation including surveillance tests in accordance with the h - methW ology in the 00CM. }l
.- .f ' ladiolqcpefjJr3jygntal_ Monitoring Program j
h[H 4 A progr.Fi shalI be provided to !uonitor the radiation and radionuclides- in the envYront of the plant. The program shall provide (1) representative 1 Id meaturements of radioactivity in the highest potential exposure pathways, ! L. ( and (2) ver2fication of the 4ccuracy of the effluent monitoring program-and ' tx codeling of environental expc a re pathways, The program shall-(1) be (] L contained in the 00CM, (2) conforte to the guidance of Appendix I to 10 CFR , I 1 '
~
, Part SO, and (3) include the folicwing: ,
k' ,,[ w ~-
- 1) Monitoring, sampling, analysis, and reporting of radiation and 4 i
f p radionuclidei in the en,ironment in accordance with the methodology and l parameters in the 00CW, I 2) - A Land use Census to ensure that-changes in the use of areas at and l pl s
-% yond the SITE 1100MDARY are identified and that modifications to the monitoring program are unde if required by the results of this census, g( and
';N
- 3) Participation in a interlaboratory Coniparison Program to ensure i. hat l independent checks on the precision and accuracy of the measurements of A .,
radioact1we materials in envirotwental sample matrices are performed as pert of the quality assurance propia for environmental monitoring. I Ip e Y a 1 f i3M5RNLI-EMIf~2 6-14b I
- 1 ,w.
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ADMINIS1RATIVE: CON 1ROL4h
. ANNUAL' REPORTS (Continued)
' dosimeter; thermoluminescent dosimeter (TLD), or film badge measurements. Small' exposures totaling less than 20%'of the f individual total dose need not be accounted for.< In.the=
', aggregate, atcleast 80% of the total whole-body dose received from external sources should-be assigned to specific major work functions;
- b. Documentation of all challenges to safety / relief valves; and c.- Any other unit unique reports requjred on an. annual basis,
- d. 'The results of specific activity analysis in which the primary-coolant exceeded the limits of Specification 3.4.5.
The following information shall be included: (1) Reactor power history starting-48 hours prior to the first sample in which the. limit was exceeded; (2) Results of the last-isotopic analysis for racloiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup system flow' history starting 48 hours' prior to the first sample in which the limit w e exceeded; (4) Graph of the 1-131 concentraticn and onc ether radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific acih.ty above the steady-state level; and (5) The l'.ine duration wnen the specific activity of the primary coolant exceeded the radioiodine limit.
' MONTHLY'0PERATING REPORTS *
-6.9.1.6 : Routine. reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam system safety / relief. valves, shall be submitted on a monthly
-basis to the ATTH: Document Control Desk, 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 15th of each month following the calendar month covered by the report.
_ ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING PEPORT* 6.9.1.7 The Annual Radiological Environmental Operating Ren.M covering operation of the unit during the previous calendar . ear shall
'be submitted before May 1 of each year. The initial report snall be submitted prior to May 1 of the year fol' lowing initial criticality.
The report shall include summaries, interpretations, and analysis of trends:of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM and-(2) Sections IV.B.2, TV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
*A single submittal may be made for a multiple unit station.
LlHERICK - UNIT 2' 6-16 l t e
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- ADMINISTRATIVE CONTROLS
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i. I SEMIANNUAL-RADI0 ACTIVE EFFLUENT RELEASE REPORT *
.6.9.1.8 The Semiannual Radioactive Effluent Release Report covering the operaticn 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 initial report shall-be submitted during firstureport period, as described above, followingLinitial
-criticality. The report shall include a summary of the quan;ities of radioactive-liquid and gaseous effluents and solid waste released from the unit. The material
'provided'shall:be (1)' consistent with the objectives outlined in the 00CM and PCP-
.and (2) in'conformance with 10 CFR 50.36a and Section'IV.B.1 of Appendix ~l to 10 CFR Part 50. ;
{' i
*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 sthuion.
LIMERICK - UNIT 2- 6-17
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-6.9.2 -Special reports.shall be submitted to the Regional Administrator of the
. Regional Office of the NRC within the time period specified for each report. !
- LIMERICK - UNIT 2 6-18 N- ,
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7 YADMINISTRATIVELCONTROLS. i RECdRDRETENTION(Continued); d .- -Records of gaseous and liquid radioactive material released to the environs. I
- e. Records of transient' or operational cycles ~ for those unit ccmponents identified .in Table 5.6.1-1.
f.. Records of reactor tasts and experiments. e g... Records of training and qualification;for' current members of the unit staff. 1 h.- Records of inservice _ inspections' performed pursuant to these-Technical i Specifications. a
- 1. Records of quality assurance activities required by the Operational _ Quality.
l Assurance Manual not listed in Section'6.10.2. J. Records of reviews performed for changes made to procedures or equipment or reviews j of tests and experiments pursuant to 10 CFR 50.59. '
.i
- k. RecordsLof' meetings of the PORC and=the NRB. I
- 1. Records of the service lives of all snubbers including the date at which the service life commences and associated installatien and maintenance records.
- m. Records of analysis required by the Radiologica) Environmental Monitoring Program.
that would permit evaluation of the accuracy of the analysis at a later date. . 1 l
- n. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION :
MANUAL and-the PROCESS CONTROL. PROGRAM. , 6.11 RADIATION PROTECTION PROGRAM 1 1 '6.11.1 Procedures-for personnel radiation protection shall be prepared consistent with the' ~ l' requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all ; operatiors i_nvolving personnel radiation exposure. i l: : 6.12' HIGH RADisTION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2)- ; of=~10 CFR Part 20, each high radiation area in which the intensity of radiation is greater ' than 100 mrem /h but less than 1000 mrem /h shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals pereitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring tevice which continuously indicates the radiation dose rate in the crea.
1 l l- ,
- Health physics personnel or personnel escorted by health physics personnel shall be exempt l t
.from the RWP issuance requirement during th performance of their assigned radiation I protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.
L 1 LIMERICK - UNIT 2 6-20 I
7 ADMINISTRATIVE CONTROLS i
~ HIGH RADIATION AREA (Continued)-
~
- 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 ".evels in the area have been established and personnel have been made knowledgeable of them. ,
-c. A health physics qualified individual (i.e., qualified in radiation protection procedures) 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 Health Physicist'in the RWP.
6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could
> 1 receive in=1 hour a dose greater than 1000 mrems shall be provided with locked 7 doors to prevent unauthorized entry, and the keys shall be maintained under the
.y administrative control of Shift Supervision on duty and/or the health physics supervision. Doors shall remain locked except during periods of. access by personnel under an-approved RWP 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 1 hour a dose in excess of 1000 mrems* that are located within large areas, such as the 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 cf the RWP, continuous surveillance direct or remote (such as use of closed circuit TV cameras), may be made by personnel
; qualified _in radic~ ion protection procedures to provide positive exposure control over the activities within the area.
6.13- PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3n. This documentation shall contain:
- 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and r
- Measurement made at 18 inches from source of radioactivity.
LIMERICK - UNIT 2 6-21
U ADMIN'ISTRATIVEl CONTROLS- 'Pg 6-22/2874 .RWG-PROCESS CONTROL PROGRAM'(Continued) a 2. A determination that the change did not reduce the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulatory,
- b. Shall become effective upon review and acceptance by the PORC and=the !
approval of,the Plant Manager..
.l 6.14 0FFSITE DOSE CALC 9LATION MANUAL'(00CM) 1 6.14.1 Changes to the ODCM: .;
o a.. Shall be documented ard records of reviews performed shall be retained as j required by Spacification 6.10.3n. This documentation shall contain:
- 1. Sufficient information to support the change together with the appropriate; analyses or evaluations justifying the change (s) and
- 2. A determination that the change will meintain.the level of l 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 adverseiy impact the accuracy or reliability of effluent, dose, or setpoint calculations. -
! b. Shall.become effective upon review and acceptance by the PORC and the approval of the Plant Manager. i L
- c. Shall be submitted to the Commission in the form.of a' complete, legible I cepy of the entire 00CM 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 l markings in the margin of the affected pages, clearly indicating the area I of the page that was changed, and shall indicate the date (e.g.,
l- month / year) the change was implemented. 6.15 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM. a . I i
- LIMERICK - UNIT 2 6-22
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- i. INDEX' DEFINITIONS . _
SECTION' DEFINITIONS (Continued) PAGE 1.22 MEMBER (S) 0F THE PUBLIC...................................... 1-4 1.23 MINIMUM CRITICAL POWER RATI0................................. 1-4
- 1. 2% OFFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.25 OPERABLE - OPERABILITY....................................... 1-4 1.26 OPERATIONAL CONDITION - CONDITION............................ 1-4 1.27 PHYSICS TESTS............... .............................,.. 1-4 1.28 PRESSURE BOUNDARY LEAXAGE..........................y........ 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY................................ *1-5 .
1.30 PROCESS CONTROL PR0 GRAM...................................... 1-5
- 1.31 PURCE - PURGING....,......................................... 1-5 1.32 RATED THERMAL P0WER........ ................................. 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY............ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME...................... 1-6 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY.............. 1-6 1.36 REPORTABLE EVENT............................................. 1-7 1.37 ROD DENSITY.................................................. 1-7 1.38 SHUTDOWN MARGIN......................................:....... 1-7 1.39 SITE B0VNDARY........................................ ....... 1-7
- 1. 41 SO U R C E C H E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7
'1.42 STAGGERED TEST BASIS......... ......................t........ 1-8 1.43 THERMAL P0WER................................................ 1-8 1.44 UNIDENTIFIED LEAKAG'E......................................... 1-8 LIMERICK - UNIT 2 ii AUG 2 5 tm .
1 INDEi'
' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements...................... 3/4 3-66 Seismic Monitoring Instruments. tion...................... 3/4 3-68 Table 3.3.7.2-1 Seismic Monitoring q
. Instrumentation....................
3/4-3-69 't Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance ., Re,quirements.....................: 3/4 3-71
- Me foto " cal N itoring strume ion. m ' . . q., .. . . .
3-73 1 (T 3e- .A31 teor ical Mo ring i
- - In mentation................... 3/4-3-74 le 4< -
Meteoro nitor Inst' tion S nee ,
. l' s u i reme nt s . . . . . . . . . . . . . . . . . . . ~. . . - 3/4 3-75 Remote Shutdown System Instrumentation and Controls..... 3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls...... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance ,
Requirements...................... 3/4 3-83 Accident Monitoring Instrumentation..................... 3/4 3-84 Table 3.3.7.5-1 ' Accident Mon,itoring Instrumen-tation............................ 3/4 3-85 Table 4.3.7.5-1 Ace'ident Monitoring Instrumenta- - tion Surveillance Requirements.... 3/4 3-87 Source Range Monitors................................... 3/4 3-88 Traversing In-Core Probe System......................... 3/4 3-89 , Chlorine Detec. tion System............................... 3/4 3-90 . Toxic Gas Detection System.............................. 3/4 3-91 Fire Detection Instrumentation.......................... 3/4 3-92
' LIMERICK - UNIT 2 ix AUS 2 5 ISH
-1
, 1 INDEk .
j LIMITING C0NDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION
.PAGE INSTRUMENTATION (Continued)
Table 3.3.7.9-1 Fire Detection Instrumentation.... 3/4 3-93, Loose Part Detection System............................. 3/4 3-97
. tive- uid Ef usntsMo ring'I umen-on. .. .... .... ... .... ....... ....... . . . ; .#98 3 u /4 3 Tabl 4.3.7.11 Radioactive Liquid Effluen~ .-
Mo n n urveillance Requirements........ 3/4 3-101-fA . PM .'.W2e Gas r: 5"lart Moni Instrumen-tation...........................toring ....................... 3/4 3-103 L
, Gu c. < < ,
Table 3.3.7.12-1 AMioactive Gaseewo E"!utat i Monitoring Instrumentation....... 3/4 3-104 Table 4.3.7.12'-1 P I!INeGasece: Eff! = t Monitoring' Instrumentation i Surveillance Requirements........ 3/4 3-107 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM. . . . . . . . . . . . . . . . .3/4 . . .3-110 3/4.3.9 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................... 3/4 3 112 Table 3.3.9-1 Feedwater/ Main Turbine Trip
. System Actuation Instrumentation.... 3/4 3-113 ;
Tabie 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-i- tation Setpoints.................... i ' 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillance Require-ments............................. 3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION.SiSTEM Recirculation Loops..................................... 3/4 4-1 L LIMERICK - UNIT 2 x
, AUG 2 5 &;
l INDEX 1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l SECTION PAGE J
, -REFUELINGOPERATIONS(Continued) 3/4.9.10 CONTR0t ROD REMOVAL Single Control Rod Remova1.............................. 3/4 9-13
~
Multiple Control Rod Remova1............................. 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level........................ 1.............. 3/4 9-17 Low Water Leve1......................................,... . 3/4 9-18 , 3/4.10 SPECIAL TEST EXCEPTIONS ,. i 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY........................... 3/4 10-1 3/4.10.2 R00 WORTH MINIMIZER..................................... 3/4 10-2 3/4.10.3 SNUTDOWN MARGIN DEMONSTRATIONS.......'................... 3/4 10 3 . 3/4.10.4 RECIRCULATION LOOPS .................................... 3/4 10-4 3/4.10.5 OXY, GEN CONCENTRATION.................................... 3/4 10-5 3/4.10.6 TRAINING STARTUPS........'............................... 3/4 10-6 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.... ..... 3/4 10-7 3/4all RADIOACTIVE EFFLUENTS 3/4.11.1' LIOUID EFFLUENTS Concentration...........................................
. 3/4 11-1 Tat >1e'4~.11.1.1.1-1 Radicactive Liquid. Waste l Sampling and Analysis P r o g r am . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-2 Dose,...................................................
. 3/4 11-5
, Liquid Radwaste Treatment System........................ 3/4 11-6 Liquid Holdup Tanks..................................... 3/4 11-7 l . 3/4.'11.2 GASEOUS EFFLUENTS Dose Rate............................................... 3/4 11-8 LIMERICK - UNIT 2 xvi Mi 2 5 mi
V . ,
.e.
INDEX' r LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADI0 ACTIVE EFFLUENTS (Continued) Tab .11.2.1 Radioac e Gaseous Waste ! Sam cond An s l gram... .................. 3/' 11 ' N'e " ele . 1/4 11-12 11 12___
- l I
- _... M ExplosiveGasMixture................................'[.'. 3/4 11-15 Main Condenser................................'....,'...... 3/4 11-16 Ve @ .... - _3/4 11 (
RfAT ni. i
@h- 11-20 ,
3/4.121 RADIOLOMC4L ENVIRO M rAL MONITO MNG
- 12. TORI AM.... ......... ........... .... 3/4 12-1
/ Ta le 3.12.1- Rad'ological Mnental ram..'.....
, f Mon toring ....... 3/, 2-3 l
' ab 3.12.1-2 Rep ng Levels For -
- p
- tivi ConcentratMnsIn Environ ental Samples. . . . . . . . . . . . . /4 12-9 Table . .1-1 _ Detection apabilities For
/ Environmept ample Analys . .... 3/4 12-10 <
f3/ .1 LAND 5 ENSU.........,.7 ......... ..... ......... ... 12-13 3 .12.3 ERLABORA Y p R00 RAM...................... 3/4 12-14 n L LIMERICK - UNIT 2 xvii AU6 2 51999
\
t INDEX BASES SECTION PAGE . 4 INSTRUMENTATION (Continued) , Seismic Monitoring Instrumentation...................... B 3/4 3-4
. M 1
- i Remote Shutdown System Instrumentation and Controls..... B 3/4 3-5 l l
Accident Monitcring Instrumentation..................... B 3/4 3-5 Source Range Monitors..........................'......... B 3/4 3-5, Traversing In-Core Probe System.......................,,. B 3/4 3-5
' 1 Chlorine and Toxic Gas Detection Systems......::........ B 3/4 3-6 l Fire Detection Instrumentation.......................... B 3/4 3-6 Loose-Part Detection System............................. B 3/4 3-6 fRadioact Liquid fflue t Monitoring Mns4 nta .... .... ,. ... . ........ h
%ww;tiv:
R;..;; a Gase: : E " t.;nt Monitorin Instrumentation.......................g. . ................ B 3/4 3-7
~
'3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM..................... B 3/4 3-7 3/4'. 3. 9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION......................................... B 3/4 3-7 Bases Figure B 3/4.3-1 Reactor Vessel Water Leve1...................... B 3/4 3-8 1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECI RCU LATION SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE -
Leakage Detection Systems............................... B 3/4 4-3 , Operational Leakage..................................... B 3/4 4-3 3/4.4.4 CH'EMISTRY..................................... ......... B 3/4 4-3 LIMERICK . UNIT 2 xix AUG 2 5198) ____i__.-________i_.____ ,_,c -
i INDEX BASES SECTION. PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATF0RM................................ B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT- FUEL STORAGE P00L. . . . . . . . . . . . B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL . . and WATER LEVEL - SPENT FUEL STORAGE POOL., ... B 3/4 9-2' 3/4.9.10 CONTROL ROD REM 0 VAL........................ ..... B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION....., B 3/4 9-2
. 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY-CONTAINMENT INTEGRITY....................s. B 3/4 10-1 3/4.10.2 ROD WORTH MINIMIZER...................... ........ B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.................... B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS............................... B 3/4 10-1 !
3/4.10.5 OXYGEN CONCENTRATION.............................. B 3/4 10-1 ;
'3/4.10.6 TRAINI,NG STARTUPS................................. B 3/4 10-1 -
3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.... B 3/4 10-1
-3/4.11_RADI0 ACTIVE EFFLU M L , \l/3 i NTS M on. 1-1 l
.......... c ..... .B_ur n-1 -
l Liquid ste Treatment System.................. B 3/4 11-2 ~ s
..................... B 3/4 11-2 g3/4.1 Q E0VS EFFLUENTS -
u %, M t!.. % ................................. 1-2
~ -
Dose - Noble es............................ B 3/4 11-3 ose - Iodine-131, Iodine-133, Tritium, and Radionuclices in Particulate Form................. B 3/4 11-3 ion Exhaust Treatment System.............. .B 3/4 11-4 LIMERICK - UNIT 2 xxii A'.is 2 5 E l .
b . :. INDEX* .
^
BASES SECTION . PAGE RADI0 ACTIVE EFFLUENTS (Continued) Explosive Gas Mixture.............................. B 3/4 11-4 Main Condenser..................................... B 3/4 11-5 ax ne 3I4.}t RADIOLOGICAL _ ENVIRONMENTAL MONITORJNG . V /4.d1 M0(1RINGe. gum....z.<.......................
.73/412-1 3/ .12 2 No sE Nsus .. ....... .... ...... ... ..... ...
93/4 2-1r
/4.12.3 INTE ORATORY COMPARISO A0 GRAM................. E 4 12-2 i 4
~
l l l 1 l LIMERICK - UNIT 2 xxiii A'.'S 2 51.9
1 INDEX-4 DESIGN FEATURES ,
~ SECTION PAGE I 5.1 SITE
. Exclusion Area............................................... 5-1 Figure 5.1.1-1 Exclusion A*ea.......................... 5-2 Low Population Zone................. ........................ 5-1 Figure 5.1.2-1 Low Population Zone..................... 5-3 Maps Definin UNRESTRICTED AREA and SITE BOUNDARY for.
Radioactive seous and Liquid E 1uents.........h.......... 5-1
\ -
Figure 5.1. -la Map Defining NRESTRICTED AREAS for , Radioactive G eous and Liquid-l Effluents...... ................. ..... 5-4 L igure 5.1.3-lb Map Defining UNR TRICTED AREAS for ( adioactive Gaseou and Liquid
- l. . E luents........... .................. 5-5 Meteorologi 1 Tower Locatl ................. ..............- 1 .
Figure 5. 4-1 Meteorolog al Tower Location........... 5-6 5.2 CONTAINMENT Configuration................................................ . 5-1 Design Temperature and Pressure.............................. 5 L Secondary Containment........:............................... . 5-7 5.3 REACTOR CORE Fuel Assemblies.............................................. 57 L Control Rod Assemblies........ :............................. 5-7 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature.............................. 5-7 Vo1ume.............,.........................................- 5-8 5.5 FUEL STORAGE . Criticality.................................................. 5-8 . LIMERICK - UNIT 2 xxiv E 2 ! 1.:ii
s
\' -o -
3l i 4 INDEX
' ADMINISTRATIVE CONTROLS !
SECTION-PAGE 6.13 PROCESS CONTROL PROGRAM (PCP).............................. 6 -! 6.14 0FFSITE DOSE CALCULATION RANUAL (0DCM)..................... 6-22 (4.E v MAJefh v CHANAES TO 14DIOKTIv3AtRIL.TMhUM . .... 6-v v - - 1
'i l
l l l-
. =
1 . LIMERICK - UNIT 2 xxviii kR 2 5 UH e
DEFINITIONS ' LOGIC SYSTEM FUNCTIONAL TEST : 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, I i.e., all~ relays and contacts .' trip units, solid state logic elements, etc of uc e r through and including the actuated device,alogiccircuit,fre to verify OPERABILI . . The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire log'ic system is tested. MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.21 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLP0 shall be the highest - value of the FLPD which exists in the core. HEMBER(S) 0F THE PUBLIC 1.22 MEMBER (5) 0F Th'E PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to niake deliveries. This category does include persons who use portions cf the site for recrea- ,
, tional, occupational, or other purposes not associated with the plant.
MINIMUM CRITICAL POWER RATIO 1.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core Tfor each class of fuel). OFFSITE DOSE CALCULATION MANUAL 14The. 'The OFFS!TE D0$E CALCULATION MANUAL (00CM) shall contain the methodology . and parameters used in the calculation of offsite doses re:.ulting from radic-active gaseous and liquid effluents, in the calculation of gaseous and liquid , t effluent monitoring Alarm / Trip setpoints, and in the conduct of the Enviroa- -- Dd 3 ; , mental Radiological Monitoring Program. The Ot>CM shall also contain (1) the ' he Radioactive Effluent Controls and Radiological Environmental Monitoring Pro-grams required by Section 6.8.4 and (2) descriptions of the infcrmation that > should be included in the Annual Radiological Environmental cperating and Semi- ' l' annualRadioactiveEffluentReleaseReportsrequiredbyspecifications6.5.1.37
? _. and6.9.1.4T 1.6 a system, suusystem, T. rain, componem, ur device snoi n vi urtnxott or nave OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubr.ication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its functibn(s) are also capable of performing their related support function (s).
OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.
- PHYSICS TESTS 1.27 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 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
LIMERICK - UNIT 2 1-4 AUG 2 5 E
DEFINITION!' PRES $URE FOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall. PRIMARY CONTAINMENT INTEGRITY
, 1.09 PRIMARY CONTAINMENT INTEGRITY shall exist when:
- a. All primary containment penett'ations required to be closed during accident conditions are either: -
- 1. Capable of being closed by an OPERABLE primary containment . 1 automatic isolation system, or ;
- 2. Closed by at least one manual valve, blind flange, or 1 deactivated automatic valve secured in its closed position,- '
. except as provided in Table 3.6.3-1 of' Specification 3.6.3.
- b. All primary containment equipment hatches are closed and sealed.
- c. The. primary contaSment air lock is in compliance with the requirements of Specification 3.6.1.3. .
- d. The primary containment leakage rates are within the limits of !
Specification 3.6.1.2. !
- e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
- f. The sealing mecharism associated with each primary containment penetration; e.g. welds, bellows, or 0-rings, is OPERABLE.
. PROCESS CONTROL PR') GRAM
~
, /.ypi.M Thf >RbtE'5HO'NTf6C7ADGAWlPCPf'shall conYatn the'curreht ttrmWas. .
sampling, analyses.- test, and determinations to be made to ensure that precan.- j ing and packaging of solid radioactive wastes based on demonstrated prccessi..2 { ,
; of actual or simulated wet solid wastes will be accomplished in such a way ns to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, N4 !,
j burial ground requirements, and other requirements solid radioactive waste. ~ governing the disposal of *( odoEidifkE~'tio a 'agenitilwaste nd/or nicess ry additives r . .~ . . , of ' ticipited w te, ans' the acc ptable' boundary condit sf the process para eterg'shall e ide tified fo eac ' waste t; pe,. ba d on lab ratory scal and/ full sc le sting or e; peri nce. Wit d taring, the PCP shall
.inci e n identif ion of conditio s that must be satisfied, ba ed on full ca e testing, to assure that watering of bead resins, powde' red resins nd filter sludges will result in volumes of free water, at the {
time of disposal, within the limits of 10 CFR Part 61 and of the low-level radioactive waste disposal site. PURGE - PURGING 1.31 PURGE or PURGING shall be 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. LIMERICK - UNIT 2 1-5 AUS 2 51919 .
4 DEFINITIONS -
-REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)
- b.- - All ~ refueling floor secondary' containment hatches and blowout. panels are closed and sealed.
- c. -The standby gas treatment system-is in compliance with the requirements of Specification 3.6.5.3.
- d. At least one door in each access to the refueling floor secondary 'I containment is closed. ,
- e. .The sealing mechanism associated with each refueling floor secondary -
containment penetration, e.g. , welds, bellows, or 0-rings, is.0PERABLE.
- f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.
REPORTABLE EVENT 1.36AREPORTABLEEVENTshallbeanyofthoseconditionsspecifi; edin Section 50.73 to 10 CFR Part 50. . ROD DENS!TY 1.37 R00 DENSITY shall be the number of control rod notches inserted as a l fcaction of the total number of control rod notches. All rods fully interted is equival,ent to 100% ROD DENSITY. SHUTDOWN AARGIN 1.38 SHUTD0h!N MARGIN shall be the amount'of reactivity by which the reactor is
=
subcritical or would be suberitical' assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is_in the shutdown condition; cold, i.e. 68'F; and xenon free. SITE BOUNDARY 1.39 The SITE BOUNDARY shall be that line as defined in Figure 5.'1.3-la. SOLIDIFIChTION - g
.40 OLID FICATION shall b th .immobi z Lion of w radioactive wastes such i I
as e por or bostoms sper t res' s, sludges nd everse s ) sis c ce - , trates a result a prc ces of tho oug mixirg th wasto ty with a i i solic cation a t(s) to - a free st ding mor 1 h with mical and physical characteristics specified in ; e PROCESS C TROL PROGRAM (PCP). l SOURCE CHECK-1.41 A SOURCE CHECK shall be the qualitative assessment of channel response
. when the channel sensor is exposed to a radioactive source.
A 4 LIMERICK - UNIT 2 1-7 AUS 2 5 HM 4
1 . 1 i l
-INSTMENTATION , _
)
METEOR OGICAL MONITORING INSTRUMEN TION- l l LIMITING C DITION FOR OPERATION s
\ 1 N
3.3.7.3 The me orological monitoring _ ins umentation channels shown in Table 3.3.7.3-1 shall b OPEiMBLE. APPLICABILITY: At I times. ACTION:
- a. With one or mor meteorological monitori instrumentation channels inoperable for m e than 7 days, prepare a d submit a Special Report to the Commission ursuant to Specification .9.2 within the next 10 days outlining t cause of the malfuncti and.the' plans for restoring the instru ntation to OPERABLE stat .
- b. he provisions of Speci ication 3.0.3 are not app i.ctb h.
SURVEILLANCE REQUI ENTS s 4.3.7.3 Each of the above'vequired meteorologi 1 monitoring instrumentation channels shall be demonstrated OPERABLE by the pe formance of the CHANNEL CHECK and CHANNEL CALIBRATION operatio,ns at the frequenc s shown in Table 4.3.7.3-1. I l . i . e B 1 - 9 LIMERICK - UNIT 2 3/4 3-73 M 2* IH'; w - . . - - _ _ . - - _ _ . - - - - - - -- - - . - - - - .- - - - -
y. f. TABLE 3. 3. 7. 3-1 : [T L u . .
~
METEOROLOGICAL MONITORINEkINSTRUMENTATION ua , N " MINIMUM Tower 1 T wer 2 ~ INSTRUMENTS INSTRUMENT . (Primary)- (Ba kup) OPERABLE
- 1. ind Speed f
- a. Elevation 1 30 feet or 159 feet 1 i
- b. Elevation 2. 175 feet or' 304 feet' 1
- 2. Wind irection -
\
- a. El vation 1 30 feet or 159 feet ,
1 E
- b. Elev tion 2 175 feet or 304 feet 1 ,
s ! 3. Air Tempera e Difference 'a-
- a. Elevationg 300 feet-
, 266feethor
- 26. feet 26-feet 1 u
. N
/ . \
\
s, i I i t 4 et e4 g e LIMERICK - UNIT 2 3/4 3 74 AL'S 2 5 ,tih, 94
t TABLE 4.3.7.3-1
^
METEOROLOGICAL MONITORING INSTRUM_ENTATION SURVEILLANCE REQUIREMENTS
)
INSTRUMENT-
\ '
CHANNEL CHECK CHANNEL CALIBRATION
- 1. Wind Speed -
Elevatio\ n1 Tower. l'and Towe 2) D SA
- b. Elevation 2 ( pwer 1 and Tower ) D \\ SA _
Wind irection
- 2. .
- a. El vation 1 (Towe 1 and Tower 2) D \ SA
- b. Elevhtion 2 (Tower and Tower 2) D SA
- 3. Air Temper re Difference \
- a. Elevation 266 - 26 ft ower1) SA ~
- b. Elevations'300 - 26 ft (To r 2) D SA s
\'
\
s g s [ - s l' L . LIMERICK - UNIT 2 3/4 3-75 A'J S 2 5 !!33 l \ -
't I
i INSTRUMENTATION: , RADI0 ACTIVE LIQUID' EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITIONJOR OPERATION \ ' l
'3.3.7.11 The radios ive liquid efflu\ent monitoring instrume ation channels t
l
- shown in Table.3.3.7. 1-1 shall be OPERABLE with their alare/t ip setpoints l set to ensure that the exceeded. The alahtripsetpoints*ilimitsofSpecification3.11.1.1arenogf these channels 'shall be accordance with the met odology and parameters in the OFFSITE D E CALCULATION I MANUA,L (ODCM).
\ \\ -
_APPLICABILIITY: - At all times. \ ACTION: a.\Witharadioactiv liquid effluent monitoring instrumenta ion , schannel alarm / trip setpoint less con'servative than required by the
'above specification immediately sus nd the release o k radioactive liquideffluentsmontoredbytheaf@fectedchannel,ordecirethe channel inoperable. l
\
- b. With less than the min mum number of ra ioactive liquid effl ent !
monitoring instrumentation channels OPER BLE, take the ACTION \shown
. in Tahle 3.3.7.11-1. Regtore the inopera le instrumentation to OPERABLEstatuswithinthytimespecifiedintheACTIONorexplain in the'next Semiannual Radioactive Effluent Release Report why this inoperabilitywasnotcorrhtedwithinthetimespecified.
- c. The provisions of Specificat'on 3.0.3 are not applicable.
, SURVEllLANCE REQUIREMERTS
\- x '
I
\ \ \:
4.3.7.11\Each radioacti e-liquid effluent gonitoring instrumentation ch shall be emonstrated OPERABLE by performance of the CHANNEL' CHECK, SOURCE ; i CHECK,CHA{sshowninTableNEL frequencie s CALIBRATION \and CHANNEL FUNCTIONAL TEST op 4.3.7.11-1. ,
\ -
, \ \
l: N) L " Excluding-the flow rat measuring devices which are not determined and adjusted ! in accordance with the 00CM. i l LIMERICK - UNIT 2 3/4 3-98 E 2 5 till J
TABLE 3.3.7.11-1 . RADI0 ACTIVE-LISUID EFFLUENT MONITORING STRUMENTATIbN l
\ \ '
l i MINIMUM i CHANNELS INSTRUMENT- OPERASLE. ACTION
- 1. - GR S RADI0 ACTIVITY MONITORS OVIDING s 1
. AU' TIC TERMINATION OF RELE E ,
- a. LTquid Radwaste Effluent Li 1 \ 100 .
I s -
- b. RHR' Service Water System Efflu nt Line 1/ loop 101
\
. . GROSS RADIOACTIVITY MONITORS NOT l PROVIDING AUTOMATIC TERMINATION .
OF RELEASE x
. Service Water System Effluent Line 1 101 Ns ..
s
- 3. FLOW RATE MEASUREMENT DEVICES- ,
\ ,
- a. Liquid Radwaste Effluent Line 1 102
'\; x.
N
- b. Dis' charge Line. 'Ns 1' 102 -
\
\.
N 1
. LIMERICK - UNIT 2 3/4 3-99 g,JG 2 51985
- 3
. i b ,
y . [ TABLE 3.3.7.11-1 (Continued) h '
'ACTIO% STATEMENTS
-ACTION 100 - With the numbei of chann s OPERABLE less than req ired by the -
Minimum Channels OPERABLE equirement, effluent rel ases may conti ue for up to 14 days rovided that prior to in'tiating a rele se:
- a. At east two independent mples are analyzed in accordance l wit Speci,fication 4.11.1. 1, and \
- b. At le t two technically qua 'fied members of the fac'ility - I staff dependently verify the release rate calculations and dis arge line valving; therwise, sus nd release of radioact've effluents via this thway. '
t A 10h'101 - Wit the number of channels OPERACLE less than Yequired by the' Mini um Channels OP ABLE requirement, eff vent releases via ' this thway may con nueforupto30dsysbrovidedthat,at least ce per 8 hours grab samples are coli cted and analyzed for gros radioactivity (bota or gamma). Beta is analyzed.at a limit o detection of least IN7 microcurie mL. Gamma is ' analyzed a a limit of det etion of at least SN7 crocurie/mL. ACTION 102 - With the numb of channels ERABLE less than regttired by- the Minimum Channel OPERABLErequhement,effluentreleasesvia his pathway may ontinue for upNto 30 days provided the flow r ta is estimated a(least once per 4 hours during actual re eases. Pump curves generated in situ may be used to estimate flow, t L l
- l. -t l
1 l' i l
. LIMERICK - UNIT 2 3/4 3-100 E25E
m. b y TABLE 4.3.7.11-1 E g RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUNENTATION SURVEILLANCE REQUIREENTS M - n " CHAfMIEL c
.j INSTRUNENT-CHANNEL
-CHECK [
SOURCE CHECK CHANNEL
-CALIBRATION FUNCTIONAL TEST n !
- 1. GR035 RADI0 ACTIVITY PWNITORS PROVIDING
-AUTOMATIC TERMINATION OF RELEASE -
/
, /
- a. Liquid Radvaste Effluent Line
/ /
, [ P. P R(3) Q(1)
- b. RHR Service Water System Effluent. tine D M R(3) Q(1)
~
- 2. bROSSRADI0ACTIVITYMONITORSNOTPROVIDING -',-
AUTOMATIC TERMINATION OF RELEASE
- a. Service Water System Effluent Line. ' M D ,
R(3) Q(2)
- i to / .
$ 3. FLOW RATE MEASUREMENT DEVICES / ~
w ,
- a. Liquid Radwaste Effluent Line / D(4) N.A. R Q ,
- b. Discharge Line D(4) N.A. R
~~
- V' .
z
/
.5 tu y - r O
w 7 - . . - ~ _ .-. _ . _ . , _ _ _.__s . , _ , . , . ,,. , , *, , '
TABLE 4.3.7.11-1 (Continued) TABLE NOTATIONS (1) -The CHANNEL FUNCTIONAL TEST shall als demonstrate that au matic isolation of this pathway and coptrol room ala annunciation occur i any oi'the' following conditions ek sts:
- 1. In trument indicates easured level above the alarm / trip setpoint.
- 2. Cir uit f2ilure. ;
- 3. Inst ument indicates a downscale fail re.
- 2) The CHANNEL FUNCTIONAL TEST sh 11 also demon trate that control room alarm i annunciation occurs if any of t following nditions exists:
- 1. Instrumen indicates measure levels abov the alarm setpoint. 7 Circuitfar(ure.
- 3. Instrument i ' :'
- icatesadownscal(failure. s s 1 (3) TheijtialCHANNEL ALIBRATION shall.be performed;u' sing one or more of the reference standaris certified by the National Bureau of Standards (NBS) orusing\standardsthakhavebeenobtainedfromsuppliersthatparticipate ,
in measur entassuranegactivitieswithNBS. These \ standards shall permit calibratin the' system ov r its intended range of eneigy and measurement range. For ubsequent CHA EL CALIBRATION, sources that have been related to the initia calibration all be used. (4) CHANNEL CHECK s 11 consist.of verifying indica' tion of flow during periods of release. CHA ELNCHECK shall\be made at least once per 24 hours on days on which continuou Vperiodic, or' batch releasesigre made. LIMERICK - UNIT 2 ' 3/4 3-102 A% 2 5130 I I I e
INSTRUMENTATION . L% PJ.f:hqvG-
., 0;.C.' A
'E GASE4E4Mt4EeT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION e le5.n 3.3.7.12 The re.g;;;t4ve gas:::: ;ffi;;r,t monitoring instrumentation channels shown in Table 3.3.7.12 1 shall be OPERABLE with th ir alarm / trip setpoints set to ensure that the limits of Specification 3.11.2 3 are not exceeded. W chr;;,/ trip ;etpeirt;* ef G,. rpliceble :"aa=1= thall be d;;;r;in;d 'a 2 re"d'^"e with the am+ hadal anu and naramataes 4a th: 00C". .
APPLICABILITY: As shown in Table 3.3.7.12-1 , ACTION: , en ay lasm Withar:d.g::tivegaster: :ffi;;r,t monitoring instrumentation
- a. .
channel alarm / trip setpoint less conservative than required by the above Specification, 4 n di iel.7 :::: nd th: r:1== e ef r: directive g:::::: Of'1;;nt: :: nit:::d by the eff;;t:d thina=1 ae det e-e + ht t
- h;rnei in;p;r: tie, aod %ks the Aefib yhW.n T@l* 4.4c7.r2 -t
- b. Withlessthantheminimumnumberofrbee'; e gas::;; :fflu:nt monitoring instrunantation channels OPERABLE, take the ACTION shown in Table 3.3.7:12-1.- Restore the inoperable instrumentation to OPERABLE status within t tracif4ad 4- the ^.CTIOF er ;;piain w)pf this inonerab414+v/m& er -et tir:
- 77::t:d' in ; tj,;;;jj ,,,.unwi 16 the next 5:rfran"m1 Radin*k+i'!e Effivuni Rwiwanu R.yv.t.
% todap und ; c~ ansxcen%l, QMjurs aN .S~$ 9 i c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable,
+.p(
s
- n. g SURVEILLANCE REQUIREMENTS -
(uN4 at h 1 S P - : G V. a 4.3.7.12 e O Each r;gf.eectia gastee ef'!uent monitoring instrumentation channel M.7. an shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, WW6E-- O" 'jf4(i,f AttfCK~~tHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the *5 n*ner.4.) I frequencies shown in Table 4.3.7.12-1. ,
'en a he t,
, menner) i "The alarm / trip setpoint4 for the Main Condenser Offgas Treatment Systea L Explosive Gas Monigoring System : d the ",eir, Cea .n;;r ^ff;;:: " rete + = at
! ";di; tier "; nit:r'_re set in accordance with Specification 3.11.2.5 and 1 3.11.2.6, re:pecti';:1y. 1 LIMERICK - UNIT 2 , 3/4 3-103 t.112 51989
(1 l l . TABLE 3.3.7.12-1~ C - (Er/[ww 6 , jF,
- aABIGAGHVE GASEGUS-EFF4UDIT MONITORING IHSTRUNENTATION
~
E
. MINIMUM CHANNELS INSTRUMENT OPERABLE' APPLICABILITY . ACTION . ,e 5
- 1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM
" EXPLOSIVE GAS MONITORING SYSTEM
** 110
- a. Hydrogen Monitor 1 .
l
- 2. SOUTH STACK EFFLUENT -
MONITORigSYSTEM
- 111' j . Noble G i Monitor 1
- l
- b. Iodine Sampler 1 112 l
l R
- 112 l Y c. Particulate Sampler 1 M .
- 113
- d. Effluent System Flow Rate Monitor 1 .
- 113
- e. Sampler Flow Rate Moi itor 1 3. NORTH TACK EFFLUENT MONIT RING SYST
- a. Nobl Gas Activity Monit 1 -
- 114' l- ~
,
- 112
- b. Iodine ampler 1 5 * * - 12 -
l c. Partipula e Sampler , 1 -
- 113 l d. Effluent Sy em Flow Rate Monitor 1 l
- 113
- e. Sampler Flow Rate @ nitor 1 M,
l n .
<.n l h m
I
~
- l .
s- 9 a- .-. ~ . . ,._z
. . _ -- _ , __, - _ _ _ _ _ . . . _ _ _ . . . __ _ _ _ _ _ _ _ _ _ _ _ . . _________m__z._____ _ _ , __
~i
, TABLE 3.3.7.12-1 ~.(Continued) l
. TABLE NOTATIONS )
44MalLt4eas.
**0uring operation of the main condenser steam jet air ejector and offgas treatment system.
***During operation of the hot maint'enance shop ventilation exhaust system.
ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser -- i offgas treatment system may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours, j ACT ON 111 - With the number of channels OPERABLE less than required by the I i Minim Channels OPERABLE e trement, effluent releases via this pa way may continue or up to 30 days provi,ded grab
- samples re taken at least on per 8 hours an these samples l
'are analy ed for gross activit/ within 24' hours. l ACTION 1 2 - With the n ber of channels OPIRABLE less than requir by the .
Minimum Chan als OPERABLE requ ement, effluent releas via this pathway may c ntinue for up to O days provided samples are con-tinuously* coll cted with auxilia sampling equipment as required in Table 4.11. 1.2-1. ACTION 113- With the number of channels OPERABLE less than required the Minimum Channels OPERABLE requirement,' effluent releases ia this < pathway may continue for up to 30 days provided the flow te is
, estimated at least once per 4 hours, i ACTION 114 With the number f channels OPERABLE le is than required by e- .
Minimum Channels PERABLE requirement, offluent releases via his
., athway may conti for up to,30 days irovided grab samples a e t kan at least once p 8 hours and the e samples are analyzed fo gross activity withi 24 hours and p ovided the mechanical vac um pumps are not opersted.
ACTION 115'- With the number of'channe ls OPERABLE less than quired by the Mini um Channels OPERABLE requirement, releases t the environment may ntinue for up to 72 hours provided that the rth Stack Efflu t Noble Gas Activik Monitor is OPERABLE; ot rwise, be in at 1 st HOT SHUTDOWN within 12 hours. L LIMERICK - UNIT 2 3/4 3-106 m i 519H w > < , - - - ,
, . TABLE 4.3.7.12-1 y E YNQst .
g RA910AeffvE GASEG N MOKITORING INSTRUMENTATION SURVEILLANCE REQUIREE NTS
~
n
. CHAfGIEL PRIDES 'IN WHICH
.g CHANNEL- -s'MNNiG CHANNEL FUNCTIONAL SURVEILLANCE y INSTRUMENT CHECK MCit- - CALIBRATION TEST IS REQUIRED
- 1. . MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING ;
i SYSTEM
. . (g
- a. Hydrogen Monitor D R Q(2)) M
- 2. N SOUTH STACK EEFLUENT NITORING SYS M ,
R, a . Noble Gas Acti .ity Monitor D M . R(2) Q(1) ,
.A.
- Y li. Iodine Sampler (4) M.A. N.A. ,
i E$
- j
" :. Particulate S ler (4) M.A. N.A. M.A -
- 1. Effluent System : low Rate Menitor D N.A. R Q Sampler Flow Rati r Monitor D ,
M.A. Q.
- 3. NO STACK EFFLUE
~
,i MONI RING SYSTEM -
-
- i
- a. Nobl Gas Activity nitor D M R(2) Q(1) -
N.A. N.A. *
- b. Iodi Sampler W (4) M.A.
~
N.A. N.A. M.A. *
- c. Partic late Sampler W (4)
- d. Effluen System Flow Rate nitor D N. A.' R Q~
E
- e. Sampler F1 , Rate Monitor _
D N. % R. Q N.
~
9
, l TABLE 4.3.7.12-1 (Continued)
TABLE NOTATIONS n .-
** During operation of the main condenser steam jet air. ejector and offgas treatment system.
-*" Ovi iny p= tier Of '". het .iriter..r,;e ehep =ti htkr. . l,.unT, sysum.
(1) The CHANNE FUNCTIONAL TEST shall also demonstrate.th t control room alarm annunc tion occurs if any of the 110 wing cond ions exists: ,
- 1. Instrumen indicates me sured levels ove the ala / trip setp int.
- 2. Circuit fai re.
- 3. I strument in ' ates a dow cale failure. -
- 4. Ins ument controls not set in operate mode. '
(2) Th initial C'ANNEL CALIBRATION sha 1 be perfo ed using"one or more of the eference tandards rtified b the Nation 1 Bureau f Standards (NBS) or u ng stand ds that h e been obt ined from uppliers hat partici te in mea urement a surance ac ivities w h NBS. Th e standa ds shall pe it l calibra ing the s st.em over ts intende range of ergy an measurement range. or subseg nt CHANNE CALIBRATI , sources at have been related to the in ial cali ' tion sha be used. - il , (% The CHANNEL CALIBRATION shall include the use of standard containing a nominal: lo ! 1. f#volumepercenthydrogen,balancenitrogen,and i
- 2. 4 volume percent hydrogen, balance nitrogen, e
l 1 LIMERICK - UN.IT 2 , 3/4 3-109
=
e' w - e -- _+y- - -
, r 3/4.11 RADI0ANIVE EFFLUENTS -
m x s 3/4.11.1 LIQUID \ EFFLUENTS
- \ ' .
CONCENTRATION t
\
LIMITING CONDITION FOR OPERATION 3.1 .1.1 The concentration of radioactive material. released in liquid effluents to RESTRICTED AREAS (s'ee Figure 5.1.3-1) shall be' limited to the concentrations spec led in 10 CFR Part 10, Appendix B, Table II, Column 2 for radionuclides - other than dissolved or estrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcuries/ml total activit . \ i APPLICABI ITY: Atalltimes.\ ACTION: .- With the con ntration of radioactive material released in liquid effluents to UNRESTRICTED REAS exceeding the above limits, immediately restore the concen-tration to wit in the above limits. \ \ \ SURVEILLANCE REQ IREMENT'S \
\
-4.11.1.1.1 Radioa\ctive liquid wastes shal'1 be sampled and to the sampling and alysis program of Table 4.11.1.1.1-1.
\
4.11.1.1.2 The result . of the radioactivity analyses. shall be used in s accordance with the methodology and parameters in the ODCM to assure that'the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. \ s
\
\
4 e
= .
i a M LIMERICK - UNIT 2 3/4 11-1
. . - . - - . .. _=. _ _ . - . _ _ . . -
, l TABLE 4.11.1.l.1-1 .
RADIDACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGR ,
\ \
g - 0 L'WER LIMIT MINIMUM TYPE OF 0F,DETEC{ ION- ! SAMPLING ANALYSIS ACTIVITY \(LLD) i LIQUID RELEASE FREQUENCY TYPE
\ FREQUENCY ANALYSIS (pci/mt) )
A. Bat'h c Waste P P Principa} Gamma Sh107 v Release. Each Batch -l Each Batch Emitters i b l Tanks s
] \ j I-131 )
1xik'
- 1. Floor Drain P M Diss'olved and-Sample Tank One Batch /M Entrained Gases '. 1x10-J.
No. 2 , (Gamma Emitters) ' 1
. 2. Laundry P M H-3 1x10 5 rain Sample d Each Batch Corrposite Gross Alpha ,
1x10 7
) Q q Sr-89, Sr'90 5x10.s
.Each Batch s Composite .
Fe-55 \. lx10 8 B. W Continuogs W Principa} Gam ma \ 5x10 7 Release Grab Sample Emitters N i , I-131 \* 1x10 8
- 1. RHR Service 1 W W Dissolved and 1x10 5 Water System f Grab Sample Entrained Gases i Effluent Line \
(Gamma Emitters) \
- 2. Service Water W M H 1x10 5
'q System Grab' Sample Composite Effluent
. Gross Alpha 1x10 7 Line i
'W Q Sr-89, Sr-90 5x10 s Grab\Sample Composite d Fe-55 1x10 8 LIMERICK - UNIT 2 3/4 11-2 E. 2 5 %
, TABLE 4.12.l.1.11.(Continued)
TABLE NOTATIONS a ihe LLD is defin d, for purposes of tO e specifications as the smallest concentration of' adioactive materia 1 i asamplethatwIllyiel a net count, , 1 above system back round, that will be{td ;ected with 95% probabili with only 5% probability of aisely concluding tha< a blank observation repr sents a "real" signal, , For a particula measurement system, w ich may include radioche ical separation: 4 66s b \ 1 LLD =
\ E-V- .22 x 108 Y exp ( 14t) "
Whe :
\ -\ c, -
LLD is the a priori 1 er limit of detection,*as defined above (as \ microcuries per unit a ss or volume), \ .
\
s cNu\is the standard deviation of the background' counting rate or of the nting rate of a blank ample as appropriate (ps counts per minute),1 E is the counting efficiency, as counts per disiritegration, V is the sample size, in un(ts of mass or volume, \ 2.22 x 105 is the number of sintegrations per min e per microcurie,- i Y is the fr'actional radiochemihal yield, when applicable, x
. AistheradIactivedecayconsUantfortheparticularradionuclide,and at for the plant effluents is the elapsed time between the midpoint of sample collection and time of counting.
Typical values of E, V, Y, and at should h used in the calculation. It should be recognized that the'LLO is defined as an a priori (before the fact) limit representing the capability of a measuremeiit system and not as an a posteriori (after the fact) limit for a particular measurement. 4 9 LIMERICK - UNIT 2 3/4 11-3 e __,m-_ , . - , _ _ , - - - , - - - - - * " - ' '
4
.t TABLE 4.11.1.1.1-1.(Continu d)
TABLE NOTATIONS b A batch re ese is the discha e of liquid wastes of discrete volume Prior to s ling for analyses each batch shall be is lated, and then thoroughly a ed to assure repr sentative sampling, he principal a emitters for hich the LLD specifica ion applies inel e e following r ionuclides: Mn- , Fe-59, Co-58, Co-60, Zn-65 Mo 99, Cs 134, C -137, Ce-141, d Co-144. This ist does not mean that nly these nuclidhs ., at to be conside d. Other gamma aks that are identifi le,togetherwit
.tho e of the above nuclides, shall a so be analyzed and repbrted in the Semi- ,
annu 1 Radioactive ffluent Release R port pursuant to speci cation 6.9.1.8. d - 4 g,,, ite sample is ne in which the uantity of liquid sampled is propot-tional the quantity' f liquid waste discharged and in which.4he method of sampling mployed resul in a specimen that is representative 4f the liquids released. \ .-
\ '
'A continuous' release is the ischarge of liq)id wastes of a nondiscrete volume,e.g.,Tromavolumeofsa system that has an input flow during the continuous release. '
N
~
,Whenever effluen\ '
t releases are in excess of the mo Stor's setpoint.
. \ \.
Na i s . L 4 6 4 LIMERICK - UNIT 2 3/4 11-4 . 25252
l l RADI0 ACTIVE EFFLUENTS . DOSE LIMITING CONDITION OR OPERATION h3.1.1.2
\
The dose or ose comr.itment to
\ \
MEMBER OF THEi PU LIC from radioactive matQialsinliquidofluentsreleasedfr.mthesitetoUNRETRICTEDAREAS (See figure 5.1.3-1) sh 11 be limited: , at During any calendar quarter to legs than or equal t 3 mrems to - the total body and to less than og equal to 10 mrems to any organ, and
- b. During any calendar year to less tbn or equal to 6_ m ems to the s totalbodyandtoJessthanorequa to 20 mrems to any organ.
APPLICABILITY: Atalltimes.\ \ . ACTION:
\ \ .
- a. With the calculated dos s \e from the serelea\
of radioactive terials inliquideffluentsexceedinganyofthetabovelimits,probareand submit to the Commission within 30 days,' pursuant to Specif ca-tion 6.9;2,a,$pecialRepohtwhichidentifsiesthecause(s) r exceeding'the limit (s) and defines the corhective actions th t have beentaken'toreducethere1asesandthephoposedcorrectiveiactions 9 to be taken'to assure that subsequent releases will be in compliance with the above limits. ThisSpecialReportshallalsoincludejhe radiological impact on finishey drinking water, supplies at the s nearest downstream drinking wat,er source. ' i
- b. TheprovisionsofSpecification\ .0.3arenotahplicable. '
i SURVEILLANCE REQUIREMENTS - \
\ g 4.11.1.2 Cumulative dose contributions from li id effluents f 9r the current \ -
calendar. quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM a eastoncepe(31 days.,
- e LIMERICK - UNIT 2 3/4 11-5 e
'm a' ' - - -
sa. .-.........s- .- .. - - . - . . . . . . . . . - . . - - . - . - . . . . . - . . . . . - - . , . - . . - -
]
RADI0 ACTIVE EFFLUENTS - LIOUID RADWASTE TREATMENT SYSTEM I 4 LIMITING CONDITION FOR OPERATION
. 1 l
l 3.11.1.3 The liquid radwaste treatment system shall'be OPERABLE and appropri-ate portions of the syster shall be used to reduce the r dioactive materials in liquid waste prior to their discharge when the projected oses due to the liquid effluent, from t e site, to UNRESTRICTED AREAS (see Figu 5.1.3-1) would exceed , 0.06 mrom to the total body or 0.2 m' m to any organ in a 31-day period. ' APPLICABILITY: A all times. g ICTION: \.
\ I
- a. With radio etive liquid waste eingdischargedwit$outtreatmentand in excess o the above limits, prepare and submit to the Commission with.in 30 d s pursuant to Specification 6.9.2.a Special Report
- which includes the following information:
- 1. Explanat\onofwhyliquidrdwastewas'beingdischargedwithout .
trea,tmentg identification of'any inoperable equipment or sub-systems, nd the reason for the inoperability,
- 2. Action (s)'akentorestorethe,inoperableequipmenttoOPERABLE status, and g.
- 3. Summary descfiption of action (s) taken to prevent a recurrence,
- b. e provisions of Specification 3.0.3 are not applicable.
. \
SURVEILLANCE, REQUIREMENTS
\
4.11.1.3.1 Doses due to liquid releases from the site to UNRESTRICTED AREAS shall be projected at least onceiper 31 days in accordance with the methodologyandparametersinthe,0DCM.
\
4.11.1.3.2 The'itquidradwastetrea,tmentsystemshallbedemonstrated
. OPERABLE by meeting,Spe:ifications 3.11.1.1 and 3.11.1.2..
e .
\
LIMERICK - UNIT 2 , 3/4 11-6
RADIDACTIVE EFFLUENTS .
.. [ ~
LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside temporary tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICABILITY: At all times. ' ACTION:
- a. With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all. additions of radioactive material to the tank and within 48 hours reduce the ,
tank contents to within the limit and describe the' events leading to this condition in the next Semiannual Radioactive Effluent Release Report.
- b. The provisions of Specification 3.0.3 are not applicable. ;
SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a repre-sentative sample of the tank's contents at least once per 7 days when radio-active materials are being added to the tank. 1 l .- l l t . i LIKFRICK - U, NIT 2 3/4 11-7 AUS 2 5 ISH
4 RADI0 ACTIVE EFFLUENTS , 3/4.11.2 GASEOUS EFFLUENTS , DOSE RATE (LIMITING CONDITIOR FOR OPERATION ! li 1 \ i
\
- 3. 1.2.1' The do'se te due to radioa\ ctive materials released in of luents from the s te to areas at or\beyond the SITE BOUNDARY (gaseous see Fig re 5.1.3-1) shall be limited to th following:
~
. For noble ga es: Less than o equal to 500 mrems/yr o the total body and les than or equal to 3000 arems/yr to the s .in, and
\
- b. For iodine-131 for iodine 133, for tritium, and for all radio-nuclides in pa ticulate form with half-lives greater than 8 days; Less thanonly.)
athways or eq al to 1500 arems/{r to any organ. .(Inha ation . APPLICABILI Y: At all time *. \
. ACTION:
\
\
- a. With the dose rate (s exceeding the a ve limits, immediate1 restore the r lease rate to w thin the above 1 its.
k
- b. The pro isions of Speci ication 3.0.3 ar not applicable. \
SURVEILLANCE RE0VIR ENTS \ \
\ i 4.11.2.1.1 The dose r te due'to noble asesingaseous\effluentsshallbe \
determined to be within the above limit in accordance with the methodology \ and parameters of the 00 M. \\ 4.11.2.1.2 The dose rate uo to iodine-131, iodine-133, tr tium, and all ' radionuclides in particula form with halfylives greater th 8 days in gaseous effluents shall be ternrined to be within the above jmitsin accordance with the methodol and paramete h of the ODCM by obtaining representative samples and pegforming analyse 6 in accordance wit the sampling and analysis program specified\in Table 4.11.2hl.2-1.
\s )
x % 9 LIMERICK - UN.IT 2 3/4 11-8 nG 2 t d _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _, - _ e. -
l i 1
,._ TABLE 4.11.2.1.2-1 E
Q 10 ACTIVE GASEOUS WRSTE SAfFL.I AfG AN' ALYSIS PROGRAM M -f 3
/
h MINURM LOWER LIMIT OF 5 SAMPLING ANALYSIS TYPE OF DETECTION (LLD)* GASEOUS RELE TYPE FilEQUENCY 'TREQUENCY ACTIVITY ANALYSIS (pCi/el) -
]
, A. - Con [tainmentPurge P Each Purge '
- P Each Purge Principal G
/
EtiittersI 1x10 4 l
/ (Pretreatment) Grab ,/ !
, Sagpie
- Mb gb . . /'
~
- 8. North Stack and p-[ncipalGammaEmitters' 1x10-p ,
/ South Stack Grab -
l Y Sample f
/ H-3 7 x08 1 C. Hot Maintenance Continuous d
[' I-131
/
1x10.sr Shop Ventilation Charcoal I-133 1x10 88 ! / **EI' Exhausth and All I i
^
Release Types Listed in 8 Continuousd ' E Princip'al Gamma Emitters' 1x10 88 above - Particulate '(I'131,Others) j
- / ., Sample y l d '
Continuous Q f Gross Alpha 1x10 88 *!
# Composite Par-ticuiate Sample
! Continuous d
p
/
Q Sr-89/ Sr-90 x10 88
- Composite Par-ticulate Sample f y ;
/ d w 0. All Release Types Continuous Noble Gas Noble. Gases 1x10
- 5 Listed in 8 above I Monitor Gross Beta o (Based on
", and the Main ,
Xe-133)
..- Condenser Off Gas ,
3 Pretreatment Radio- ' activity Nonitor
/ _
) /'
, - -+. . - - -._ ____ _ __._ _ ___ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
l TABLE 4.11.2.1.2 1.(Continued) TABLE NOTATIONS }
\
a The LLD is defined, for purposes o these specificatio'ns as the smallest concentration of radioactive materi 1 in a sample that w 11 yield a not count, i abovesystem\ background,thatwill e detected with 954 p obability with only 5% probability of falsely concludin that a blank observa ion represents a
, "real" signal. 4
\For a parti ular measurement sys om, (which may include radiochemical i eparation):\ *
\ l
\
4.66s b LLD = k E V 2.22 x 108 Y exp (-Aat)
\
Where:
\:\
LLD isithe a priori lower limit of de ection as defined 'above,(as , microcuries per unit mass or volume), 0 i
, \
s b isthestandarddeviationofthebacgroundcounting'rateokofthe counting rate of a blank sample as appro riate (as counts per minute),
\
E is the cou'nting efficiency (as counts per disintegration), '
\
V is the sample size (in units of mass or v'olume),
\
2.22 x 108 is the number of disintegrations kr minute per'microcurke,
\
Y'is the fractional radiochemical yield (when ' applicable), \
\
A is. the radioactive decay constant for the particular radionuclide, and
., r \
At is'the elapsed time between midpoint of sample \ collection and time of counting (for plant effluents, not environmental shmples) \ The value of sb used in the calculation of the LLD for a detection'systek background counting' shall rate orbe based ofsthe counting on therate.of actualthe observed. blank samples variance (as of theg$ppropriate) rather' than on an unverified theoretically predicted variance. Typical values of E, V, Y, and at shall be used in the calculation.
\ '
\ r It should be recognized that the LLD is defined as an a pr' tori (before the fact) limit representing the capability of. a measuremelit sygtem and not as
, an 3 posteriori (after the fact) limit for a particular measurement.
N LIMERICK - UNIT 2 3/4 11-10 m,2 5 Hi3
.,d 3l 5l 4
TABLE 4.11.2.1.2-1,(Continued) TABLE NOTATIONS . J 5 Sampling and analys s shall also be performe following shutd n, startup, ) or a THERMAL POWER chapge exceeding 15% of th RATED THERMAL P ER within ' a 1-hour aeriod. This\ requirement does not a ly if (1) analy s shows j that, the )0SE EQUIVALEy 1-131 concentration i the primary coo ant has not increased more than a factor of 3; and (2) the pain c.ondenser o fgas prr ' treatmentradioactivityhonitorshowsthateff1bentactivityhasnot l increated more than a factor of 3. \ c Sample shall be changed a\ least once perY 7 days and analyses shall be i . complet within 48 hours a'f ter changing, or af te removal from sampler. Sampling \shall also be perf6rmed at least once per 24 hours for at \least
-l 7 da 15%ofysRATED following each shutdonn, startup, or THERMAL POWER change exteeding THERMAL POWER in\1 hour and analyses completed within 4$ hours of changing \ When samples ~ collected for 24 hours arh analyzed, the i corresponding LLDs ma This requirement
\ does not apply if (1)y concentration in be the incressed primary by acoolant factor of 10has analysis not \ s' increas hows that more,.than a
\ factor of 3; and (2) the noble gas, monitor shows that effluent activitys \ ,
has not increased more than a factor of 3.
\
i \
\ \ s
\
d Th'e ratio of the ' sample flow rate totthe sampled stream flow rate shall be known for the time period covered by each dose or dose ' rate calculatio\ ; made\inaccordance,withSpecifications\.11.2.1,3.11.2.2,and3.11.2.3. 3 n\ e The principal gamma emitters for which the LLDespecification' applies include \the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m and Xe-138 for gaseous emissions and Mn-54, Fe 59, Co-58, Co-60, Zni65, Mo-99, I-131, Cs-134, Cs-1371 Ce-141 and Ce-144 f'or par- - ticulate e'aissions. Thitslist does not mean that only these nuclides are to be consiWered. Other gamma peaks which a're identifiable, toge\her with those of the\above nuclides, shall also be anal Semiannual Ra' ioactive Effluent Release Report pursuant \yzed and reported ik the to Specifica-tion 6.9.1.8. s f Under the provis ons of footnote e. above, only noble gases need to be considered. \
\
g Deleted. N
\ g
- h. Required for the hot maintenance shop ventilation exhau t only during opera '
tion of the hot maintena,nce shop ventilation exhaust system.
\ s
\
\ \
\
LIMERICK - UNIT 2 3/4 11-11 E25W e
. . - - ---_e--~ ,,..--,.--1.-, - . , , . -- , . - . - . - . . - - - . . , - . , - - - - - , , - - - - .
RADIDACTIVE EFFLUENTS , D0SE - W BLE GASES LIMITING CONDITION FOR OPERATION \ l
.11.2.2 The air dose due to noble gases rele& sed in gaseous efflu nts from he site to' areas at and beyo d the SITE BOUNDA Y (see Figure 5.1.3- ) shall limited to the following: ,
- a. During any calendar gyarter: Less than or equal to 10 mrad for -
\' gamma radiation and less than or equal ko 20 mrads for beta radiation, and
- b. During any calendar yeart less than or e ual to 20 mrads for gamma radiation and less than o equal to 40 mr s for beta radiatio .
APPLICABILITY: At all times. ..- ACTION:
- a. With the calculated air dose om radioactive hobia gases in gaseous '
effluents exceeding any of the bove limits, phpare and submit to the Commission within 30 days, rsuant to Specification 6.9.2, a Spacial Report *which identifies e cause(s) for\ exceeding the limit (s) and defines the correcti actionsthathavebeentakento reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases 4 11 be in compliance with the above limits,
- b. The provisions of Specification 3.0.3 a not applica le.
SURVEILLANCE REQUIREMENTS \ 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
~
LIMERICK - U, NIT 2 , 3/4 11-12 a
i RADI0 ACTIVE EFFLUENTS - DOSE - I0 DINE- 1, 10 DINE-133. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM 1 LIMITING CONDIT QN FOR OPERATION I\ - 1
\3.'11.2.3 The dose \ OF THE' P 'BLIC from iodine-131, iodine-133, to a MEMBER tritium, and all radionuclides in parti late form with half- ives greater than 8 days in gaseoks effluents released from the site to at s at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shell be limited to the following:
\ -
- a. During any calendar quarter: Less than or equal to 1 arems to any organ and,
- b. ' During any calendar year: Le u t'han or equal to 30 mrees to any
\ organ.
APPLICABI'tITY: At all times. ACTION:
- a. With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-li ' s greater than 8 days, in gaseous effluents exceeding any of t e above limits, prepart and submit to the Commission within 30 days, .ursuant to Specification 6.9.2, a Special Report which identifies the'cause(s) for exceeding the limit and defines the corrective actions that,have been taken to reduce the releases and the proposed corrective a'ctions to be taken to assure that subsequent releases will be in compliance with the above limits. s b.
\
The provisions of Specification 3.0.3 are not applicable.
\ \
SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dos contributions fo the current calehdar quarter and currentcalendaryearfo(r; iodine-131, iodine-233, tritium,andradionuclides in particulate form with hvif-lives greater than 8 days shall be determined in accordance with the methodology and p'arameters in the ODCM at least once per . 31 days. , e 0 LIMERICK - URIT 2 3/4 11-13 m25W
RADIDACTIVE EFFLUENTS
\ ;
VERTILATION EXHAUST TREATMENT SYSTEM
\
LIMITING CONDITION FOR OPERATION
\
3.11.2.4 The VENTILATION E HAUST TREATMENT SYSTEM shall be OPERABL and
\ r
. appropriate portions of the system shall be used ito reduce radioact ye materials in gaseolvs waste prior to their discharge when the projected doses e to gaseous effluent releases from the site to areas at and beyond the S TE -
BOUNDARY (see Figure 5.1.3-1) when averaged over 31 days would excee 0.6 mrem - to any organ in a 31-day period. PPLICABILITY: At all times. . kTION:
- a. With gaseous waste being discharged without treatment .and in excess of the above limits, prepare and submit to the Commission within 30 day ~s, pursuant to Specification 6.9.2, a Spei:ial Report which includes the following information:
- 1. Explanation of why gaseous radweste was b'eing discharged without treatment, identification of any inoperable equipment or subsystem %, and the reason for the inoperability, t 2. Action (s)'taken to restore the inoperable equipment to OPERABLE
\ status, and
\ 1 3g Summary description of action (s) taken to prevent a recurrence.\
- b. ThA provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site to areas a't and beyond theSITEBOUNDARYshallbeprojectedatleastonceper31daysinaccordance with the methodology and parameters in the ODCM.
~
4.11.2.4.2 The VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by meeting Specifications 3'.11.2.1, 3.11.2.2, and 3.11.2.3. LIMERICK - UNIT 2 3/4 11-14 M25W
1 i I RADIDACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be limited.to less than or equal to 4% by volume. APPLICABILITY: Whenever the main condenser air ejector system is in operation. ACTION:
- 4. With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentr.atton to within the limit within 48 hours.
- b. The provisions of Specification 3.0.3 are not applicable.
- e SURVEILLANCE REQUIREMENTS
- 4.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydro 3.3.7.12-1 of Specifica-tion 3.3. gen 7.12.monitors re* quired OPERABLE by Table .
l es
, e LIMERICK - UNIT 2 3/4 11-15 AM 2 5 BB
. i 1
RADI0 ACTIVE EFFLUENTS MAIN CONDENSER , i LIMITING CONDITION FOR OPERATION l 3.11.2.6 The rate of the sum of the activities of the noble gases Kr-85m, Kr-87 Kr-88, Xe-133, Xe-135, and Xe-138 measured at the recombiner af ter-condenser discharge shall be limite(. to less than or equal to 330 mil 11 curies /second. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3*. ,, ACTION: With the rate of the sum of the activities of the specifie' noble d gases at the recombiner after-condenser discharge exceeding 330 mil 11 curies /second, l- restore the gross radioactivity rete to within its limit within 72 hours or be ' in at least HOT SHUTDOWN within the next 12 hours. . '. g SURVEILLANCE REQUIREMENTS 4.11.2.6.1 The rate of ,the sum'of the activities of noble gases at the recombiner after-condenser discharge shall be continously monitored in accor-dance with Specification 3.3.7.12. 4 4.11.2.6.2 The rate of the sum of the activities of the specified noble gases from the recombiner after-condenser discharge shall be determined to be within the limits of Specification 3.11.2.6 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the recembiner after condenser discharge:
- a. At least once per 31 days,
- b. .Within 4 hours following an increase, as indicated by the Main Condenser Off-Gas Pretreatment Radioactivity Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level or air in-leakage, in the nominal steady-state fission gas release from the primary coolant. ..
- c. The provisons of Specification 4.0.4 are not applicable.
*When the main condenser air ejector is in operation.
LIMERICK - UNIT 2 3/4 11-16 AM 2 51959 4
. - - - ~ - , - - , m - -
ILADI0 ACTIVE EFFLUENTS VENTING OR PURGING - LIMITING CONDITION FOR OPERATION
\
3.11.2.7 VENTING or PURGING of the Mark II containment shall be through the \ standby gas treatment system. ' APPL!bBILIH: Wh hever the containment is y nted or purged." s ACTION:
\ \
, s. Withthereqbirementsoftheabovespecificationnotstisfied,
\ suspend all Vf TING and PURGING of the ontainment.
- b. T provisions of Specification 3.0.3 at not applicable..
\ \ -
\
SURVEILLANCE REQUIREMENTS s i
\ \
\.11.2.7.1 The containment shall;be determined to be aligned for VENTING or RGING through the standby gas treatment system within\4 hours prior to s art of and at least once per 12 hours during VENTING or PURGING of the co tainment.
4.1 2.7.2 Prior to use of the purge' system through the standby gas treatment syst ' assure that: g i a Both standby as treatment system trains are OPERA E whenever the purge system s in use, and s
- b. s Whenever the purge system is in use during OPE. RATIONAL CONDITION 1
'or 2 or 3, only one of the standby gas treatment system trains may be used.
i \ k
\ .
\ :
s s .
*Except for the one inch /two inch vent valves to the Reactor Enclosure '
Equipment Compartment Exhaust Filters when used for containment pressure control and nitrogen.make-up operations. e g LIMERICK - UNIT 2 3/4.11-17 gg 5 U0
1 l RADI0 ACTIVE EFFLUENTS
/4.11.3 SOLID RADWASTE TREATMENT ,
L ITING NOITION FOR OPERAT N 1 1 i 3.11 3 Radibactive wastes shal be SOLIDIFIED or dew ered in accordance w h the OCESS CONTROL PROGRAM to et shipping and trans ortation requirements durin transit, and disposal sit requirements when re ived at the disposal site. s g . ( \i APPLICA ILITY: At all times. ACTION: -; ,
- a. ith SOLIDIFICATION or dewatering not meeting df'sposal site and ,
s hipping and transportation' requirements, suspend shipment of the i adequately processed wastes and correct the PROCESS ~ CONTROL PROGRAM, trecurrence.
, procedures and/or the so1{d waste system as necessary to prevent
\
- b. With the SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1h test the ig roperly processed waste in each container to ensure that it meets burial ground and shipping requirements recurrence. and\ (2) take appropriate' administrative \ action to prevent ,
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3.1 The PROCESS CONTROL PROGRAM shall be used to verify that the properties of the packeged waste meet the minimum stability requirements of 10 CFR Part 61 and other requirements fee transportation to the disposal site and receipt at the disposal site.
~
e 9 9 e LIMERICK - UNIT 2 3/4 11-18 td 2 5 W
4
, M ;
y RA010 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE .
'l,
]
LIMITING CON (h! TION FOR OPE TION ! t 3.11.4 The an\ nual (calendar year) dose or d se commitment to any MEMBER OF i HE PUBLIC, due to releases f radioactivity and radiation, f uranium fuel . cle sources shall be limite to less than r equal to 25 mreps to the total dy or an organ, except the! thyroid, which hall be limited to less than or e ual to 7 arem. \
\
APPLICABILITY: At all times. i 1 g . ACTION: i Sa. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the. limits of Specifi-3.11.1.2b., 3.11.2.2a., 3.11.2.2b. ' cation or 3.11.2.3.1.1.2a.lculations b., ca shall be made including ect radiJ. dIr.3.11.2.3a. tion , contributi s from the reactor units and from outside' storage tanks , to determi whether the above limits of Specification 3.11.4 have been exceed d. If such is the case, prepare an'd submitito the Commission thin 30 days pursuant to Specification 6.9.2, a Special Report that efines the co,rrective action to be taken to, reduce subsequent re' eases to prevent recurrence of exceeding the above limits and in ludes the schedule for achieving'conformance with the . above limits. This Special Report as defined in 10 CFR 20.405c, shall include nanalysisthatestImatestheradiationexposure (dose) to a ME ER OF THE PUBLIC from uranium fuel cycle sources. ' includingallefluentpathwaysanddirectradiation,for}hecalendar year that inclu s the release (s) covered by this report. It shall alsodescribelealsofradiationandconcentrationsofradioactive material involve and the cause of the exposure levels or tions. If the es imated dose (s) exceeds the above limits, goncentra- and if the release condit on resulting in violation of 40 CFR Part 90 has notalreadybeencrrected,theSpecialReportshallinclude}sa request for a varia ce in accordance with the provisions of 40 CFR Part 190. Submitta of the report is considered a timely request, and a variance is g nted until staff action on the request isi complete. \ \ -
- b. The provisions of Spe(ffication 3 0.3 are not applicable.
SUR EILLANCE REQUIREMENTS \ _-
\
\
4.11. 1 Cumulative dose contrib tions from liquid and gaseous effluents shall e determined in accordance ith Specifications 4.11.1.2, 4.11.2.2, and , 4.11.2. , and in accordance with t e methodology and parameters in the ODCM. 4.11.4.2 If the cumulative dose cor 3.11.4, A TION a. Cumulative dose co\tributions ntributions from exceed the limits direct radiation defined fromin unit oper ion shall be determined in accordance with the methodology and parameters in the ODCM. \ , LIMERICK - UNIT 2 3/4 11-20 AUG 2 5 2
RADIDACTIVE EFILUENTS S RVEILLANCE RE REMENTS (Continued) \ es s
\
4.\1N3.2 If the 5\0 IDIFICATION method is used, the PROCl S CONTROL PROGRAM t shaligbe used to verify the SOLIDIFICATIO gof at least on 1 representative te t l specimen from at least every tenth batch ofs each type of wet radicactive was 1 solutichs)lter sludges) spent resins, evapor tor bottoms, ansodium sulfate \ (e.g.,\fi
- s. If-any test specimen fails to verif SOLIDIFICATION, the SOLIDIFICA- .
TIONofthebajchundertestshall suspended unti such time as additonal test specimens can be obtained, alternative OLIDIFICATION pahameters can be determined in accor' dance with the PR ESS CONTROL \ PROSRAM, and a sbbsequent test verif144 SOLIDIFICATION. SOLIDIFICATION \ of t e batch may then be resumed using the alternative S LIDIFICATION para. ters determi ed by the PROCESS C0 TROL PROGRAM. , ,
- b. If the initial test s SOLIDIFICATION, the $pecimen 0 CESS CONTROLfrom PROGRAM a bate shall.of waste provide fails to,yerify for the collection and testings of representative test specimens from each consecutive batch of the same type of w 't w'aste until at least consecutive initial test \ specimens demo..Jtrate SOLIDIFICATION,ythree sThe PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specificatibn 6.13, to assure SOLIDIFICATION of subsequent batches of waste. 'x
'\ v j \
3 k \ l
. g e
f . i LIMERICK - UNIT 2 3/4 11-19 AUL 2 5 IM
I fi 5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM - LIMIT!M CONDITION bR OPERATION 1 1
- 3. .1 The radiolog cal environmental e nitoring program s all be conducted as pecified in Table 3.12.1-1.
APPL CA8ILITY: At allstimes. ACT!0k: . a With the radio ogical envirorimenta)_ monitoring program not being l conducted as spbcified in Table 3.12.11, prepare andisubmit to the j al Environmental Operating Commission, Report per Specif1 in the 1 cation Annual 6.9.1.7, Radiologi(description a of.the\ reasons for
\notconductingtheprogramasrequir&dandthepla'nsfo preventing
( b. recurrence.
\
ththelevelofradioactivityastheresultofplant'effluentsin l angenvironmental sampling medium at a s ocified'Tocation exceeding thes reporting lev $1s '.of Table 3.12.1-2 when averaged over\any calendar quartter, prepare' ad submit to the Comminsion within 30 days, pursuant to Shecificat' - 9.2', a S i forehceedint N limit (s) pecial Report'that and defines identifiesactions the corrective tme cause(s) to be l taken to redwet .adioactive affluents so that the potentia 1\ annual ! dose to a MEMBER OF THE PUBLIC is less thanithe calendar year limits . of'Spechfications 3.11.1.2, 3.11.2.2 and 3.11.2.3. When more than ' one of the radionucli' des in Table 3.12.1-2 are detected in the' sampling medium, this report shall be submitted if:
\ . .
g concentration (1) , ! reporti g level (1)
- concentration (2) reporting level (2) + .. 1 1.0 When radionuclides other than those in Table 3.12,1-2 are detected and are the re'sult of plant effluents, this report shall be submitted ifthepotentialannualdosetoaMEMBEROFTHEPUBLICisequalto 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 radioactfyity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in -
the Annual Radiological Environmental Operating Report.
, ,. x .
\
e i 4 LIMERICK - UNIT 2 3/4 12-1 M 25M:I 4
- -,- ,.-,--, .,,., , -r----- . . , - - _ -
l RADIOLOGICALINVIRONMENTALMONITORING '
, (
LIMITINGCONDI\IONFOROPERATIbk(Continued) 1
!\' i 1
j r ACT 0N: (Continu'td) ,
,\ c. With milk or fresh leafy vegetable samples unavailable from one or i
more of the sample locations required by Table 3.12.1-1, identify locations for estaining replacement samples and add them to the
, radiological environmental monitoring program within 30 days.' The i specific locations from which samples were unavailable may then . t be deleted from the monitoring program. Pursuant to Specifica-tion 6.9.1.8,' identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s). - '
- e. \
- d. The provisions of Specification 3.0.3 are not applicabi'e.
SURVEILLANCE REQUIREMENTS ' s 4.12.1.1 The radiologien1 environmental monitoring sample's shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12.1-1,. the detection capabilities required by Table 4.12.1-1 y- ,
. LIMERICK - UNIT 2 3/4125 tm b W .
r- TABLE 3.12.1-1 S
- g , RADIOLOGICAL ENVIR0fGENTAL MONITORING PROGRAM i
. NtNSER OF REPRESENTATIVE -
i c EXPOSURE PATHWAY SAMPLES AND SAMPLIN(i AIS TYPE AfD FREMNCY SAMPLE LOCATIONS g ,)
'5 AND/OR SAMPLE / COLLECTJ6ft FREQUEf1CY OF AIIALYSIS ,
-d j m 1. DIRECT ' IATION(b) 40 routine monitoring stations At le'ast Quarterly. Gamuse dose at least either with two or more dosimeters - quarterly. l or with one instrument for measur,ing
' j and recording dose rate. contirmsously i
, / ;
'/ placed as follows: (1) Aq i6ner ring of stations, one in'each ~
meteorological secto6 in the general 3 area of the SITE-tOUNDARY; (2) An ' l outer ring of-stations, one in each ! l meteorological sector, in the 3-to ! R
, 9-mile range from the site; (3) The y l' balance of.the stations placed in .
- M spect.d ' interest areas such as !
I J,
/ population derr.es, centers, nearby resi-j schools, and in 1 or 2 !
i / areas to serve as c,ontrol stations. .
- 2. AIRBORNE
- Radiciodine and Samples from 5 locations: / Continuous sampler Radioiodine Cannister: l' Paff ticulates ' operation with sample I-131 analysis weekly.
l .
- a. 3 samples from close to the cc11ection weekly, or '
- 3 SITE BOUNDARY locations.(in more' frequently if l 4 different sectors)pf-the required by dust Particulate Sampler- -
loading.' highest calculated annual Gross beta r , average groundlevel x/Q. .. activity sis !
- b. I sample-from the vicinity followi 11ter :
l community having one of the * [ l , highest calculated annual .j, ge ! i:, highest groundlevel y/Q. ,j p ,I ,,g g , o - composite (by l m / location) at *least. ; a ,/ quarterly. t l l . 'i
TABLE 3.12.1-1 (Continued) C j!! RADI000GICAL ENVIl1000 ENTAL NDNITORING Pfl0 GRAM i g ._- g -
. MINSER OF REPRESENTATIVE e EXPOSURE P419 DAY SAMPLES AND SAMPLING Als TYPE Afe FREQUENCY
' i'im AND/OR SAMPLE SAMPLELOCATIONSk COLLECTION FREQUENCY OF ANALYSIS
/
m 2. AlftBORNE (Continued) /
/ -
I sample fryar'a control loca -
,/ / tion, as,for example 15-30 ka-
,/ distant'and in the least prev-al wind direction. =
- 3. WATERBORNE ,- ,
I
- a. Surface 1 sample upstream. Composite sampi over Gamma isotopic analysis'
~
1 sample downstream. .- 1-month period. monthly. Composite for w , tritium analysis quarterly.
~
2
- b. Ground
*/ Samples from 1 or 2 sources
@arterly. h isotopic' and tritium only if likely to be affected.h ,
7 analysis quarterly.
- c. Drinking 1 sample of each of,1-to Monthly. (Composite) I-13 lysis on each of the,neare ,sL water supplies c ite when the dose that,could-be affected by its . alculated for the con
- disfharge. sumption of the water !
j is greater than 1 area i
, , per year. Composite 'i 1 sample from a control . ' for gross beta and gamme location. ,
" isotopic analysesaponthly. !
Composite for triti , , . , . analysis quarter 1 .
- d. Sediment I sample from downstream area Semiannually. Gamma iso ic analysis i from with existing or potential semi 11y.
shoreline recreational value. : 5
.'a* . !
,_ , , ~ - - - -
~
/ T 3.12.1_1 (Continued)
C / 4 RADIOLOGICAL ENVIR0001 ENTAL MONITORING PROGRAM a - .
,n , -
__ __ / NUf9ER OF REPRESENTATIVE e EXP05UBE' PATHWAY SAMPLES AND SAMPLING AND TYPE Age FREQUENCY
'5 Ape /OR SAMPLE SAMPLE LOCATIONS (a) COLLECTION FREQUENCY OF ANALYSIS m , ..
t
- ^
4 - to / [ INGESTION -
/
/
- a. Nilk a Samples from milking animals Semimonthly when ani- Gamuna isotopic and I-131
. J.
/' in 3 locations within 5 km ,
distance having the highest
- mais are on pasture, analysis semimonthly when monthly at other times. animals are on pasture; , ;
dose potential.' If there are monthly at other times. j none, then, I sample from milk- l
- inq animals in each of 3 areas I between 5- to 8-km distance --
~
where doses are calculated to be 5 ~
- greater than 1 arem per yr[
w / - i 7 1 sample,from milking animals at a control location (15-30 km t Ja' distant and in the least , f prevalent wind direction). *
- b. Fish and a. I sample of two recreationally Sample in season, or Invertebrates important species in vicinity semiannually..if they ~~ analysis _ Gamma onisotopic edible of plant discharge ares.
are pot ~ seasonal. portions. I sample of same , species in - areas not influenc ant ! discharge. .- i
- t.
- e, y ,
v. s. _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ - - _ _ _ =
i ,_ TABLE 3.12.1-1 (Continued) E- 7 g / RADIOLOGICAL ENVIROISEllTAL IWIIITORING PROGRAM M
~
IRNSER OF REPRESEIITATIVE ,
,g EXPOSURE PAT W T SAMPLES AND SAMPLIIIG AfS TYPE AIS FREQUENCY ,
y Ale /OR SMIPLE ' , SAMPLE LOCATIONS (,) COLLECT 10ll FREQUE91CY OF ANALYSIS N 4. IIIGESTION . 7' . l ,/(Contipued)' - f' - ( .. Food Products Samples of 3 different Monthly when,'available, Gamuna isotopic' and kinds of broad leaf if milk samp' ling is
- I-131 analysis.
vegetation grown nearest not performed. .
.einch sf two different
.offsite locations of highes't y'
l - predicted annual average
, ground level D/Q if milk -
g . sampling is not performed. / i y 1 sample of each of th 9 s'imilar th1y when available, r - isotopic' and m broad leaf vegetationrown g if milk sampling is'-I-13L analysis. 15-30 km distant inethe least - not performed. N l prevalent wind direction if' milk sampling is not performed. 0 e~' '
~ * %,
- ) '
I
E
- 4!
'I i .
i' TABLE 3.12.1-},(Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
/
a
,/ TABLE NOTATIONS
~
Specificparametersofdistanceanddirectionec't'o'rsfromthecenterlineof e the two reactors andadditienti descri tion where per inent, shall be provided for each and everf' sample location 4n able 3.12.1-1 in a table and figure (s) in the ODCM. Deviations are pemitted Pom-thereq)edsamplingscheduleif specimens are unobtainable due to hazardous c'ondi ons, seasonal unavailability, maifunction of-automatic sampling equipment and r legitimate reasons. If -
- specim(eps shal madearetounobtainable due to sampling complete corrective equipinent action prior malfunction, to the send of the nextevery effort sampling od' All deviations from the sampling schedule shall be documented in the :
Annua} Radiological Environmental Operating Report pursuant.to Specification 6.9,1.7. It isIrecognized that, at times, it may not be possible or practicable to con-tinueltoobtainsamplesofthemediaofchoiceatthemostdesiredlocation or time. In these instances suitable alternative media and jlocations may be chosen for the particular pathway in question and appropriate s'ubstitutions ' made within 30 days in the radiological environmental menitoring program. Pursuant to Specification 6.9.1.8, identify the cause of the unavailability of samples for that pathway and identify the new lo, cation (s) for obtairling replacement samples in the next Semiannual Radioactive Effluent Release Report l and also~ include in the report a revised figure (s) and table for the ODCM, reflecting the new location (s). , x ,
/
b l 0ne or more instruments, such as a pressurized ion ch' amber faraneasuring and ! recording dose rate continuously may be used in place o~f, or in a ition to, integrating dosimeters. For the purposes oh.this table, a thermol minescent dosimeter (TLD) is considered to be on.e.phosphpr; two or more phos hors in a packet are considered as two or mort dosimeters; Film badges sha not be used as dosimeters for measuring' direct radiationx
/ ,/ >
C Methodology for-rec of radiciodirIshall be described in'the 00CH. J Airborneparticulatesample-fl(tersshallbeanalyzedforgrossbetaradio activity 24 hours or smre after sampling to allow for redon and thoron daughter decay. .,If' gross beta activity in air particulate samples is greater than 10 times the yearly mean of control, samples, gamma isotopic analysis shall be performed on the individual samples. '
/ .
' Gamma isotcIpic analysis means the identification and quantification of gamma-emitting /r'adionuclides that may be attributable to the effluents from the facility.
The "up
. ofthehtreamsample"shallbetakenatadistancebeyondsignificantinfluence ischarge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. *
/
- l l LIMERICK - LINIT 2 '
3/4 12 7 m25W
-- m
TABLE 3.12.1-1 (Continued) . RADIOLOGICAL ENVIR0t04EhAL MONITORING PROGRAM n TABLENOTA\ IONS 1 9 A composi e sample is one n which the quan ity (aliqvot) of iquid sampled isproportionaltothequanityofflowingliquidandinwhic the method of l sampling employed results in a specimen that is representative f the liquid flow. ::n this program compos te sample aliquot.s shall be colle ted at time ntervals that are very short e.g., hourly) r41stive to the compositing . riod(e.g., thly) in order to assure obtaining a representat ve sample, h Gro ndwater samples shall be taken when this source is tapped for drinking or irri ation purposet in areas where e hydraulic gradient or recharge properties s are s table for contamination. \
\
I Thedoseshallbecalculatedforthemaimumorganandsaegrou',using\the s methodoJogy and parameters in the ODCM. -
\ \
\
\
's N\
9 4 9 8 e e 9 LIMERICK - UNIT 2 , 3/4 12-8 E. 2 5 W . _ _ . . . - ~ - - , , _ _ _ - , , - . -. -- . . . . - . . - - - - - . - - - ~ . - .
C l z 9 TABLE 3 12.1-2 E REPORTING LEVELS FOR RADI0 ACTIVITY CO TIONS IN ENVIROISENTAL SAfrLES - Reporti evels
-e ,-
/ p m .
WATER AIRBORNEPARTICUtITE FISH MILE F000 PRODUCTS ANALYSIS (pCi/L) OR GASES (pC1/m3 ) (pci/ , wet) (pci/L) (pci/kg, wet) H 20,000* Mn-54 1,000 30,000 Fe-59 400 10,000 < Y* Co-58 1,000 30,000 f
; a Co-60 300 10,000 '
Zn-65 300 20,000
- Zr-Mb-95 400** -
I-131 2 0.9 , 3 100
~
Cs-134 30 10 1-000
, ' 60 1,000 l
Cs-137 50 20 2,'000 70 ,2,000 l ; 200**
'~
p , Ba-La-140 - 300 ,
. *For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists,.a value of p 30,000 pCi/L may be used. +
/
v, /
$ ** Total for parent and daughter. -
=
e 9
C t z g TMLE 4.12.1-1 DETECTION CAPABILIJ FOR ENVIRGISENTALJAFLE ANALYSISI *)
. L R LIMIT OF DETECTI D)(b)(c) ,
i to -
/ -
t AIR 80RNE PARTI'CULATE FISM MILK ,/ i ANALYSIS idATE[/L) (pC1 OR' GASES (pC1/m2 ) (pCS/kg, wet) (pci/L) F000 (pCi/kg. PRODUCTS
, wet (pCi/kg,[.dry)
SEDISENT
' O.01 Gross Beta' 4 ,
,/
- H-3 2000 '- ,/ /
Mn-54 15 i30'
^
w' Fe-59 30 .
- ,/'260 /
[ - g Co-58, 60 15 -/ 130 / 4 Zn-65 30 260 -
- o Zr-95 30 /
Mb-95 15 ' /
/ .
I-131 1(d) 0.07
~
, 1 60 -
! Cs-134 15 0.05 130 ' 15 e 60 150 Cs-137 18 0.06
- 150 18 80 180 Ba-140 60 La-140 15
' 60 15 .
/ ~!
* /
e ,/ O' 5-
,a ,- , . _ _ , _ .. . _
.. . _ .- -. ~_ . - _ - - - - - _
, 1 I
TABLE 4.12.1-1 (Continued)
., TABLE NOTATIONS (a)Thislis does not mea that only these nucli es are to be conside d. Other I peaks the are identif ble at 95% confidence level, together with hose of I the above nuclides, sh 1 also be analyzed an reported in the Annu 1 Radio-logical En ironmental Operating report pursuan; to Specification 6. 1.7.
(b) Required de ection capa lities for thermolumi scent dosimeters use for I
-environment measuremen:4 ue given in Regulatory Guide 4.13. .
ined, for pu ose of these specifications,.as the smalle t con-
)TheLLDis'de{radioactivematerialinasampl centration of.- e that will yield a net co nt j (abov.t system background) thht will be detected with 95% probability wi h only 5% probabt,11ty of faliely concluding that a blank observation represents a"real".signalj 1
\
\ i i
\
\ \
\ .
For a particular measurement system (which may include, radiochemical sekration): \ g LLD =
.,4.66 s b- \ \
\
E V 2.22 Y, exp(-AAt) i
\ s \
where \ \;
~
LLD ih the "a priori" lower limit of detection es de(ined above (as i picocupies per unit mass or volume), s is the standard deviation of the b'ackground counting rate or of the cNunting rate-of a black sample as appropriate (as counts per minute), < 1
\
E is-the counting efficiency (as countsiper disintegration), Vistheshmplesize(in'unitsofmassohvolume), s t , 2.22 is the pumber of disintegrations per minute per picoeurie,
\-
Yigthefractionalradiochemicalyield(whenapplicable), ; A is'the radioactive decay constant for the particular radionublide, and j At for vironme\ntal samples is the elapsed time between sample collection (or end. f the sample collection period) and time of counting. l Typical values of E, V, Y, and At shoyld be used in the calculation.
\ ,
A LIbiERICK - UN,IT 2 3/4 12-11 g25W
.N -
- o Table 4.12.1-1 (Continued) .
TABLE NOTATIONS' n It'shouldb.recognizdthattheIDisdef edasanlahriori(befo the fact) lini.t oprasenting the capab 11ty of a asurement stem and n t as an posterio'ri after the et) limit for a par icular measu ement. Ana yses all be perf ed in suc a manner hat the s'ated LLDs w 1 be achie d under r tine condit ns. Occas nally ba ground f' ctuations, avoidably all s pie sizes, t presence f interfe ing nucli es, or other uncontrolla e ci unstances ma render the e LLDs un chievable In'such ca es, the co rib-uti fact'rs o sha be identi ied and scribed i the Annual adiological _ ; Envi noen al Opera ing Repor pursuant o Specif atdon 6.9.1.7.1 (d)LLD or drinking wgter samples. 'g ' -
'\'
i
/
,e 4 ,e e
e 9 4 9 e S e S L L l LIMERICK - UN.IT 2 3/4 12-12
t RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE C{NSUS
\
LIMITING CONDITION FOR OPERATION \ 1
\ 1 3.12.2 A la'nd use census shall be conducte and shall identify wit in a distance of 8 km (5 miles) the location in a th of the 16 meteorolog cal sectors f the nearest milt minal, J the neare t residence and the nea est-garden
- greater the M f500 ft 2) produ ing broad leaf vegetatiop.
APPLICABI TY: At oli tks, \ \*
\
ACTION:
. a.
\
\ \
With a land use census 1dentifying a 1 ation(s) which yields a . calculated dose or dose commitment grea er than the values curren\ tly being calculated in Specification 4.11.2.3, identify the new loca s tion (s) in the next Semiannual Radioactive Effluent Release Report. pursuant to Specification 6.9.1.8. E
\
- b. Withalandusecensusidentifyingalocation(shthatyieldsa . \.
calculated dose or dose commitment (via the same exposure pathway) \ 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new loca-tion (s) to the. radiological environmental monitoring program within 30 days. The sampling location (s), excluding the control station location, having the lowest calcu' lated dose or dose commitment (s) (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this ~1and use census was conducted. Pursuant to Specification 6.9.1.8, identify the new 1 location (s) in the next Semiannual Radioactive Effluent Release Report
-and also include in the report a revised figure (s) and table for the 00CM. reflecting the new location (s).
- c. e provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE \ REQUIREMENTS The land use census shall.be Landucted during the growing seasoq at
\4.12.2 least once per 12 months using that information that will provide the best tesults, ~such as by a door-to-door survey, aerial survey .or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
\
\
- Broad leaf vegetation sampling of at least three different kinds of vegetation .
may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications fcr broad leaf vegetation ~ sampling in Table 3.12.1-1 item 4.c. shall be followed, including analysis of control samples. LIMERICK-- UN.IT 2 , 3/4 12-13 m 2 5 IMS 2
RADIOLOGICAL ENVIRONMENTAL MONITORING ' 3/4.12.3 TERLABORATORY C PARIS 0N PROGRAM
\
LIMITING CONDITION FOR OPERATION
\
\ j 3.12.3 Analyses.shall be performed on* radioactive aterials suppl ed as part '
of an Interlaboratory. Comparison Program which has en approved by the Commission. APPLICABILITY: At'all times. . ACTI :
\
- a. With analys s not.being performed as required 'above, report tbe corrective actions taken to prevent a recurrende to the Commission in the Annual Radiological Environmental Operating Report pursuant o Specificatioh 6.9.1.7.
\
- b. They provisions of Specification 3. 3 are not app icable.
\ x 5 RVEILLANCE REQUIREMENTS
\
4.12. The Interlaboratory Comparison Program d, sha 1 be descr
. A su iry of the results. obtained as part of the ab ye required Int riaboratory Compari on Program shall be included in the Annual Radiological Environmental '
Operati 1 Report pursuant to Specification 6.9.1.7. .
)
e 0 O 4 4 LIMERICK - UNIT 2 3/4 12-14 EG252
,- - . . - . . -. ,~ .,,
A - - A44-.- m._ a
* .*j' . .
- a g
=
4 s
" ,. t .
i e-%. e-
' 'I [,
. - j k
BASES FOR SECTIONS 3.0 AND 4.0 k LIMITING CONDITIONS FOR OPERATION " AND
. SURVEILLANCE REQUIREMENTS- .
9
. i 1.
l [ 1 .. > l' I 1. \ . GO
.. .u -
. ?.. ,
- t t
INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to - the reactor vessel. This instrument (. tion does not provide actuation of any of the emergency core cooling equipment. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is ac'ceptable on the basis that the dif- - forence between each Trip Setpoint and the Allowable Value is an allowance for
. instrument drift specifically allocated for each trip in the safety analyses. -
3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION l The control rod block functions are provided consistent with the requir0-ments of the specifications in Section 3/4.1.4, control Rod . Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control. rod block. ' Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2).the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available en selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General-Design Criteria 19, 41, 60, 61, 63, and 64. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the retsponse of those features important to safety. This capability is required to permit comparison of the measured response to . that used in the design basis for the unit. s3/4.3.7.3 METEOROLOGICALMONITORINdINSTRUMENTATION The OPERABILITY of the teorological monitoring instrumenta ion ensures at'50Kicientmeteorologi ata is v lable for estimating' po ntial radia-t on doses to Qeblic as a sult of rou ine or g dental,telea e of radio-ac ivt sateria s to the ai.mosphehe. Thiscaabili1yisrequifedto' valuate the nded for i itiatihg p rotectife measures t pro'ect t e he,alth an safety of ttie publig. This i umentati'n e i s consis en with t e ecommend t ons of Regul oryG(ide1.23" ite Meteo ogical P g ams," Fe r ary, 1972. LIMERICK - UNIT 2 B 3/4 3-4 E25W l
INSTRUMENTATION BASES METE 01t0 LOGICAL MONITORING INSTRUMENTATIO (Continued) Sed compil d ince a uary 72 rovid c relati between av- ' tio 1 Towir 1 and le atio 1 Towe 2), nd b twe n Ele at n 2 (T wer 1) and lev ti >n 2 Tow r 2) his orr lation ser es as jus ific tion or the - use te ppro i e To e 2 ins r nt as a ack-up t the T er instrument as shown Table .3.7.3 .
~
3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS - The OPERABILITY of the remote shutdown system instrumentation and controls i ensures that sufficient capability is available to permit shutdown and mainte-nance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is requited in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part'50, Appendix A. The Unit 1 RHR transfer switches are included only due to their.p'otential impact on the RHRSW system, which is common to both units. ' 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is con-4 sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation - for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. ,
~
3/4.3.7.6 SOURCE RANGE MONITORS __ The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup
- and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the inter-mediate range monitors are on scale, adequate information is available without
_ the SRMs and they can be retracted. 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM - The OPERABILITY of the traversi$g in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from Z . use of this equipment accurately represent the spatial neutron flux distribution
= of the reactor core.
The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e., detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core', thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demon- - i strated by comparing the detector (s) output in the resultant heat balance calcu-lation (P-1) with data obtained during a previous heat balance calculation (P-1). 1 LIMERICK - UNIT 2 B 3/4 3-5 I i ! AUG 2 5 IH3 d
INSTRUMENTATION .
.)Q
-SASES
'k 3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS i The OPERABILITY of the chlorine and toxic gas detection ' systems ensures
.that an accidental chlorine and/or toxic gas release will be detected promptly and-the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection, ypon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to ~
provide the required protection. The detection systems requir'ed by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95 " Pro-tection of Nuclear Power Plant Control Room Operators'against an Accidental Chlorine Release," February 1975. 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION ;. OPERABILITY.of the detection instrumentation ensures..that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extin-guishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an ' integral element in the overall facility fire protection program. Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than
' dete: tors that are installed solely for early fire warning and notification.
Consequently, the minimum number of OPERABLE fire detectors must be greater. The loss-of detection capability for fire suppression systems, actuated by fira detectors, represents a-significant degradation of fire protection for any att a. As a result, the establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system , and avoid or mitigate damage to primary system components.- The allowable out-of-service times and surveillance reduirements are consistent with the recom-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System. of Light-Water-Cooled Reactors," May 1981. 3/4.3.7.11 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instru-mentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. LIMERICK - UNIT 2 . B 3/4 3-6
,, M6 2 5 IM
INSTRUMENTATION , BASES bbME 3/4.'3.7.12 - A.MC GAS 50tHHfft9fM7 MONITORING INSTRUMENTATION Th= red!:::tiv; ; :;e.. Tileent in:tr&-a+=+ 4aa is pre'?!d;d te :.; ,it:-
- nd cent.el, e; :ppli:ble. +ha_ releases of radioactive matarimic in naseous '
- ffleente during ect.el er pet;nti:1 r:100 ; vi gaseous effiuvni... Th: :hrm/
trip e rt?? a+ 5 fr= + h=== iactre ent; ; hell be cel;.leted in e;;erden;; with i l th: pr;;ed., e; S th: 000", te eneere thet the ;,lerr.v% ip wiii occur prior i.o er::: din; the limit; ef 10 Oi" P;rt 20. This instrumentation e4ee includes provisions for monitoring the concentrations of potentially explosive gas mix- ~ tures in the off gas 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. 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM i ! Thisspecificationisprovidedtoensurethattheturbinehv'erspeed . protection system instrumentation *and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection , from turbine excessive overspeed is required since excessive overspeed cf the
- turbine could generate potentially damaging missiles which could impact and
- i. damage safety related components, equipment or structures.
3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION
-The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the. event of failure of feedwater controller under maximum demand.
i e LIMERICK - UNIT 2 B 3/4 3-7 m 2 5 ISS3
3/4.11 RADI0 ACTIVE EFFLUENTS BASES \, k \ 3/4.11.1 LIQUID \FFLUENTS \ [ 3/4.11.1.1 CONCEN\ RATION l This specifica ion is provided to ensure that the concentrati n of radio-l ctive materials rel ased in liquid waste effluents to UNRESTRICTE AREAS will
. e less than the con ntration levels specified in 10 CFR-Part 20, pendix B, ble II, Column 2. his limitation provides additional assurance t at the levels o
radioactive materia s in bodies of water 19 UNRESTRICTED AREASiwiil result in . !. ex osures within (1) the Section II. A design o jectives of Appendix I,\ 10 CFR . p Pa 50, to a MEMBER OF'THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to the L pop lation. ! The concent ation limits for dissolved or entrained noble bases are based upon the assum tion that Xe-135 is the controlling radioisotope and
-its MPC in air was converted to an equivalent concentration in water usisg the methods described in the International Commission on Radiological Protectton p
(ICRP) Publication 2. ' , \ TheYequired detection capabilities for radioactive materiah in liquiy waste
- l. samples are tabulated in terms of the lower limits of detection (LLDs). Detpiled discussion \of the LLD, and other detection limits can be found in HASL Procedures i Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Djtec-tion and Quarttitative Determination - Application to Radiochemistry," Anal. Chem. _40, L 586-93 (1968)\ and Hartwell, J. K., " Detection Limits f 9r Radioanalytical Counting i L Techniques," Atlantic Ri,chfield lianford Company Report ARH-SA-215 (June 1975).I 3/4.11.1.2 OOSE \\
This specification is provided to implement the requirements of Sections II.A,
-III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Opera-ion implements the guides set forth in Section II.A of Appendix I. The ACTION tatements provide the required operating flexibility and at the same time imple-nt the guides set forth in Section IE A cf Appendix I to assure that the releases of\ radioactive material in liquid effluents will be kept "as l'ow as is reasonably
-achievable." Also, for fresh water sites}with drinking water sypplies which can be potentially affected by plant operations, there is reasonable assurance that the ,
operation of the facility will not result fp radionuclide concentrations in the finished drinking water that are in excess gf the requirements of The dose calculation methodology and parameters in the ODCMment imple\40 CfR Part 141. the require-ments in Section III. A of Appendix I that conformance with the guides of Appendix I be show by calculational ocedures based on odels'and data, such that the actual
~
, exposure of a MEMBER OF THE UBLIC through appr priate' pathways is unlikely to be substantially underestimated The equations spe ified in the 00CM for calculating thedosesquetotheactualrlease'ratesofradaactivematerialsinliquidefflu-ents are consistent with the methodology provided n Regulatory Guide 1.109,
." Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose Af Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents
'from:AccidentakandRoutineReactorReleasesforthePurposeofImplementing Appendix I," Aprsil 1977.
\
This specification applies to the release of radioactive materials in liquid ef fluents from the\ site. s s LIMERICK - UNIT 2 ' B 3/4 11-1 AUG 2 51933
- e :9 '
RADI0 ACTIVE EFFLUENTS
- BASES
' 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM .
-Th requir ent that he approp late portio s of this system be sed when s cifie provi assuran that th releases o radioact ve materia s in liquid of luents will b ept "as w as is asonably a ievable". This sp ification i lesents the re irements f 10 CFR 0.36a, Gene 1 Design Criterion 60 of App ndix A o 10 C Part 50 d the de ign objecti e given p Section I.D of.
Appe dix I t 10 CF Part 50. The specified limits overningitheuseo appro-pria portio s of t liquid r dwaste treatment syst m were specified a a suit le fraction of the dose de ign objehtives set f th in S tion II.A of , AppendjxI,10CFRPart50,fori x uid eff ents. , , 3/4/11.1.4 LIQUID HOLDUP TANKS T,he tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable,of holding the tank contents and that de not have tank overflows and surrounding area drains connected'to the liquid radwaste treatment system.. . 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 syrface water supply in an UNRESTRICTED AREA.
3/4.11.2 GASEOUS EFFLUENTS . 3/4 11.2.1 DOSE RATE This s cificati n is provided to ensure that the dose at any time at and be end the S TE BOUNDA Y from gasedys effluents from the sit' will be w1 hin the annual do e limits, f 10 CFR Part 20 to U ESTRICTED ARE 5. The ann al dos limits a the dos associated ith the c centrations o 10 CFR Par 20, Appehdix B, Ta le II, Co umn 1.- The limits-p ovide reasona le assuranc that radioactive mat rial disc arged in ga eous efflu nts will not esultinthh exposure of a M MBER OF T PUBLIC in n UNRESTR CTED AREA, ei her within of outside the SIT BOUNDARY, to annual a erage concentrations ex eeding the \ limits specified in Appendi B Table II of 10 CFR Part 20 (10 FR 20.106(b)(1)). 1 For MEMBERS OF T PUBLIC w o may at times be withi.n the SITE B4UNDARY, the i occupancy of that MEMBER OF s HE PUBLIC will usuallyibe sufficiently low to com-pensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Ex pies of calculations' for such MEMBERS OF THE!PUBLIC, with\ he appropriate oc pancy factors, shall be given in the s ODCM. The specified \ lease rate limits restrict, kt all times, the corresponding ganga and beta-d6se r es above background to a MEMBER OF THE PUBLIC at oi\,beyond the \ SITE BOUNDARh to essthanorequalto500aress/yeartothetotalbodyortoleisthanor equ 1 to 3003 mrems/ygar to the skin. TPese release rate limits also restrict, at a 1 times, the corresponding throid dtse rate above' background to a child via t e inhalation pathway to less 'than or equal to 1500'mrems/ year. Thisspecification\a'ppliestotherelea:3e of radioactive materials in gaseousieffluents from the site. -
\. s LIMERICK - UNIT 2 8 3/4 11-2
,. AUC 2 5 IH3
RADI0 ACTIVE EFFLUENTS " A * . BASES ! DOSE RATE (Continued) The quired etectici capability f radioactive materi s in gaseous L we te samp es are t bulate in terms of th lower limits of det etion (LLDs). i Det iled di cussion f the D, and other tection limits can b found in HAS Proced es Manu 1, HASL 300 (revised an ually), Currie, L. , " Limits
-for iualitat e Detec ion and\ Quantitative De erminatio - Applic tion to !
L Radic hemistr ," Anal. Chem. 40, 586-93 (1968 and Hart ell, J. K 1 tion iLits for Radioan tical Counting Techn ques," At1 ntic Ric\, hfield "Detec- - Hanfo Compan \ Report l
-SA-215 June 1975). \ ;
3/4.11. 2 DOSE-N06i.EG\SES ' Thisspecifictionisp\ovided oimplementtherequire3ents'of Sections .I.B III. and IV.'A of App'endix I,10 CFR Part 50 \ The Limiting
- Condition for OperatN,on implements the guides set fo' r th in Sectio'n II.B of Appendix I. The ACTION statements pro 9ide the requir'ed operating flexibility and at the same time implement the guides set forth in.Section IV. A of Appendix I to assure that the rel' eases of radioactive material in' gaseous effluentsswill bekept"aslowasisreasonablyachievagle". The Surveillance itequirements i implement the requirements in Section III.A of Appendix I that conformance n
with the guides of Appendix I be shown by calculational procedures based on models and data such thatx he t actual exposdre of a MEMBER'0F THE PVBLIC through s
' appropriate pathways is unlikely to Qe substantial-ly underestimated. The dose calculation established in\the 00CM for calc 61 sting the doses due f.o the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide \1 .109, '"Calcula-tion o Annual Doses to Man from Routink Releases of Reactor Effluents for the Purpt.se f Evaluating Compliarice with 10\CFR Part 50, Appendix \1," Revision 1, Octobe.' 1 77 and Regulatory Guide 1.111, " ethod( for Estimating Atmospheric (Transport nd Dispersion of Gaseous Efflue s in Routine Releases from Light-Vater Cooi. Reactors," Revision 1, July 19 7 with\ site spepific\ dispersion s
curves and ap osition methodology. TheODCMequationsprovidedfArdetermining theairdosenatandbeyondtheSITEBOUNDARYsarebayeduponthehistorical , av rage atmos'pheric conditions, s
\
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3/4 11.2.3 00$E - 10 DINE-131. 100INE-133, TRITIUM, AND RADIONUCLIDES IN < PARTICULSTE FORN - t
- s \
This specif:Ication is provided to implements the requirements of Sectiods II.C, III. A, and IV. A of Appendix I,10 CFR Part '50. The Limiting Conditions for Opieration are the guides ' set forth in Section II.C of Appendix 1. The ACTION statements provide the requirAd operating \ flexibility and at the same time implement the guides set forth~ in Section IK. A of Appendix I to assure that the releases of stadioactive materials in gaseous effluents will be kept"aslowasis'teasonablyachievable."\The00CMcalculationalmethods specified in' the Suryeillance Requirements i'mplement the\ requirements in SectionIII.AofAppendix'IthatconformanceYiththeguidesofAppendixIbe shown by calediational procedures based on models and data) such that the
; actual exposure of a MEMBER OF THE PUBLIC thro 0gh appropriate pathways is LIMERICK - UNIT 2 ) B 3/4 11-3
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RADI0 ACTIVE EFFLUENTS T' 4 BASES [i . X)SE-10 DINE-131.10 DINE-133, TRITIUM. AND RADIONUCLIDES IN PARTIC' ULATE l: 'ORM (Continued) \ . unlikehy to be substant 11y underestimated. Tfw ODCM calculational methods i for calculating the dose due to the actual release rates of the subje t , 1 materials are consistent ith the methodology provided in Regulatory G ide 1.109,
" Calculation of Annual Dos s to Man from Routine Releases of Reactor E fluents 1 for the Purpose of Evaluat g Compliance with 10 C Part 50, Appendix I,"
Revision l', October 1977 an Regulatory Guide 1.111 " Methods for Estimating - Atmospheric Transport and Di ersion of Gaseous Effl ents in Routine Releases from Light-Water-Cooled React s," Revision 1, July 977 with site specific dispersion curves and depositi methodology. These squations also provide for. i determining the actual doses ba ed upon the historica' average almospheric i' conditions. The release rate sp cifications for iodin -131, iodine-133, tritium,'and radionuclides in pa iculate form with hal -lives greater than. 8 days are dependent on the exist g radionuclide pathw s to man'in areas at
*and beyond the SITE BOUNDARY. The athways which were e amined in the develop-pont of these calculations were: ( ) individual inhalat n of airborne radf,o-quclides, (2) deposition of radionuc ides onto green leaf vegetation with subsequent consumption by man, (3) de osition onto grassy reas where milk ~ .
animals and meat producing animals gra e with consumption f the milk and me t by man, and (4) deposition on the grou with subsequent ex osure of man, k 3/4.11.2.4 VENTILATI EXHAUST TREATMENT SYSTEM p The requirement th t the appropriate p rtions of this syst m be used, when specified, provides reas nable assurance tha the releases of r ioactive '
\ materiais in gaseous affl ents will be kept " s low as is reason ly achievable."
s This spehification implem ts the requirements of 10 CFR 50.36a, eneral Design
\ Criterion \60 of Appendix A o 10 CFR Part 50, a d the design obje ives given n Section\II.D of Appendix to 10 CFR Part 50. The specified li its governing t a use of propriate'porti s of the systems we'e specified as a suitable fr etion of t e dose design o ectives set forth in Sections II.B and II.C of AppndixI,10$FRPart50,for'gaseouseffluents.
3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided t'o ensure that the concentration of poten-tially explosive gas mixtures contained in the main condenser offgas treat-ment system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen below its flammability limit provides
' assurance that the releases of radioactive materials will be controlled in -
conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part-50. O e LIMERICK - UNIT 2 B 3/4 11-4 AUG25 E
RADI0 ACTIVE' EFFLUENTS
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BASES i I 3/4.11.2.6 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main l condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently , discharged directly to the environment wi.thout treatment. This specification ' implements the requirements of General Design Criteria 60 and 64 of Appendix A , to 10 CFR Part 50, 3/4.11.2.7 VENTING OR PURGING , s Thisspecificakionprovide reasonable assurance that releases from dr ell purging operations will ot exceed the annual dose limits of 10 CFR ~ l -Par 20 for UNRESTRI TED AREAS. , i 3/4. .3 SOLID RADWA TE TREATMENT T e OPERABILITY o the solid ra aste system sures that the s stem will i be avai able for.use w never solid r dwastes requi e processing and ackaging ; prior. t being shipped ffsite. This pecification implements the req irements of 10 CF 50.36a and Gen ral Design Cri erion 60 of ppendix A to 10 C Part 50. q The proce s parameters i luded in esta lishing the. OCESS CONTROL PRO AM , may inclu , but are not imited to wast type, waste H, waste / liquid /s idifi- i cation age / catalyst rati s, waste oil e ntent, waste rincipal chemical constituent mixing and c ring times. > 3/a.11.4 TOT L DOSE This spect ication is pr vided to meet t e dose limita ions of 40 CFR Par 190 l that ave been-i corporated in o=10 CFR Part by 46 CFR I 25. The specifica ion requir s the prep ration and su mittal of a Sp ial Report wh never the calcu-
; lated ses from ant radioact e effluents ex eed twice the esign objective doses o Appendix For sites ontaining up to four reactors, it is highly niikely hat the r ultant dose o a MEMBER OF E PUBLIC will xceed the se limit of:40 CF Part 190 if he individual actors remain ithin the r orting r quirement evel. The' ecial Report w 11 describe-a utse of ac ion that hould res t in the li itation of the nual dose to member of the ublic to ithin the 40 CFR Part 190 limits. Fo the purposes the Special Repo t, it may e assumed that the do e commitment to the MEMBER OF E PUBLIC from ther urantum fuel cy le sources s negligible, w~th the excepti that dose c ntributiohs from oth nuclear el cycle facili ies at the sam site or with n a radius 1 of 5 mile must be c sidered. If t dose to any MEMBER OF THE P LIC is estimated to xceed the equirements of 0 CFR Part 190, the Special R4 port with ) request r a varian e (provfded th release conditions i resulting 1 v iolatioh of 40 CFR Part 190 h ve not already een corrected), in j accordance th the prqisions of 0 CFR 190. 1 and 10 CFR 0.405c, is considered to be a timel request apd fulfill the requi nts of 40 CFR Part 190 until j NRC staff acti n is completed. An ir. ividual i -not considered a MEMBER OF THE l PUBLIC during ahy period in which he/she is engaged in carrying out any operation '
that is part of the nuclear fuel cycle. LIMERICK - UN.IT 2 B 3/4 11-5 A'$ 2 5 2
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t 3/4.12' RAD LOGICAL ENVIRONMENTAL MONITORING
} BASES'
\
3 4.12.1 MONIT RING PROGRAM The radiological environmental nitoring program te utred by this spec fica-ti provides representative measurem ts of radiation an of radioactive sat rials in those\ exposure pathways a d for those radionuclides that lead to the pighest potential radiation exposures of MEMBERS OF THE\PUBLIC resulting from;the station operation. Thismonit#ingprogramimplem4ntsSectionIV.B.2 of. Appendix I to 10 CFR Part 50 and thereby supplements the ' radiological efflue t . monitoring program by'iverifying that the measurable concentrations of radioactiv , materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental ex'posure pathways. The initially specified monitoring program kill be effective.for at least the first 3 9 ears of commercial operation. Follbwing this period, program changes may be initiated based on operational experience. The required detection' capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environrtental 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 measure-ment system and not as an a posteriori (after theifact) limit for a.particular measurement.- - Detailed discussion of 'the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually); Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968); and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford . Company Report ARH-SA-215 (June 1975);
\ '
3.4.12.2 ' LAND USE CENSUS Thisspecificatiokisprovidedtoensurethatchanges.intheuseofareas ' at d beyond the SITE BOUNDARY are identified and that modifications to the
.moni ring program are mhde if required by the results of this census. The i best formation from the\ door-to-door survey, aerial survey or consul'ing with I al agricultural authorities or any combination of these methods shall
.be used.\ This census satisfies the requirements of Section IV.B.3 of Appendix I
! to 10 CFRsPart 50. Restricting the census to'. gardens of greater than 500 l square fee't provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the 1 minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Re6ulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the f^ollowing assumptions were used: (1) that 20% of l 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. a \ [ l
- LIMERICK - UNIT 2 8 3/4 12-1 g250
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.. . .. i RADIOLOGICAL ENVIRONMENTAL MbkT0itING (4ASES , ,
1 3 .12.3 INTE. LABORATORY COMPARI N PROGRAM l: \
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\ l The requirement for participation in an Interlaboratory Comparison Program is p vided to ensure that independept checks- on the precision and accuracy of the measurements 'of radioactive material in environmental sample matrices are performed as part 'of.the quality assurance program for environmental monitoring in order.to demonstrate that the results are reasonably valid for the purpose -
-of Section IV.B.2 of Appendix I to 10 CFR Part 50.
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- 5. 2 CONTAINMENT '
CONFIGURATION i 5.2.1. The primary containment is a steel lined reinforcect concrete structure consisting of a drywell and suppression chamber. The drywell' is a steel-lined reinforced concrete vessel in the shape of a truncated cone on top of a water filled suppression chamber and is separated by a diaphragm slab and conricted to the suppression chamber through a series of downcomer vents. The drywell , has a maximum free air volume of 243,580 cubic feet at a minimum suppression , pool level of 22 feet. The suppression chamber has a maximum air region of ! 159,540 cubic feet and a minimum water region of 122,120 cubic feet. - DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is di' signed and shall be maintained for: L a. Maximum internal pressure 55 psig.
- b. Maximum internal temperature: drywell 340*F.
suppression pool 220'F.
- c. Maximum external to internal differential pressure 5 psid.
. d. Maximum floor differential pressure: 30 psid, downward.
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~ ADMINISTRATIVE CONTROLS
- PROCEDURES AND PROGRAMS (Continued) .
- 6. 8. 4 The foliowing programs shall be established, implemented, and maintained:
- a. Primarv Coolant Sources Outside Containment ,
A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a i
'
- serious transient or accident to as low as practical levels. The .
systems include the core spray, high pressure coolant injection,
' reactor core isolation cooling, residual heat removal, post accident tampling system, safeguard piping fill system, control rod drive scram . '
, discharge system, and containment air 'anonitor systems. The program '
y shall include the following: ',
- 1. Preventive maintenance and period (c visnal inspection requirements, and
- 2. Integrated leak test requirements' Tor each system at refueling cycle intervals or less. , .
- b. In-Plant Radiation Monitorinc '
A program which will ensure the capabi.lity to accurately detemine the airborne iodine concentration in vital areas under accident conditions. This program s, hall include the following:
- 1. Trainingofpersont)el, _ , '.
? .;
- 2. Precedures for monitoring, and .'
- 3. Provis'icns for maintenance of samp. ling and analysis equipment.
'. 'c . Post-accident Samplinc* '
' A program 'which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous afflu-ents, and containment atmosphere samples under accident conditions. -
The program shall include the following:
- 1. Traini.ng of personnel,
- 2. Procedures for sampling and analysis, and
- 3. Provisions for maintenance of sampling and analysis equipment.
"Not required until prior to exceeding 5% of RATED THERMAL POWER.
eneric Letter 89- 01 Enclos re 3-6.0'ADMINISTRATIVECONTR$LS ! 4/h PROCEDURES AND PROGRAMS h d d 6geda150w3*FPsrogrNatNtgAhb1Ti%ey/ikip y ,
.h..R'adioactiveEffluentControlsProgram 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.
i 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 gaseots monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the d ODCM,
- 2) Limitations on.the. concentrations of radioactive material l released in liquid effluents to UNRESTRICTED AREAS conforming to 1
-10 CFR Part 20, Appendix B, Table II, Column 2,
- 3) Monitoring, sampling, and analysis of radioactive liquid and =
gaseous effluents in accordance with 10 CFR'20.106 and with the methodology and parameters in the ODCM, ; 4). Limitations 'on the annual and quarterly doses or dose commitment I to a MEMBER OF THE.PUBLIC from radioactive materials in liquid effluents released from each. unit to WNRESTRICTED AREAS conform-ing to Appendix I to 10 CFR Part 50,
- 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and '
current calendar year in accordance with the methodology and parameters in the ODCM at -least every 31 days, } Limitations on the operability _and use of the liquid and gaseous j 6) effluent treatment systems to ensure that the appropriate 1 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 conforming to Appendix I to 10 CFR Part 50,
- 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY ]
conforming to the doses associated with 10 CFR Part 20, Appendix B,. Table II, Column 1, l
/ ZIAE91% LLNM $. . ggQ
neric. Letter 89-01 . Enclosure 3'
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12 1 2; ADMINISTRATIVE CONTROLS 6.8.4 . Radioactive Effluent Controls Program (Cont.) :
- 8) Limitations on the annual and quarterly air doses resulting from' ,
noble gases released in gaseous effluents from'each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to
, 10 CFR Part 50, ,
I
- 9) Limitations on the annual and quarterly doses to a MEMBER OF THE.PUBLIC.from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond-the -
SITE BOUNDARY conforming to Appendix- I to 10 CFR Part 50,
- 10) Limitations on venting and purging of the Mark 11 containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable r. . m .m m u m,- -
ano
- 11) Limitations on the annual dose or dose commitment to any MEMBER E
OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190, p -Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radio- _ l nuclides in the environs of the plant. The program shall provide (1) representative measurements of radioacti.vity in the highest potential exposure pathways, and (2) verif,1c'ation of the accuracy of the effluent monitoring program and.modeling of environmental expo'- sure pathways. The program 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, '
b 2) A Land Use Census to ensure that changet in the use of areas at ; t and beyond the SITE BOUNDARY are identified and that modifica-tions to the monitoring program are made if required by the j results of this census, and
- 3) ParticipationinafInterlaboratoryComparisonProgramtoensura 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.
H4 b
%ana Onks
b,- , a 3893224150
'Q ADMIN!STRATIVE CONTROLS y.,
~
, 6.9 REPORTING RE0VIR!MENTS y
ROUT!NE REo0RTS - 6.9.1 In addition to the applicable recorting reovirements of Title 10. Coce ' of Federal Regulations, tne following eports shall be submitted to tne Regional Administrator of tne Regional Of fice of the NRC unless otnerwise notec. . STARTUP REPORT
- 6. 9.1.1 A summary report of plant startup and power escalation testing shall
- be submitted following (1) receipt of an Operating License, (2) amencment to the license ir.volving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel sucplier, and (4) mocifications that may have significant!y altered the nuclear, thermal, or hydraulic performance of the unit.
- 6. 9.1. 2 The startup report shall, address each of the tests identified in Sub-section 14.2.12 of the final Safety Analysis Report and shall include a descrip- -
tion of the measured values of the operating conditions or characteristics , obtained during the test program and a comparison of these values with cesign predictions and specifications. Any corrective actions that were required to obtain satisf actory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report, , i
' 6. 9.1. 3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence -
ment of cemmercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the startup report does not cover all three events (i.e. , initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted ; at least every 3 montht until all three events have been completed. ANNUAL REPORTS
- 6.9.1,4 Annual reports covering the activities of the unit as described below . i for the previous calendar year shall be submitted prior to Marsh 1 of each year, j
.The initial report shall be submitted prior to March 1 of the year following initial criticality.
- 6. 9.1 ; Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the nut.er of station, utility, and other personnel (including contract *rs) receiving exposures greater ;
than 100 mrem /yr and their associated man-rem exposure according to ! work and job functions ** (e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance 3 (describe maintenance), waste processing, and refueling). The dose ' assignments to various duty functions may be estimated based on pocket "A single submittal may be made for a multiple unit station.
**This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.
LIMERICK-UNIT [ 6-15 h g tyg ;
mese .
,.- = , - . . - _
fc., ADMINISTRATIVE CONTROLS ' 3893224150 p - , D.'.9 ANNUAL REPORTS (Continued)
.. T dosimeter, ments. Smallthermoluminescent exposures totalling less dosimeter tnan 20% (TLD),
of the or film bad!e in ividual measure 6fb " total dose need not be accounted for. In the aggregate, at least Mi ' SO% of tne total wnole body dose received from external sources should f3 r . be assigned to specific major work functions; ! h,[ b. Documentation of all challengas to safety / relief valves; and I 5". : ' c. Any other unit unique reports required on an annual basis. !
- d. The results of specific activity analysis in which the primary 6 '
N, coolant exceeded the limits of Specification 3.4.5. The following - ' information shall De included: (1) Reactor power history starting l
- b. 48 hours prior to the first sample in which the limit was exceeded;
- k. (2) Results of the last isotopic analysis for radioicdine performed l h,' prior to exceeding the limit, results of analysis while limit was gl. exceeded and results of one analysis af ter the radiciodine activity ff O.
was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup
- P system flow history starting 48 hours prior to the first sample in I. which the limit was exceeded; (4) Graph of the I-131 concentration '
h, and one other radiciodine isotope concentration in microcuries per s,-_ oram as a function of time for the duration of the specific activity
!.. inbove the steady-state level; and (5) The time duration when the i
? specific activity of the primary coolant exceeded the radiciodine
'y ' limit.
('U MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, includ-e ing documentation of all challenges to the the main steam system safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555,.with a - 3 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, b ANNUAL RADIOLUGICAL ENVIRONMENTAL OPERATING REPORT *
~
/ '
6.9.1.h The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted s before May 1 of each year. The report shall include summaries, interpreta-
- f. tions, and analysis of trends of the results of the Radiological Environmental X Monitoring Program for the reporting period. The material provided shall be 5 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. 4
'A single submittal may be made for a multiple unit station.
3 - EK* UNIT [ 6-16 Amendment No. 20 MMi9I$ 1
- - _ . - _ _ _ - _ _ . - - - - - - - - - - - - - - - - - - - - - ~ ~ ~ ~ ' ~ ' " ~ ~ ~ ' ~
. t ADMINISTRATIVE CONTROL $ g r, ANNUAL \ RADIOLOGICAL \NVIRONMENTAL \PERATING REPO\T (Conti
\
environm\ ental survei anceactivitiesforthereprtperiod,kn\ uding a om-parison (as appropria g), with preoperational stu es, operation 1 control and previous environmental surveillance reports an , an assessment of the o - erved impacts of the plant operation en the envirohment. The reports shal a so includi the g results\of' land use censuses required by Specification 3.1 2.
\
The Annual Radjological Environmental Operating Repor\s shall include the res its
\
of a 1 radiological environmental tamples'and of all environmental radiation \ measu ments taken during the report period pursuant toithe locations specifiedi in the tables andsfigures in\the OFFSITE DOSE CALCULATION MANUAL, as well as . summari ed and tabulated results of these analyses and measurements in the format of the tablesin the Radiological Branch Technical Position, Revision 1, November'979. In the event that some individual res'ults are not available for inclusion ith the report, the r' aport shall be submitted noting and explairring the reasons for the missing results. The missing data shall be i submitted as
/
soonaspossqleinasupplementaryreport. The reports sh'all also include the following: a summary descri ion of the radiological environmental monitoring program; at least two legible maps **
**0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
LIMERICK.- UNIT 2 6-16a
; AM 2 5 im
1 3892300520 l . . l l
- ADMINISTRATIVE CONTROLS O ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- c in all sa plin ocations kgyed' a tance ydirections fr m center ine o .tgble e resdits g ho*i.: l ens e par e t n in .he In glab atory Comp venetorplan,/r son Tootam ired )y pe f ati n 3 3; disc sslo orgi devi)tions the Sappli g e e of le 4 .1- no discu sion at Analftas io/wM he LLD uir by Tab e - 2. -1 was qt achie ble. .
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT
- 6.9.1.6The Semlannual 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 l'ywid and gaseous effluents and solid waste released from the unit. The material r,covided shall be
.(1) consistent with the objectives outlined in the OCCM and PCP and (2) in cen- .
formance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. e,. m- ,y,.s.~,,,......, ..s.........., . . . . . .. u , . . . , . . . , gvi nouie ive Ma rials in Licuid MGaseegsMf fhR . from light-Wat ecledNuc1har Fow an , " Revision.IrJuh 1974, with dA mmarized on a f ng th forw.1 of Appendi Bt 0~ f' , riathQasis Ne /j e annual Radioactive . fluen Release Repor e submitte& 6 ' days after a ry l' f ea h yearAh'all includ an annual su ofhovylymetebrological ed o er the previous yea T thethm f n h ub-ty-hour listing ' .agnetic tape o mary nual su ma M i.her in wind speed, w%direc - tion atm sphe ic sT3bRity, and ciDitation (if m asured), or in the fo . of f r t frequenc distributioins of ind spee , di ection, atmospheric / Cabil'ty ** This same repoft shall nclude an ssm nt of the rad ation dee, . to the r joettive liquid a d a effluents released from the unit o s t) i on dur g the previous ndar year. This same report . hall also inclueataisesQen of the radiat oses from radioa,tive li ind gase-ouve" fluents to MEMB LQF THE PUBL c ue- w U=' ti ities inside the41TE B0Q@ RY (Fig es5.1.3lajind5.1.3-lb)duringthe por period. Al assump-tions se n kingJbe assessments (i e. c act*vity, expos t'ime and lo n) sh. W be inc ded in these r orts. The asse smen 4f ra 'ation desq[Fr5iic1 be p4 4c:te}d incordance with e metne sh d pa of. . the u SE TION .A(UAL(00CM). y N
$*A single submittal may be made for a multi unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
1 O LIMERICK - UNIT f. 6-17 Amendment No. 26 M I I IS
ADMINI$TRATIVE CONTROLS
. SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)
The Semiannua Radioactive fluent Release port to be submi ted 60 days fter Ja vary 1 of e ch year shall 1so include an assessment of radi tion doses o the likely most exposed MEMBE OF THE PUBLIC f m reactor releasts and othe near uranium el cycle sour s (including do es from primary e' fluent pat - ways nd direct 'diation) for he previous cal pdar year to show onformance with CFR Part 0, Environmen al Radiation Protection Standards for Nuclea Power eration, ceptablemethdsforcalculathgthedosecontrfutionfro liquid d gaseous ffluents are ven in Regulato Guide 1.109, Re 1, October 77. The S mian ual Radioac ive Effluent elease Reports s all include the f lowing infor. nation for each ty' e of solid wa te (as defined i 10 CFR Part 61) khipped ffsite dur g the repor period':
- a. Con iner volume
\ b. Tota 1\ curie quant 4y (specify other determine by measurement or s estimate, \ -
- t. Principal radionuclikes (specify hether determin by measurement '
\ or estimate),
d.\ Source ok waste and pr cessing empi yed (e.g., dewattred spent resin,
\ compacted ' dry waste, ev orator bott ms),
- e. ' Type of container (e.g., SA Type A, Type B, large Q}u4ntity),
1 and i \
- f. SOLIDIFICATION agent or ab orbent (e.g. cement; urea f'rmaldehyde). o
\
\T Semiannual Radioactive Effluent R ease Report shall include a' list and de ription of unplanned releases from he site to RESTRICTED AREASs of radio-act ye materials in gaseous and liquid (fluents mad during the reporting period.
\ \
\ '
The Semiannual Radioactive Eftluent Release Reports shall include any changes made during the reporting period to the PRO SS CONTROL PROGRAM (PCP) and to the ODCM, as well as a listing o new locatio s for dose c'alculations and/or environmental monitoring identif(ed i by the land use census pursuant to Specification 3.12.2. '
\
4 SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for 6:e report. e
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( ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION
- 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
St.10.2 The following records shall be retained for at least 5 years:
..a. Records and logs of unit operation covering time interval at each
,,. power level. '
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to
. nuciaar safety. '
- c. All REPORTABLE EVENTS.
' d. Records of surveillance activities, inspections, and calibt'ations required by these Technical Specifications. ,
- e. Records of changes made to the procedures required by Specification
- 6.8.1. .
- f. Records of rad'ioactive shipments.
- g. Records of sealed source and fission detector leak tests and results,
- h. Records of annual phy'sical inventory of all sealed source material of record.
- 6.10.3 The following records shall be retained for the duration of the unit Operating License: .
- a. Records and drawing changes reflecting unit' design modifications made '
to systems and equipment described in the Final Safety Analysis _ Report,
- b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories. .
- c. Records of radiation exposure for all individuals entering radiation control areas. .
t l. LIMERICK - UNIT 'I. 6 ,_ . - t l P.T 'C Le a-l' ll . e. . . . . . . ..
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i ADMINISTRATIVE CONTROLS ' RECORD RETENTIE (Continued)
, d. Records of gasecus and liquid radioactive material released to the environs, ,
- e. Records of transient or operational cycles for those unit components identified in Table 5.6.1-1. ,
I
- f. ' Records of reactor tests and experiments. .
- g. Records of training and qualification for curr.ent members of the unit *
. ,- staff. .
- h. Records of inservice inspections performed pursuant to these Technical -
, Specifications.
- i. '
Records of quality assurance activities required by .the' 0perational Quality Assurance Manual not listed in Section,6.10.2.
- j. Records of reviews ment or reviews of, performed for changes made tests and experiments to procedures pursuant o'r equip-to 10 CFR 50.59. '
l*N k. Records of meetings of the PORC and the NRB. . .
- 1. Records of the service lives of all snubbers ineluding the date a't which maintenance the service life commensva ano assosiot:d installation and records,
- m. Records of =a=1 yds required by the Radiological Environmental Honitoring Program that =ct'1d permit evaluation of the accuracy of the analysis at a later date. .. .- .
. . # #. Records of reviews performed n r changes made to the OFFSITE DOSE . ;.. .-
CALCULATION MANUAL and the PROLESS CONTROL PROGRAM. . 2c. , 6.11 RADIATION PROTECTION PROGRAM .
'6.' 11.1 Procedures for perscnnel radiation protection shall be prepared con-sistent with the requirements of 10 CFR Part 20 and shall be approved, main-l tained, and adhered to ,for all operations involving perscnnel radiation exposure. -
62 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 Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /h but less than 1000 mrem /h shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiatien Work Permit (RWP)*. 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.
" Health physics personnel or personnel escorted by health physics personnel shall be er.empt frem the RWP issuance requirement during the perfarmance of
.their assigned rsdiation protection duties, provided they are otherwise foilowing plant radiation protection procedures for entry into high radiction areas.
LIMERICK - UNIT 2 6-20 K gjm . I g g, 4 ggegaggym es e ge OG' 6 850 4 4 ^ 9' .' '
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- 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 J may be made after the dose rate. lev,els in the area have been estab- , I lished and personnel have been made. knowledgeable of them, j
- c. A health physics qualified. individual (i.e., qualified in radiation protection procedures) with a radia
- who is responsible for providing po, tion dose sitive rate control over monitoring the activi- device ties within~the area and shall perform periodic radiation surveil-lance at the frequency specified by the Health Physicist in the RWP.
6.12.2 In addition to the requirements'of Specification 6.12.1, areas accessible -
, to personnel with radiation levels such that ,a major portion of the' body could receive in 1 hour a dose greater than 1000 mrems shall be provided with locked doors to prevent unauthorized entry, and the keys shall b'e maintained under the administrative control of Shift Supervision on duty ahd/or the health physics supervision. Doors shall remain lock'ed except during periods of access by personnel under an approved RWP shich 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 1 hour a dose in excess of 1000 mrems* that are located within large areas, such as the 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 RWP, continuous .
surveillance direct or remote (such as use of closed circuit TV cameras), ray
, be made by personnel qualified in radi.ation protection procedures to provide
-positive exposure control over the activities within the. area.
6.13 PROCESS CONTROL PROGRAM (PCP) Changes to the PCP: l l
- a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.3/. This documentation shall
-(
contain: h} ( 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and wnni---..., we ..+ w e r__-mm o~~- -. _
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" Measurement made at 18 ' inches from source of radioactivity.
,( i LIMERICK - UNIT h 6-21 O l C'M - --- *
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- 2. A-determination that the change did not reduce the overall conformance of the solidified waste product to existing
.ee keri: 1 ::
3' app t J4 fe, tc.ps s, &:;t:;; ;,r,s ety; owl, d F&vJJk3e . Q h0 e s bWut.rG14 wesi~ b. Shallbegm effe on review App r ov e h gi g up % p.'and acceptance by the PORC.cw O b 94 0FFSITE DOSE CALCULATIDN MANUAL (00CR) Changes to the 00CH: , a. Shall be documer.ted and records of reviews performed shall be retain-ed as required by Specification 6.10.3p. This decementation shall contain: A
- 1) Suf /lcient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and .
- 2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR
!' Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and l not edversely impact the accuracy or reliability of effluent, i dose, or setpoint calculations,
- b. .
i Shall become effective after review and accept.ance by the [URG) and the approval of the Plant Manager. c. Shall be submitted to the' Commission in the form of a complete, . legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the 00CM was made. Each change shall : i be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shali - indicate the date (e.g., month / year) the change was implemented, v.t 4 6.s.q.v nn. 6.v mosvr soony sv $n. i .wiv w m .... .i . - . (liquid, gaseb s, and solid):
- a. Shall be ported to the Commission ' the Semiannual Radica tive Effluent Re se Report for the perio n which the change wa made effective. Th discussion of each chang shall contain:
- 1. A summary of th evaluation that led to -
the change could made in accordance wit edeterminationthat\ 10 CFR 50.59;
\
LIMERICK - UNIT $ 6-22 U & fW i l
- 4 0 eph -
46 . ($e fees. . *6 GG sen 9 . 89- - -
9 1
}f4 ADMINISTRATIVE CONTROLS h MAJ0k CHANGES 70 *A010 ACTIVE WASTE TREATMENT SYSTh5 (Continued)I -
\ 2. for(hechangewitho Suf icient detailed nformation to t tally support t e reason benefit of add tional or suppl mental info ation;
- 3. Adeta\leddescription f the equipment, components, an processes involve and the interfaces with other p nt systems;
- 4. An evaluat. ion of the chang which shows.th predicted role es of radica6tive materials in liquid and gase us effluents an or .
quantity ofs solid waste thats differ from tho e previously pre ictedinthelicenseapplicationandamendeehtsthereto;
- 5. evaluationo\fthechangewhichshowstheex\pe.cted meximum exp'osures to individual in the UNRESTRICTED AREA \and to the general population that differ from those previou ly estimated in the license application and amendments thereto,';
y ,
$. A comperison of thes predicted releases of Tadioactive, materials, in liquid and gaseous effluents and in solid waste, to,the actual
\ releases, for the period prior to when the
\
- 7. \ An estimate of the exposure to plant operating personne1 as a s
x changes are tp be made; s
\
' result of.the change; and, '
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- 8. Documentation of the fact that the change was reviewed and found acceptable by'the PORC.
\
- b. Shall be reviewed and accepted by dhe PORC prior toximplementation.
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Q. LIMERICK - UNIT 2 6-23 IdG 2 5 lhi
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OFFSITE DOSE CALCULATION KANUAL
' REVISION 8 I
s- , LIMERICK GENERATING STATION s , UNITS 1 AND 2 i PHILADELPHIA ELECTRIC COMPANY DOCKET NOS. 50-352 and 50-353 i . PORC Approval: , Manayer
![ IO __ ,
G Radiation Control and Chemistty/ Nuclear and Environmental / Section Approval M* Manager
/
LGS Healtn Physics Representative , Mcotr- /d /74064Y Radiation Control and Chemistry / L Nuclear and Environmental /' 4s/r/. // . 4Ab.WA' L so j T' Representatives , c7 l. 1 l l
9 stodem Table of Contents ! Page 1 Introduction Part I"- Radiological Effluent Controls
- I,- l
- 1. 0' Definitions I-7 2.0 Basis for Effluent Monitoring Controls I-7 2.1 Meteorological Monitoring Instrumentation 2.2 Maps defining unrestricted areas and site boundary for I-7 radioactive gaseous and liquid effluents I-7 2 . ,3 Radioactive Liquid Effluent Monitoring Instrument 4 tion I-12 ,
2.3.1 Concentration , I-12 2.3.2 Dose I-13 2.3.3 Liquid Radwaste Treatment System ' I-13 2.4 Radioactive Gaseous Effluent Monitoring Instrumentation I-13 , - 2.4.1 Dose Rate I-14 2.4.2 Dose - Noble Gases 2.4.3 Dose - Iodine-131, Iodine-133, Tritium, and I-14 Radionuclides in Particulate Form I-15 2.4.4 Ventilation Exhaust Trentment System I-16 2.4.5 Venting and Purging . I-16 2.4.6 Total Dose ' I-17 2.5 Monitoring Program I-17 2.5.1 Land Use Census I-17 2.5.2 Interlaboratory Comparison Program I-18 2.6 Applicability I-20 3.0 Effluent Controls I-20 3.1 Meteorological Monitoring Instrumentation i
e I-25 3.2 Liquid Effluents 3.2.1 Radioactive Liquid Effluent Monitoring I-25 Instrumentation I-28
' 3.2.2 Concentration .
s I-32 3.2.3 Dose I-33 3.2,.4 Liquid Radwaste Treatment System - , 3.3 Gaseous Effluents 3.3.1 Radioactive Gaseous Effluent Monitoring I-34 Instrumentation I-41 3.3.2 Dose Rate I-45 3.3.3 Dose - Noble Gasts 3.3.4 Dose - Iodine-131, Iodine-133, Tritium, and I-46 Radionucildea in Particelate Form : I-47 3.3.5 Ventilation Er.haust Treatment System I-48 3.3.6 Venting or Purging I-49 3.3.7 Total Dose 3.4 Radiological Environmental Monitoring I-50 3.4.1 Monitoring Program I-62 3.4.2 Land Use Census I-63 Interlaboratory Comparison Program 3.4.3 I-64 3.5 Annual Radiological Environmental Operating Report I-65 3.6 Semiannual Radioactive Effluent Release Report I-67 3.7 Major Changes to Radioactive Waste Treatment Systems l l 6 li
)
i
LIMERICK GENERATING STATION - OFFSITE DOSE CALCULATION MANUAL ! TABLE OF CONTENTS t
'- II. Calculational Methodologies 1.0 Liquid Ef fluents" 1
1.1' Radiation Monitoring Instrumentation and Controls - , 1 1.2 Liquid Effluent Monitor Setpoint Determination ; 2 r 1.2.1 Radwaste Discharge Monitor 6 1.2.2 Service Water Radiation Monitor 7 1.2.3 RHR Service Water Monitor 8 1.'3 Liquid Effluent Dose Calculations - 10 CFR 50 8 1.3.1 Member of the Public Dose - Liquid Effluents 14 1.4 Liquid Effluent Dose Projection 2.0 Gaseous Effluents 15 2.1 Radiation Monitoring Instrumentation and Controls . 16 2.2 Gaseous Effluent Monitor Setpoint Determination 19 2.2.1 North Vent Noble Gas Effluent Monitors 22 2.2.2 South Vent Noble Gas Effluent Monitors 25 . 2.2.3 Noble Gas Effluent Fractional Contribution , 27-2.2.4 Noble Gas Effluent Default Se; points 31 2.2.5 WRAM Noble Gas Effluent Monitor 38 2 . 3. Gaseous Effluent Dose Evaluations 39 2.3.1 Site Boundary Dose Rate - Noble Gases 2.3.2 Site Boundary Dose Rate - Radiciodine and 43 Particulates 44 2.3.3 Noble Gas Air Dose 2.3.4 Radioiodine and Particulate Dose Calculations 46 10 CFR 50 iii
50 2.3.5 Annual Dose Evaluation 55 - 2.4 Gaseous Effluent Dose Projection , 56 i i 3.0 Special Dose Analysis 56 3.1 Total Doses to Members of the Public 57 3 . 2, Doses Due to Activities Inside the Site Boundary ' 59
'4.0 Radiological Environmental Monitoring Program 59 4.1 Sampling Program 60 4.2 Interlaboratory Comparison Program APPENDICES Appendix A - Technical Specification /0DCM Matrix Appendix B - Radiological Environmental Monitoring Program - Sample Type, Location and Analysis O
iv
LIST OF TABLES I 3.1-1 Meteorological Monitoring Instrumentation : I 3.1-2 Meteorological Monitoring Instrumentation Surveilla.1ce
' '
- Requirements '
l I 3.2-1 Radioactive Liquid Effluent Monitoring Instrumentation I 3'.2-2 Radioactive Liquid Effluent Monitoring Instrumentation
- Surveillance Requirements I 3.2-3 Radioactive Liquid Waste Sampling and Analysis Program I 3.3-1 Radioactive Gaseous Effluent Monitor'ing Instrumentation I 3.3-2 Radioactive Gaseous Effluent Monitoring Instrumentation ,
Surveillance Requirements I'3.3-3 Radioactive Gaseous Waste Sampling and Analysis Program I 3.4-1 Radiological Environmental Monitoring Program I 3.4-2 Reporting Levels for Radioactivity Concentrations In Environmental Samples 13.4-3 Detection capabilities For Envirtsamental Sample Ana~eysis II 1-1 Ingestien Individual Dese Factors (Daipj) For Liquid Effluents II 1-2 Assumptions Used In Limerick Liquid Effluent Dvse
- Evaluations ,
II 2-1 Dose Factors For Noble Gases II 2-2 Individual Dose Factors (DFk,1,n) For Gaseous Effluents II 2-3 Assumptions Used In Limerick Gaseous Effluent Dose Evaluation II 2-4 Nearest Gaseous Effluent Dose Receptor Distances (Meters) A-1 Technical Specification /ODCM Matrix B-1 Radiological Environmental Monitoring Program l i l v I
LIST OF FIGURES I 2.1-1 Meteorological Tower Location l I 2.2-1 Map Defining Unrestricted Areas for Radioactive Gaseous and Liquid Effluents I 2.2-2 Map Defining Unrestricted Areas for Radioactive Gaseous and Liquid Effluents II 1.3.1-1 LGS Dilution vs River Flow (Outfall) II ' 3.1-2 LGS Dilution vs River Flow (Citizen's Utility) II 1.3.1-3 LGS Dilution vs River Flow (Phoenixville) II 1.3.1-4 LGS Dilution vs River Flow (Philadelphia Suburban) II'1.3.1-5 LGS Dilution vs River Flow (Crew Course) t I.T l-1 Liquid Effluent Flow Diagram ,
.II 2-1 Gaseous Effluent Flow Diagram - North Stack :
II 2-2 Gaseous Effluent Flow Diagram - South Stack 4 II 2-3 Gaseous Effluent Flow Diagram - Hot Maintenance Shop u Environmental Sampling Locations on Site or Near B-1 the Limerick Generating Station B-2 Aquatic and Terrestrial Environ:nental Sampling Locations at Intermediate Distances from the Limerick Generating Station B-3 Airborne and TLD Environmental Sampling Locations at Intermediate Dictances from the Limerick Generating Station B-4 Environmental Sampling Locations at Remote Distances from the Limerick Generating Station I l vi 1 I
t !
- n. TABLE A-1 TECHNICAL SPECIFICATION /ODCM MATRIX-CDCM ODCM '
Section I Equation ; 0- Technical- Requirement Section II jpecification i l 3.2.1 1-1 i 6.8.'4.d.1 1-2' . , i 1-3 1-4 l 6.8.4.d.1 3.3.1 2-1 (or 2-9) 2-2 (or 2-9) j 2-4 (or 2-10) 2-5 (or 2-10) l 3.2.2 1-1 6.,8.4.d.2,d.3 1-2 ! 1-3 [
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DEFINITIONS The following terms are defined so Thethat uniform defined interpretation terms' appear in of these controls may be achieved. these capitalized type and shall be applicable throughout , controls. 1.1 ACTION ACTION shall be that part of a control which prescribes remedial measures required under designated conditions. * / t 1.2 CHANNEL CALIBRATION A CRANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the The CHANNEL parameter which the channel monitors. CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall The CRANNEL include the CHANNEL FUNCTIONAL TEST. CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. 1.3 CHANNEL CHECK A CRANNEL CHECK shall be the qualitative assessmentThis of channel behavior during operation by observation. determination shall include, where possible, comparison of
- the channel indication and/or status with other indications and/or status derived from independent instrument channels ,
measuring the same parameter. 1.4 CRANNEL FUNCTIONAL TEST A CRANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips,
- b. Bistabic channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested. - (T Engr.-/cl O l H. P. .ro
.: hs I-l Date Rev. i'
i 1.5 DOSE EQUIVALENT I-131 i DOSE EQUIVALENT I-131 shall be that concentration of I-131_, i microcuries per gram, which alone would produce the sa:ve l thyroid dose as the quantity and isotopic mixture The of I-131, I-132, 1-133, I-134, and I-135 actually present. j thyroid dose conversion factors used for this calculation
*- shall be those listed in Table III of TID-14844, i
" Calculation of Distance Factors for Power and Test Reactor l
Sites."
- 1. 6' FREQUENCY NOTATION '
The FREQUENCY NOTATION specified for the performance of . Surveillance Requirements shall correspond to the intervals > defined in Table 1.1. 1.7 MEMBER (S) OF THE PUBLIC MEMBER (S) OF THE PUBLIC shall include all persons Thiswho are not occupationally associated with the plant. category does not includeAlso employees excludedoffrom thethis utility, its category contractors, or vendors. to service equipment or to are persons who enter the sitThis category does include persons who make deliveries. use portions of the site for recreational, occupational, or other purposes not associated with the plant. . 1.8 OFFSITE DOSE CALCULATION MANUAL
- The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid ,
effluents, in the calculation of gaseous and liquid offluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring ' Program. The ODCM shall also contain (1) the Radioactive Effluents Controls and Radiological Environmental Monitoring Programs required by Technical Specification 6.8.4 and (2) descriptiens of the infprmation that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Relea ; 6.9.1.8. (3) description of meteorological monitoring controls. > Engr.;$ Chb H. P. 4 < 5._ Date w /. / rt Rev. ? I-2
F~ l l.9 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE 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 v equipment that are required for the system, subsystem, are train, component, or device to perform its function (s) also capable'of performing their related support function (s). - i 1.10 OPERATIONAL' CONDITION - CONDITION An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as defined in the Technical Specifications. 1.11 PURGE - PURGING PURGE or PURGING shall be 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.
1.12 P.ATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat
. transfer rate to the reactor coolant of 3293 MWt.
1.13 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CPR Part 50. 1.14 SITE BOUNDARY The SITE BOUNDARY shall be that line ps defined in rigure I2.2-la. 1.15 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source., 1.16 THERMAL POWER transfer THERMAL POWER shall be the total reactor core heat rate to the reactor coolant. Engr.I(/S H. P. JCe%5 I-3 Date 3 /n/f o Rev. Q l l
3.17 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or anyare
.or for industrial, commercial, institutional, and/or recreational purposes. '
1.18 VENTILATION EXRAUST TREATMENT SYSTEM - < A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or , radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through_ charcoal adsorbers and/or HEPA filters for the purpose of removing ' iodines or particulates from the gaseous exhaust stream (such a system is prior to the release to the environmentnot considered to have an Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXRAUST TREATMENT : SYSTEM components. 1.19 VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration orused Vent, otherin operatingco not provided or required during VENTING. system names, does not imply a VENTING process. e Engr.TihIS H. P. Txd5
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v 1-4
TABLE lol SURVEILLANCE FREQUENCY NOTATION I NOTATION PREQUENCY s S At least once,per 12 hours. 3 D At least once per 24 hours. - ; a W At least once per 7 days. M At least once per 31 days. i O At least once per 92 days. SA At least once per 184 days. , At least once per 366 days.
.A R
At least once per 18 months (550 days). S/U Prior to each reactor startup. P Prior to each radioactive release. N.A. Not applicable. 4 O i Engr.dII b
!! . P. w /s Date '5 /; /9e Rev. #
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l - 2.1-METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the' meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to, evaluate the need for initiating protective measures to protect the health and. safety of the public. This instrumentation is contistent with the recommendations of Regulatory Guide 1.23 . ,
"Onsite Meteorological Programs," February,1972.
Site data compiled since January 1972 provide correlation betweenElevation1(lower 1)andElevation1(Tower 2),and ' between Elevation 2 (Tower 1) and Elevation 2 (Tower ?). This -
^ correlation serves as justification for the use of the appropriate Tower 2 instrument as a back-up to the Tower 1 instrument as shown in Table 1 3.1-1.
MAPS DEFINING UNRESTRICTED AREAS AND SITE 80VHDARY FOR RADIOACTIVE GASEQUS AND LIQUID EFFLUENTS 2.2_ Information regarding radioactive gaseous and liquid 1 ef fluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE-BOUNDARY that are accessible to MEMBER OF THE PUBLIC, shall be shown in Figures 12.2.-la and 12.2.-lb. 5 The exclusion area and low population zone shall be as shown in Figures 12.2.-1c and 12.2-ld. e 2.3RADI0ACTIVELIQUIDEFFLUENTMONIT0QNGINSTRUMENTATION The= radioactive liquid effluent instrumentation is provided . to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual i or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the . procedures in the ODCM part 11 to ensure that the alarm / trip wi'.1 - occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this int,trumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of ' Appendix A to 10 CFR Part 50. , e Engr. 4 d:b ! H. P. ,.M Date 7 770 , Rev. _F 1-7
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- l2.3,1 CONCENTRATION ,
This control is provided to ensure that the concentration of radio' active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B. Table !!, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will res exposures within (1) the Section ll.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to < the population. The. concentration limits for dissolved or entrained noble gases are based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLO, and other detection limitsA.,can be found
" Limits for in HASL
' Procedures Manual, HASL-300 (revised annually), Currie, L. Qualitative Detection and Quantitative Determination - Application "toand Hartwell, J. K. Detection Radiochemistry," Anal. Chem. 40. 586-93 (1968), , Limits for Radioanalytical Countina Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June,1975). 1 2.3.2 DOSE This Control is provided to implement the requirements of Sectionsthe The Control implements II.A. ' III.A. and IV.A of Appendix I, 10 CFR Part 50. The ACTION statements provide guides set forth in Section II.A of Appendix 1.the required operating f forth in Section IV.A of Appendix ! to assure that the releases of radioactive , material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the- , operation of the facility will not result in radionuclide concentrations in the : finished drinking water that are in excess of the requirements of.40 CFR Part 141. The dose calculation methodology and parameters in the 00CM implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data, s that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 -r CFR Part 50, Appendix 1," Revision 1, October, 1977 and Regulatory Guice 1.113
" Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor I," April, 1977.
Releases for the Purpose of Implementing Appendix This control applies to the release of radioactive materials in liquid effluents from the site. Engr. 'd ' H. P. e *C5 te _d E' l-12
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23l3 LIQUID RA0 WASTE TREATMENT SYSTEM 7 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". This specification-implements the requirements of 10 CFR 50.35a, General Design-Criterion 60 of Appendix A to 10 CFR Part 50 and the-design objective given in Section 11.0 of Appendix ! to 10 CFR Part 50. The specified-limits governing the use of appropriate portio n of the liquid radwaste treatment system were specified.as a suitable fraction of the r;ose design objectives set forth in Section
'll.A of Appendix 1, 10 CFR Part 50, for liquid effluents. , ;
2.4' RADI0 ACTIVE GASE0US EFFLUENT MONITCAING INSTRUMENTATION The radioactive gaseous effluent inttrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CM to ensure that the-alarm / trip will occur prior'to exceeding the limits of 10 CFR Part 20. . The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and_64 of Appendix A to 10 CFR Part 50. 2.4.1 DOSE RATE This' control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the dose associated with
.the concentrations of 10 CFR Part 20, Appendix B. Table II, Column I. These limits
" ovide reasonable assurance that radioactive material discharged in gaseous-effluents 11- not_ result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to-annual average concentrations exceeding the limits .
;specified in Appendix B. Table 11 of 10 CFR Part 20 (10 CFR 20.106(b)(1)). For MEMBERS OF THE PUBLIC who may at times-be within the SITE BOUNDARY, the occupancy factor for that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in Examples of
-the' atmospheric diffusion factor for above that for the SITE BOUNDARY.
calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given-in the ODCM. The specified release rate limits restrict .at all times, the
. corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE B0UNDARY to less than or equal to 500 mre,ms/ year to the total body or to-less than.or equal-to 3000 mrems/ year to the skin.- These release rate limits also i restrict, at all times, the corresponding thyroid dose rate above background to a' child via the inhalation pathway to less than or equal to 1500 mrems/ year.
This specification applies to the release of radioactive materials in gaseous
. effluents from all reactors at the site.
Engr. H. P./A> jfM Date A h/fc Rev. W I-13
s;Pg B J/4 61-Jooa m - r.au .)
. I DOSE RATE (Continued).
The required detection capability for radioactive materials in gaseous waste samples Detailed discussion of the are tabulated in terms of-the lower limits of detection (LL0s). LLO, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised
- annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination K.,
- Application to Radiochemistry," Anal. Chem. _40, 5B6-93 (1968), and Hartwell, J.
" Detection Limits for Radionnalytical Counting Techniques." Atlantic Richfield Hanford Company Report (ARH-SA-215 (June 1975). . ,
)
2.4.2 OOSE - NOBLE GASES This control 15 provided to implement the requirements of Sections 11.8. Ill.A, and IV.A of Appendir, i, 10 CFR Part 50. The control implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required opera;ing flexibility and at the same time implement the guides set forth in Section IV.A of Appendix ! to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section Ill.A of Appendix ! that conformance with the guides of Appendix 1 be shoWn by calculational precedures based on appropriate pathways is unlikely to be substantially underestimated. The dose calculation established in the 00CM for calcula:ing the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the I," Revision 1, October Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routing Releases from Light-Water Cooled Reactors,"
Revision 1, July 1977 with site specific dispersion curves and deposition methodology. The 00CM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric cor.ditions. 2.4.3 DOSE - 100lNE-131,100lNE-133, TRITIUM. AND RADIONUCLIOES IN PARTICULATE TDRM This control is provided to implement the requirements of Sections II.C, !!!.A, and IV.A of Appendix 1, 10 CFR Part 50. The controls are the guides set forth in Section !!.C of Appendix 1. The ACTION statements provide the required operating flexibility and at the same tirle implement the guides set forth in Sectioq IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The 00CM calculational methods specified in the Surveillance Requirements implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways Engr. G d H . P . 4( S Date .uNO Rev. ." l-14 l l l
m *8 d l
.00SE-100lNE-131, 100 lNE-133. TRITIUM, AND RADIONUCLIOES IN PARTICULATE ;
FORM (Continued) l is unlikely to be sabstantially underestimated. The 00CM calculafional methods for ! calculating the doses due to the actual release rates of the subject materials are l consistent with the methodology provided in Regulatory Guide 1.109, " Calculationl Doses to Man from Routine Releases of Reactor Revision I,Effluents for the October 1977 andPurpose RegulatoryofGuide Evaluating Compliance with 10 CFR Part 50, Appendix 1," 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efi in Routine Releases from Light-Water-Cooled Reactors," Revision 1. July 1977 [w specific dispersion curves and deposition methodology. [ determining the actual doses based upon the historical average atmospheric conditions The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in i particulate form with half-lives greater than 8 days are dependent Theon the existing pathways which j radionuclide pathways to man in areas at and beyond the SITE B0VHDARY.(1) individua were examined in the development of these calculations were: l airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation wit ' subsequent consumption by man, (3) deposition onto grasiy areas where milk animals meat-producing animals graze with consumption of the milk and meat by man, and (4 deposition on the ground with subsequent exposure of man. ! 2 . 4 . 4, VENTILATION EXHAUST TREATMENT SYSTEM f The requirement that the appropriate portions of this system be used, when specified provides reasonable assurance that the releases of radioactive materials in gaseousThis : effluents will be kept "as low as is reasonable achievable." ! requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix ATheto 10 C 50, and the design objectives given in Section 11.0 of Appendix 1 to 10 CFR Part 5j Specified limits governing the use of appropriate portions of the systems were sp ; F . as:a suitable fraction of the dose design objectives set forth in Section !!.B and ll.C o Appendix 1, 10 CFR Part 50, for gaseous effluents. 2.4.5 VENTING OR PURGING : This control provides reasonable assurance that releases from drywell purging . operations will not exceed the annual dose limits of 10 CFR Part 20 for UNRESTR ' AREAS. ' 2.4.6 TOTAL 00SE i TPis control is provided to meet the dose limitationsThe of 40 CFRrequires control Part 190the tnat have the J been incorporated into 10 CFR Part 20 by 46 CFR 18525. preparation and submittal of a Special Report ahenever the calculated doses from For sites radioactive effluents exceed twice the design objective dose OF THE PUBLIC will exceed the dose' limits of 40 CFR Part 190 if the remain within the reporting requirement level. Engr.- d : H . P . - An Odte 3 / A'70 aev. e: 1-15 4
-m , . , , .
^w:
j , 2 4.6 TOTALDOSE~(Continued) The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within z For the purposes of the Special Report, it the 40 CFR Part 190 limits. may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from. other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facil,ities at the If the dose same to site or within a radius of 5 miles must be considered. any' MEMBER OF THE PUBLIC is estimated to exceed the require , the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is
- completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
2.5 RA010 LOGICAL ENVIROMENTAL MONITORING PROGRAM . The raciological erivironmental monitoring program required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposuresThis of MEMBER OF THE PUBLIC resulting from the station operation. monitoring program implements Section IV.B.2 of Appendix 1 to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifyirs that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The'LL0s required by Table 4.12.1-1 are considered optimum for routine environmental measurements in industrial laboratories, it should be recognized that the LLO is defined as an a_ priori (before the fact) limit representing the capability of a measurement system and not an a posteriori_ (after the fact) limit for a particular measurement. Detailed discussion of the LLO, and other detection limits, can be found in HASL Procedures Manual HASL-300 (revised annually); Currie L. A., " Limits for Qualitative Detection and 40, Quantitative 586-93'(1968);Determination and - Application to Radiochemistry" Anal. Chem. Hartwell, J. K., " Detection Limits for Radicanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). Engr.C'M H.P. @5
"' W 4*
l 1
- - - . _. ' - -in ;.,w,,,
- b. s;Pg B 3/4 12-lodt74 - rick _ l 1
L
- 2.5.1 LAND USE CENSUS F This control is provided to ensure that changes in the use of areas at and beyond the SITE' BOUNDARY are identified and that modifications to the monitoring
=
The best information
' program _are made if required by the results of this census.from the door-to-d authorities or any combination of these methods 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 exoosure pathweis via leafy vegetables will be-identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: . (1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/ square meter. 2.5.2 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 environmental monitoring in order to demonstrate that the results are reasonably valid for the purpose of S'.ction IV.B.2 of Appendix I to 10 CFR Part 50. Engr.Gt(;- H . P . eks
~
l l-17 1
-2.6 APPLICABILITY #
2.6.1 Compliance with the controls contained in the succeeding controls section is required during the OPERATIONAL CONDITIONS or other conditions-specified therein; except that upon failure to meet the controls , the associated ACTION t requirements shall be. met, 2.6.2 Noncompliance w'ith a control shall exist when the reauirements'of the control and associated ACTION requirements are not met within the specified time intervals. If'the control is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 2.6.3 There are no actions-in the 00CM which would require an operational condition chance. 2.6.4 There are no restrictions on changing operating conditions in any of the ; controls of the ODCM. i The associated bases from the LGS Technical Specifications applies to this section E ngr .-~Tb*k H. P. /;f3. v I I-18
. i a
A 4 , . APPLICABILITY SURVEILLANCE REQUIREMENTS 2.6.5 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual controls unless otherwise stated in an individual Surveillance Requirement. 2.6.6- Each Surveillance Requirement shall be perforti.ed within the specified time ; interval.with: i
'a. ~A maximum allowable extension not to exceed 25% of the surveillance ,
interval, but i
- b. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times.the specified surveillance interval.
2.6.7 Failure to perform a Surveillance Requirement within the allowed surveillance interval defined by control 2.6.6, shall constitute noncompliance l with the OPERABILITY requirements for a control. The time limits of the ACTION l' requirements are applicable at the time it is identified that a SurveillanceTh.; Rsquirement has not been performed. , to 24 hours to permit the completion of the surveillance when the allowable outage Surveillance i time limits of the ACTION requirements are less than 24 hours. .l requirements do not have to be performed on inoperable equipment. Engr H. P / IDate # Rev. < l i 1-19 f 1 i
s;Pg 3/4 3-73/odg74 - RWG
-METEOROLOGICAL MONITORING INSTRUMENTATION y
CONTROLS' _
*- 3.1 The meteorological monitoring instrumentation channels shown-in Table ,
13.1-1 shall be OPERABLE., APPllCABILITY: At all times.
' ACTION:
- a. With one or more meteorological _ monitoring instrumentation channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to LGS Technical-Specification 6.9.2 within.the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status.
4
-SURVEILLANCE REQUIREMENTS l
3.1.1 Eachoof the above required meteorological monitoring instrumentation ! channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CAllBRATION operations at the frequencies shown in Table L3.1-2. I i f Engr,s35455 H. P. fewes. Date 3/;g/fo Rev. ti ' n I-20 i i I 1
.,,,y.,. ... ,..
f7 ~ ' . E TABLE 13.1-1 , i u
' ": METEOROLOGICAL MONITORING INSTRUMENTATION t
MINIMUM. Tower. 2 INSTRUMENTS Tower 1 (Primary) (Backup) OPERABLE ; INSTRUMENT Wind Spe5d 1. 1,
'a. Elevation-1 30 feet or' 159 feet -
-1
- b. Elevation 2 175 feet ~ or 304 feet 2
- 2. Wind Direction
- a. Elevation 1 30 feet or 159 feet l' 1
Elevation 2 175 feet or 304 feet Y
'.b.
- 3. Air Temperature Difference 266 feet- 300 feet-
- a. Elevations 26 feet 1 26 feet or' ;
NOTE: i
-The meteorological towers shall be located as shown on Figure 13.1-1. i 1
I 1 ' Engr.31['A$
' H . . P . Jt.L i 'Date 1?//y0 Rev. '/
l-21 I:
ii._ - -~24 > p!l f SjPg.J/4-J-/D/Ouise - nnu , TABLE'I3.1-2 , m
. METEOROLOGICAL MONITOR l _NG INSTRUMENTATION SURVEILLANCE REQU ,
4 CHANNEL CHANNEL CHECK ~ CAllBRATION INSTRUMENT .
- 1. Wind-Speed !
D . SA ,
- a. Elevation 1 (Tower 1 and Tower 2)
SA I b. Elevation 2(Tower 1landTower2) D
. I
- 2. Wind Direction D SA-a.. Elevation 1-(Tower 1andTower2)
D SA
- b. Elevation 2 (Tower.1 and-Tower 2) >
I ' 3.. Air Temperature Difference I D SA 1
- a. Elevations 266 - 26 ft (Tower 1)
D SA
- b. ' Elevations 300 - 26 ft (Tower 2) l 4
l i Engr.ep(dkf H. , P . ?r:% Date 3/40 Rev. p t 1-22 ll l> 1 4 l
- stPg 5-6/odem:74 - kth -
- %\
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': g.yg- -. TOWER #1 .,*,.
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FIGURE 13.1-1
' METEOROLOGICAL TOWER LOCATION ,4,p H . P . Ex'$
Date 3/./YC Rev. 3 1-23
,4.... ,
y , RAD 10ACTJVE LIQUID EFFLUENT MONITORING INSTRUMENTATION' e . . j CONTROLS '
.. =3.2.1 l'n eccordance with LGS TS 6.8.4.d.1. the radioactive liquid effluent
. monitoring instrumentation channels shown in Table 13.2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Control 3.2.2 are not exceeded. The alarm / trip setpoints* of these channels shall be determined and adjusted'in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).
APPLICA8ILITY: At all times. t ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel' alarm / trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid
' effluents monitored by the affected channel, or declare the channel inoperable. ., l
-3 '
- b. With less than the minimum number of radioactive liquid effluent -
monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 13.2-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain in the next Semiannual Radioactive Effluent Release Report why this inoperability w as not corrected within the time specified.
'I SURVEILLANCE REQUIREMENTS i
3.2.1.1 Each radioactive liquid effluent moni'toring instrumentation channel shall be demonstrated OPERABLE by performance of.the CHANNEL CHECX, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 13.2-2. ,
'* Excluding the flow rate measuring devices which are not determined and adjusted in accordance.with,the ODCM.
Engr.7 H. P. M , Date z/zij_fo Rev. A A, 1-24
RADIDACTIVELIQUIDEFbLbT55hif0RINGINSTRUMENTATION MINIMUM ; CHANNELS ACT10N :I INSTRUMENT _ OPERABLE.
- 1. GROSS RADIOACTIVITY-MONITORS PROVIDING '
AUTOMATIC TERMlHATION OF RELEASE Liquid Radwaste Effluent Line 1 100
. . a.
- b. RHR Service Water System Effluent tin'e 1/ loop 101 .
-2. ' GROSS RADIOACTIVITY MONITORS NOT PROVIDlHG AUTOMATIC TERMINATION OF RELEASE Service-Water System Effluent Line 1 101 a.
- 3. FLOW RATE MEASUREMENT DEVICES Liquid Radwaste Effluent Line 1 102 a.
1 102
- b. Discharge Line
; -ACTION STATEMENTS ACTION 100 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE' requirement, effluent releases may continue for up to 14 days provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Table 13.2-3, and
- b. At least two technically qualified members.of the facility-
' staff independently verify the release rate calculations and discharge line valving; y Otherwise, suspend release of radioactive effluents via-this pathway. ACTION I0l '-- With the number of channels OPERABLE'less-than required by the Minimum Channels OPERABLE requirement, effluent releases via this , pathway may continue for up te 30 days provided that, at least once per 8 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma). Beta is analyzed at a limit i of detection of at least IN7 microcurie /mL. Gamma is analyzed at. a limit of detec, tion of at least SN7 microcurie /mL. ACTION 102 - 'With the number of channels OPERABLE less thdn required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for'up to 30 days provided the flow rateEamp is estimated at least once per 4 hours during-actual releases. curves generated in situ may be used to estimate flow. Engr. 1 %' H. P. LOl l-25 Date 3/:/Y u ' Rev. +~
~
~s_..,
_ . ,, ~ s;Pg 3/4 3-101/odi74 - ndG TABLL 13.2 ,
~
RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT CHANNEL SOURCE CHANNEL FUNCTIONAL. CHANNEL CHECK CAllBRATION TEST CilECK INSTRUMENT
- 1. GROSS RADI0 ACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELEASE P P R(3) Q(1) ,..
- a. Liquid Radwaste Effluent Line ~
.0 M R(3) Q(1)
- b. RHR Service Mater System Effluent Line
- 2. GROSS RADI0 ACTIVITY MONITORS NOT PROVIDING AUTOMATIC TERMINATION OF PELEASE M R(3) Q(2)
- a. Service Water System Effluent Line D
- 3. FLOW RATE MEASUREMENT DEVICES N.A. R Q
- a. Liquid Radwaste Effluent Line D(4)
- N.A. R Q
- b. Discharge Line .D(4) ,
l t l
-s Engr.
H.'P. /st . Date 4/i/fo ) Rev. SV 1 .- - . . . - . . , . . ., . . , . -. ., ,. . . - - , . - - - . . _ . , , - - - .
~
V_ , , i 5 s TABLE ' 13. 2-2 ' (Cont in'ved) - . j TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic
-isolation of this pathway and control room alarm annunciatio,n occur if any
,of the following conditions exists:
~
?
-1. Instrument indicates measured. levels above the alarm / trip setpoin't.
- 2. Circuit failure.
- 3. -Instrument indicates a downscale failure.
i (2) The. CHANNEL FUNCTIONAL TEST shall also demonstrate that control room al-i 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.
The initial CHANNEL CALIBRATION shall be performed using one or more of
, (3) the reference standards certified by the National Institute of Standards '
and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. .l These standards shall permit calibrating the system over its intended
~
For subsequent CHANNEL range of energy and measurement range. CALIBRATION, sources that have been relhted to the initial-calibration shall be used. CHANNEL CHECK shall consist of verifying indication of flow during periods f (4) of release. CHANNEL CHECK shall be made at least once per 24 hours-on.
' days on.which continuous, periodic, or batch releases are made, l
p Engr. H. P. AA'7-et Date ::,fi/9c' 1-27 ,/ Rev.
\
~
i sI Pg 3/4 11-1/o0 /4 - Kwu , t CONCENTR4 TION- ^
- t CONTROLS 3.2.2 :n accordance with LGS TS_6.8.4.d.2 and 3, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see figure 12.2-la,b) shall be limited to the concentrations specifiedLin 10 CFR Part~
20, Appendix B. Table II, Column 2 for radionuclides other than dissolved or entrained nobles gases. Fo ; shallbelimitedto2x10gdissolvedorentrainednoblegases,.theconcentrat microcuries/mi total activity. - , f 1 APPLICABILITY: At all times. t ACTION: With-the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits. . SURVEILLANCE REQUIREMEN15 3.2.2.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 13.2-3. 3.2.2.2 .The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of' Control 3.2.2. , i i a e
-Date 3 '/90
'Rev.
l' , I-28 4 t
..,,,m..-
c TABLE 13.2-3 , RADIOACTIVE L10010' WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM lYPE OF 0FDETEC{l0N l l SAMPLING ANALYSIS ACTIVITY (LLO) LIQUID RELEASE , ANALYSIS (uCi/ML) i TYPE FREQUENCY FREQUENCY .. j P Principa} Gamma 5x10-7 A.s Batch Waste P Emitters Release Each Batch Each Batch j b Tanks 1-131 1x10-6 M Dissolved and lx10-5
'1. Floor Drain P Entrained Gases. !
Sample Tank One Batch /M No. 2 (Gamma: Emitters) _l i M H-3 1x10-5
- 2. Laundry P Drain Sample Each Batch Composite d 3 Tank Gr6ss Alpha 1x10-7 f P .Q Sr-89, Sr-90 5x10-8 f d
Each Batch Composite Fe-55 1x10-6 W Principa 5x10-7 ;
-B. W Continuogs Release Grab Sample Emitters} Gamma ;
l l-131 lx10-6 W Dissolved and 1x10-6
- 1. RHR' Service W Water System Grab Sample ~ Entrained Gases f (GammaEmitters) ,
Effluent Line , M H 1x10-5 l
- 2. Service Water W System Grab Sample Composite d l Gross Alpha lx10-7 Effluent Line f j
Q Sr-89, Sr-90 5x10-8 W d I Grab aamp'le Composite
- Fe-55 lx10-6 Engr.4dlS H . P . ' /& j Date :4hBC' Rev. /
l-29 0 1
' TABLE 13.2 (Crntinued) '
- TABLE NOTATIONS _
aThe LLO is defined, for purposes of these controls, as the smallest . concentration of radioactive material in a sample that will yield .a net count, above system background, that will' be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" 4
.. . signal.
For a'particular measurement system, which may include radiochemical separation: - , l, 4.66s "~ L LO = """""""b""".
"y 6~,"e~x~p~(~- A A t )
E
- V ' 2.22 x 10 Where: '
LLO is the a priori lower limit of detection as defined above (as microcuries per unit mass or volume), is the standard deviation of the background counting rate or of the sb counting rate of a blank sample as appropriate (as counts per minute). E is the-counting efficiency, as counts per disintegration, V is the sample size, in units of mass or volume, 2.22 x 10 6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, l X is the radioactive decay constant for the particular radionuclide, and A' t for the plant effluents is the elapsed time between the midpoint of l ", sample collection and' time of counting. Typical values of. E V, Y, and at should be used in calculation. It should be recognized that the LLO is defined as' an a priori (before the fact) limit representing the capability of a measurement. system and not as an a-posteriori (after the fact)-limit for a particular measurement. Engr.4V H. P. 46 Date 3/dc I-30 ' Rev. 4
s;Pg 3/4 11-4/od:74 - RWG. TABLE 13.2-3 (Continuedl ' TABLE NOTATIONS b A batch release is the discharge of liquid wastes of a discrete volume.
-Prior to sampling for analyses, each batch shall be isolated, and then l thoroughly mixed to assure representative sampling.
^ /
c The principal gamma emitters for which the LLD specification app the following radionuclides: This list does not mean that only these nuclides Cs-137, Ce-141, and Ce-144.Other gamma peaks that are identifiable .together with are to be considered. those of the above nuclides, shall also be analy2ed and reported in the j Semiannual Radioactive Effluent Release Report pursuant to. Control 3.6. d A composite sample is one in which the quantity of liquid sampled is, ' proportional to the quantity of liquid waste discharged released. 8 A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release, f Whenever effluent releases are in excess of the monitor's setpoint. , I 1 Engr. Gb H . P . )WN Date -s he/o ' 1-31 Rev. V
E ~x s;Pg 3/4 11-5/odt74 s . 0051 CONTROLS !
.- 3.2.3 . In accordance with LGS TS 6.8.4.d 4 and 5 the dose or dose commitm to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from-the site to UNRESTRICTED AREAS (See Figure 12.2-la) shall be limited:
'a.- During any calendar quarter to less than or equal ~to 3 mrems te the total body and to less than or equal to 10 mrems to any organ, and l
- b. -During any' calendar year to less than or equal to 6 mrems to the total j body and to less than or equal to 20 mrems 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, prepare and submit to the Commission within 30 days, pursuant to Technical- Specification 6.9.2,
-3 a Special Report which identifies the cause(s).for exceeding the limit (s) -
I and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions.to be taken to RSsureThisthat subsequent releases will be in compliance with the above limits. i Special Report'shall also include the radiological impact on finished . l drinking water supplies at the nearest downstream drinking water source. 3 l 1 SURVEILLANCE REQUIREMENTS
)
i 3.2.3.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance l with the methodology and parameters in the ODCM at feast once per 31 days.
'l,
- -l Engr. v H . P . /a ';
'1 28 U Date l-32 pey, l
l
----------___-___a
s;Pg 3/4 11-6/odt74 - RWG LIOUID RADWASTE TREATMENT SYSTEM CONTROLS h
- 3.r.4 In accordance with LGS TS 6.8.4.d.6 the liquid radwaste treatment ,
system shall be OPERABLE and appropriate portions of the system shall be used t reduce the radioactive materials in liquid waste prinr to their discharge when the l projected doses due to the liquid effluent, from the site, to UNRESTRICTED AR -- (se'e Figure 12.2-la) would exceed 0.06 mrem to the total body or 0.2 mrem to any i organ in a 31-day period. APPLICABILITY: At all times. ACTION: a. With radioactive liquid waste being discharged without treatment and in ' excess of the above limits, prepare and submit to the Commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report which includes the following information:
; i
- 1. Explanation of what liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, .;
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE Status, and
-3. Summary description of action (s) taken to prevent a recurrence.
SURVEILLANCE REQUIREMENTS i Doses due to liquid releases from the site to UNRESTRICTED AREAS sha i 3.2.'4.1 be projected at least once per 31 days in accordance with the methodology al
. l parameters in the ODCM.
The liquid radwaste treatment system shall be demonstrated OPERABLE b 4.2.4.2 meeting Controls 3.2.2 and 3.2.3. 4 Engr. M b H . P . N<A v Date gfo I-33 Rev. 1
s;Pg 3/8 3-103/od874 - R4G-b . RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION' s H CONTROLS.
'1 3.3.1 In accordance with LGS TS 6.8.4.d.1, the radioactive gaseous effluent
- monitoring instrumentation channels shown in Table 13.3-1 shall be OPERABLE with ;
;their alarm / trip setpoints set to ensure that. the limits of Control 3.3.2 are not exceeded.. The alarm / trip setpoints of the applicable channels shall be 3 determined'in accordance with the methodology and parameters in the 00CM.
- < l APPLICABILITY: As shown in Table 13.3-1 ACTION: ,
i
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the atove Control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel !'
inoperable. ! Y
.b. With less than the minimum number of radioactive gaseous effluent-monitoring instrumentation channels OPERABLE, take the ACTION shown in t
' Table 13.3-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Semiannual Radioactive Effluent Release' Report.
l
-SURVEILLANCE REQUIREMENTS-3.3.1.1' Each. radioactive gaseous effluent mohitoring instrumentation channel ,
shall be demonstrated OPERABLE by performance of the CHANNEL CHECK. SOURCE
. CHECK, CHANNEL CAllBRATION, and CHANNEL FUNCTIONAL TEST operations at the L frequencies.shown in Table 13.3-2.
s' Engr."TV)b. H.P.ggr$; < Date snv1fc.' ' I-34 Rev. _ji__ l l l , I t l l
, , , . . . - . . . . . - - - . . - . . . - . . .~ ~ .
- - . ~ . . ~ . ,. .. ~ . . - - ~ . - . . . ~.~ - -.
( , s;Pg 3/4 3-104/od:/4 - R W
~
' 1ABLE 13.3 l l-i-
- .RADIDACTIVE GASEOUS EFFLUENT MONITORIE iWSTRUMENTATION I
MINIMUM CHANNELS APPLICABILITY ACi10N OPERASLE INSTRUMENT ,
- 1. 50u!H STACK EFFLUENT MONITORING SYSTEM
- 111 1
- a. Moble Gas Activity Monitor ,
- 112 todine Sampler I b.
1
- 112'
- c. Particulate Sampler
- 113 1
- d. Effluent System flow Rate Montter
- 113 I
- e. Sampler flow Rate Monitor +
- 2. NORTH STACK LFFLUENT MONITORING SYSTEM
- 114 1
- a. Nchle Gas Activity 'rionitor
-
- 112 1
- b. Iodine Sampler 'l
'* 112 1
- c. ' Particulate Sampler
- 21.2
" Effluent System flos Rate Monitor I d.
-*' 113-1
.e. Sampler flow Rate Monitor ,
.=.
C : Engr. .--
. H. P. fr D> ? -
I-35 f _hi
, . w - .
5 I", - _ - _3 '
. ; e. "
+#,
- 3. ] C.,y':J
,2 *
~
~
s;Pg 3/4 3-105/od:74 - RWG h TABLE 13.3-1 (continued) s : P8@l0 ACTIVE GASEOUS EFFLUENT MONITORING INSIRUMENTATION MINIMUM CHANNELS APPLICABILITY ACT10t3
-OPERABLE-INSTRUMENT ,
1
- 3. HOT MAINTEilANCE' SHOP VENT!tATION EXHAUST RADIATION MON 110R *
* ** 112 1
- a. Iodine Sampler _
** 112 1
- b. Particulate Sampler
" 113 1
- c. Effluent System flow Rate Monitor
** 113 1
- d. Sampler flow Rate'Honitor 89 l
l
' Engr.[W H. P. g C. .
Date 3 /. //0 I-36 Rev. //- I! !?dn j j .- e , ver - , .*
~
e w,.'
steg he a-Avofvws. - nn., TABLE 13.3-_,(Continued) 1 TABLE NOTATIONS i
- At all times, s
** During operation of the hot maintenance shop ventilation exhaust system.
i ACTION STATEMENTS ACTION 111 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for , gross activity within 24 hours. With the number of channels OPERABLE less than required by the AC,110N 112 - Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 13.3-3 ACTION 113 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requiremett, effluent releases via this pathway may continue for up to 30 d..ys provided the flow rate is estimated at least once per 4 hours ACTION 114 - With the number of channels OPERABtE less than required by the Minimum Channels OPERABLE requiremint, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours and provided the mechanical vacuum pumps are not operated. Engr.@b H. P. Wh I-37 fc' l
s;Pg 3/4 3-107/od:74 - RWG TABLE 13.3-2 j
- RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
; CHANNEL MODES IN WHICH SOURCE CHANNEL FUNCTIONAL SURVEILLANCE CHANNEL CHECK CHECK CALIBRATION TESI IS RFQUIRED
.NSTRUMENT
- 1. SOUTH STACK EFFLUENT MONITORING SYSTEM
- a. Noble Gas Activity Monitor D M R(2) Q(1)
Iodine Sampler W(3) N.A. N.A. N.A.
. b.
N.A. N.A. N.A.
- c. Particulate Sampler W(3) r
- d. Effluent System flow Rate Monitor B N.A. -R Q
- e. Sampler flow Rate Monitor 0 N.A. R Q
?. NORTH STACK EFFLUENT MONITORING SYSTEM
- a. Noble Gas Activity Monitor D M R(2) Q(1)
N.A. N.A. N.A.
- b. Iodine, Sampler W(3)
Particulate Sampler N.A. N.A. N.A.
- c. W(3)
- d. Effluent System flow Rate Monitor D N.A. R Q
- e. Sampler flow Rate Monitor D N.A. R Q Engr.Ivd H . P . L' C 1-38 Date J, 'O Rev. _ < _
G- - . . - - . _ . . - ~ _ -
M s;Pg 3/4 3-108/od:74 - RWG 1ABLE 13.3-2 (Continued) RADI9 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURV
, CHAtlNEL MODES Ill WHICH TUNCTI0tlAL SUPVEILLAriCE SOURCE CilANNEL IS PEQUIRED CHANNEL TEST CHECK CALIBRATION CHECK INSTRUMENT
- 3. HOT MAINTENANCE SHOP VENTILATION EXHAUST RADIATION MONITOR "
N.A. N.A. W(3) N.A.
- a. Iodine Sampler "
N.A. N.A. W(3) N.A.
. *b. Particulate Sampler
- R Q D N.A.
- c. Effluent System flow Rate Monitor "
R Q D N.A.
- d. Sampler flow Rate Monitor Engr. . 7. t[F H. P. _.
- Date I-39 Rev. _
? e
slPg 3/4 3-109/od:74 - RWG TABLE 13.33 (Continued)' TABLE NOTATIONS
- At all times. .
** During operation of the hot maintenance shop ventilation exhaust system.
(1)' The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room annunciation occurs if any of the following conditions exists:-
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale f ailure.
4 Instrument controls not set in operate mode. The initial CHANNEL CAllBRATION shall be performed using one or more of (2) the reference standards certified by the National Institute of Standards and Testing (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with HIST. These standards shall permit calibratingFor thesubsequent system over its intended CHANNEL range of energy and measurement range.CAllBRATION, sources that have been shall be used. The iodine cartridges and particulate filters will be changed at least (3) once per 7 days. Engr.Ii[ih U H.P. Kro I 40, Date XQCe Rev.
s;Pg 3/4 11 8/od 74 - RWG > GASEOUS EFFLUENTS j DOSE RATE i CONTROLS 3.3.2 In accorda'nce with LGS TS 6.8.4.d.7 the dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the , SITE BOUNDARY (see Figure 12.2-la) shall be limited to the following4
- a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and
- b. For iodine-131, for iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 dayst less than or equal to 1500 mrems/yr to any organ. (Inhalationpathwaysonly.)
APPLICABILITY: At all times. ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limits. i SURVEILLANCE REQUIREMENTS 3.3.2.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of the ODCM.
3.3.2.2 The dose rate due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous I effluents shall be determined to be within the above limits in accordance with the methodology and parameters of the ODCM by obtaining, representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 13.3-3. i i ' I-41 Date sufffo Rev. S'
s;Pg 3/4 11-9/oua . - - l IABLE 13.3-3 RAD 10ACTIVC GASEOUS WAS1E SMPt.ING AND ANALYSIS PROGRM DEltC110N (LLD)a ANALYSIS INE Of (mCl/mL) SAMPLIllG ACIIV11Y ANALYSIS FREQUEDICY FREQUENCY GASEOUS RELEASE TYPE , P P
?rincipal Gamuna Emitters **I 1 x 10-4 Each Purge Each Purge A. Contairment Purge Grab (Pretreatment) '
Sample t Principal Gasuna Emitters
- 1x10-4 gb Ilorth Stack and Mb B.
South Stack - Crab -6 1x10
- Sample H-3 Ix10-I2 '
! d yc I-131 1x10-10 Not Maintenance Continuous I-133 C. Charcoal Shop Ventilation 57 1e Exhaust 9 and All lx10-II Release Types d yc Principal Gamma Emitters
- Continuous (1-131,Others) gedin8 Particulate Sample _
Gross Alpha lx10-II d Q Continuous Composite Par-ticulate Sample - Sr-M, Sr-M 1x10~II d Q Continuous Composite Par-ticulate Sample lx10-6 d Noble Gas lloble Gases All Release Types Coatinuous Gross Beta or Gamuna (Based on D. Monitor Xe-133) Listed in B above ~ 4 Engr. d t?$ H. P. TikG Date y fa
' ' I-42 e
i 3;Pg 3/4 11-10/od:74 - RWG TABLE 13.3-3 (Continued) . TABLE NOTATIONS , aihe LLD is defined, for purposes of these Controls, as the smallest concentration of radioactive material in a sample that will yield a nct count, l '- above system background, that w!11 be detected with 95% probability with only 5%
$ probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, (which may include radiochemical - ,
' separation):
4.66sb ---- LLD =
---------------10 E- * ---
E
- V
- 2.22 x Y * --
e xp---(-16 t ) Where: LLD is the a prior! lower limit of detection as defined above (as microcuries perunitmassorvoluce), - is the standard deviat'on vf the background counting rate or of the
;sbcounting rate of a blank s.wple as appropriate (as counts per minute),
E is the counting efficiency (as counts per disintegration), : V is the sample size (in ur.its of matt or volume), 2.22 x 10 6 is the number of oisintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable). A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples) The value of sb used in the calculation of the LLO for a det orofthecountingrateoftheblanksamples(asappropriate)ratherthanon an unverified theoretically predicted variance. Typical values of E V, Y, and At shall be used in the calculation, it should be recognized that the LLD is defined an an a priori (before the fact) limit representing t.he Capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Engr.G H . P . J. VS Date y on I-43 Rev. n i
f slPg 3/a 11-11/odt/4 - HW5 TABLE 13.3-3 (Continued) TABLE NOTATIONS b Sampling and analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1 hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT l-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the main condenser offgas pre-treatment radioactivity monitor shows that effluent activity has not increased more than a factor of 3. - - c Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour and analyses completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding This reouirement does not apply if LLDs may be increased by a factor of 10. (1) analysis shows that the DOSE EQU! VALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas
- monitor shows that effluent activity has not increased more than a factor of 3.
d 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 Controls 3.3.2, 3.3.3, and 3.3.4. e The principal gamma emitters for which the LLO specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m and Xe-138 for gaseous emissions and Mn-54, Fe-59 Co-58, Co-60, 2n-65, ThisMo-99, list 1-131. Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.Other gamma j does not mean that only these nuclides are to be considered. l peaks which are identifiable, together with those of the above nuclides, shall l also be analyzed and reported in the Semiannual Radioactive Effluent Release ' Report, pursuant to Control 3.6 f Under the provisions of footnote e. above, only noble gases need to be considered. g Required for the hot maintenance shop ventilation exhaust only during ! operation of the hot maintenance shop ventilatiori exhaust system. Engr. ! H. P. < Date NA l-44 Rev. j_ 1 ! l ! 1
s%Pg 3/4 11-12/od874 - RWG DOSE - NOBLE GASES CONTROLS 3.3.3 In accordance with LGS TS 6.8.4.d.5 and 8, the air dose du'e to noble gases released in gaseous effluents, from the site to areas at and beyond th
'BOUNDA1Y(seeFigure12.2-16)shallbelimitedtothefollowing: '
Less than or equal to 10 mrads for gamma a. During any calendar quarter: radiation and less than or equal to 20 mrads f Less than or equal to 20 mrads for gamma
- b. During any calendar year:
radiation and less than or equal to 40 mrads for beta radiation. APPLICABILITY: At all times, gTION:
- a. With the. calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reducej releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
SURVEILLANCE REQUIREMENTS 3.3.3.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. Engr. Q 8 H. P. md te ic 1 45 i
s;Pg 3/4 11-13/odt74 - RWG ; DDSE - 10 DINE-133. TRIT!UM. AND RADIONUCLIOES IN PARTICULATE F a CONTROLS l 3.3.4 In accordance with LGS TS 6.8.4.d.P. the dose to a MEMBER O from iodine-131, iodine-133, tritium, and all radionuclides in particulate form
- with half-lives greater than 8 days in gaseous effluents released, from the site to areas at and beyor 4 theSITEBOUNDARY(seeFigure12.2-la)shallbelimitedto the following:
Less than or equal to 15 mrems to any organ
' a. During any calendar ouarter:
and, Less than or equal to 30 mrems to any organ.
- b. During any calendsr year:
APPLICABILITY: At all times, gTION:
- a. With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed correJtive actions to be taken to assure that subsequent releases will be in compliance with the above limits.
SURVEILLANCE REQUIREMENTS 3.3.4.1 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than B days shall days. Engr. 6 8 H. P. Mio I-46 gte
5;Pg 3/411 14/0d174 - JtWG
. 4 VENT!LATION CXHAUST TREATMENT SYSTEM i
CONTROLS i 3.3.5 In accordance with LGS TS 6.8.4.d.6, the VENTILATION EXHAUST TREATMENT I SYSTEM shall be OPERABLE and appropriate portions of the system shall be usti to reduce radioactive materials in gaseous waste prior to their discharge when t a . projected doses due to-gaseous effluent releases, from the site to aress at ano l bey'ond the $1TE BOUNDARY (see Figure 12.2-1) when averaged over 31 J . , exceed 0.6 mrem to any organ in a 31-day period. j APPLICABILITY: At all times. l ACTION:
- a. With gaseous waste being d sc.targed without treatment, and in excess of the above limits, prepare aid submit to the Commission within 30 days, pursuant to Technical Specificition 6.9.2, a Special Report which '
includes the following inform 0 tion:
- 1. Explanation of why gaseous radwaste was being discharged without treatment, identification o' any inoperable equipment or subsystems, and the reason for the inope* ability.
- 2. Action (s)takentorestoretheinoperableequipmenttoOPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
SURVEILLANCE REQUIREMENTS I Doses due to gaseous releases from the site to areas at and beyond the 3.3.5 1 SITE boundary shall be projected at least once per 31 days in accordance with the methodology and parameters in the 00CM. 3.3.5.2 The VENTILATION EXHAUST TREATMENT SYSTE'M shall be dem OPERABLE by meeting Controls 3.3.2, 3.3.3 or 3.3.4 - Engr. H. P. ! Date a I-47 Rev. I
\
#~'
VENTINGORPURC]NG CONTROLS 3.3.6 in accordance with LGS TS 6.8.4.d.10. VENTING or PURGING of th containment shall be through the standby gas treatment system. Whenever the Containment is vented or purged.'
. APPLICABILITY _:
ACTION: ~ a. With the requirements of the above specification not satisfied, susp'end all VENTING and PURGING of the containment. SURVEILLANCE REQUIREMENTS The containment shall be determined to be aligned for VENTING or 3.3.6.1 PURGING through the standby gas treatment system within 4 hours prior to start of and at least once per 12 hours during VENTING or PURGING of the containment. Prior to use of the purge system through the standby gas treatment 3.3.6.2 system assure that; a. Both standby gas treatment system trains are OPERABLE whenever the purge system is in use, and b. Whenever the purge system is in use during OPERATIONAL CONDITION 1 or 2 or 3. only one of the standby gas treatment system trains may be used.
*Except for Ee one inch /two inch vent valves to the Reactor Enclosure Equipment Compartment Exhaust filtert when used for containment pressure control and nitrogen make-up operations. '
Engr. T H. P. Date o 1-48 Rev. , __ N _ _ ~ ~ ^ ~ - - - - - - - _ _
stVg .1tse d - at w, ~ ^ .... . , r l RADIDACTIVE EFFLUENTS TOTAL OOSE CONTROLS-
** 3.3.7 in accordance with LGS TS 6.8.4.d.11, the annual (calendar year) dose or ,
dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactiv " radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be liinited to less than or equal to 75 mrems. APPLICABillTY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materia liquid or gaseous effluents exceeding twice the limits of Control .
3.2.3.a. 3.2.3 b., 3.3.3.a., 3.3.3.b., 3.3.4.a., or 3.3.4.b., calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks If to determine Such whether the is the case, above limits of Control 3.3.7 have been exceeded. prepare and submit to the Commission within 30; d to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits-and includes the schedule for achievingT conformance with the above limits. CFR 20.405c, shall include an analysis that estimates the radiation
- exposure (dose) to a MEMBiR OF THE PUBLIC from the uranium fuel cycle sources, including all ef fluent pathways and direct radiation, for the It calendaryearthatinclulestherelease(s)coveredbythisreport. '
shall also aescribe levtis of radiation and concentrations of ra material involved, and the cause of the exposure levels orIf the concentrations. resulting in violation 40 CFR Part 190 has not if the release conditiot already been corrected, the Special Report shall include a Submittal request for a variance in accordance vith the provisions of 40 CFR Part 190. of the report is considered a timely request, and a variance is granted until staff action on the request is complete. SURVEILLANCE REQUIREMENTS ._ Cumulative dose con,tributions from liquid and gaseous effluents shall b 3.3.7.1 3.2.3.1, 3.3.3.1. and 3.3.4.1, and in determined in accordance with Controls accordance with the methodology and parameters in the ODCM. 3.3.7.2 If the cumulative dose contributis's exceed the limits defined 3.3.7a. Cumulative dose contributions from dicect r Eng Part II. l
"* Wf 1
s;Pg 3/4 12-lodc;;74 - RWG- [ RADIOLOGICAL ENVIRONMENTAL C NITORING PROGRAM ' CONTROLS 3.4.1 In accordance with LGS TS 6.8.4.f.1, the radiological environmental monitoring' program shall be conducted as specified in Table 13.4-1. APPLICABILITY: At all times. ACTION
- a. With the radiological enviernmental monitoring program not being conducted as specified in Table 13.4 1, prepare and submit to the l Comission, ie the Annual Radiological Environmental Operating Report per Control 3.5, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
'b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 13.4-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s)forexceedingthelimit(s)anddefinesthecorrectiveactions to be taken to reduce radioactive effluents 50 that the potential annual dost to a MEMBER OF THE PUBLIC is less than the calendar year limits of Controls 3.2.3, 3.3.3 and 3.3.4. When more than one of the radionuclides in Table 13.4-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentration (2) reporting level (1)
+
reportinglevel(2)
+
...{1.0 When radionuclides other than those in Table 13.4 2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Controls 3.2.3, 3.3.3, and 3.3.4. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
.S 1-50 Date Rev.
C
sjPg3/4 12-2/odem:74 - RWG
. RADIOLOGICAL ~ ENVIRONMENTAL MONITORING I CONTROLS ACTION (Continued)
- c. With milk or fresh leafy vegetable samples unavailable from one or more !
' of the sample locations required by Table 13.4-1, identify. locations !
for obtaining replacement samples and add them to thespecific The radiological environmental monitoring program within 30 days. locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Technical Specification 6.14, submit as a part of or concurrent with the Semiannual Radioactive Effluent Release Report, a complete, legible copy of the entire 00CM, including a revised figure (s) and table for the ODCM reflecting the new : location (s). SURVEILLANCE REQUIREMENTS 3.4.1.1 The radiological environmental monitoring samples sna11 be collected pursuant to-Table 13.4-1 from the specific locations given in the table and figure (s) in the ODCM and shall be enslyzed pursuant to the requirements of Table 13.4-1, the detection capabilities required by Table 13.4-3. Engr.4 d H. P. Date a 1-51 Rev. ____; l l
~
lABLE 13.4-1 RADIOLOGICAL ENVIRONMENTAL MONIl0 RING PROGRAM NUMBLR Of RLPRESENIAllVE SAMPLING AND TYPE AND FREQUENCY
- . EXPOSURE PATHWAY SAMPLES AND COLLECTION FREQUENCY OF ANALYSIS AND/0R SAMPLE SAMPLE LOCATIONS (a) 40 routine monitoring stations. At least Quarterly. Gamma dose at least
- 1. IslRECT RADIATION (b) quarterly.
either with two or more dosimeters or with one instrement for measuring and recording dose rate continuously placed as follows: (1) An inner ring of stations, one in each meteorological sector, in the general area of the Si1E BOUNDARY: (2) An outer ring of stations, one in each meteorological sector, in the 3-to-9-mile range from the site; , (3) The balance of the stations placed in special interest areas - such as population centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations.
- 2. AIRBORNE Samples from 5 locations: Continuous sampler . Padioiodine Cannister:
Radiolodine and I-131 analysis weekly. Particulates operation with sample
- a. 3 samples from close to the collection weekly, or 3 SilE BOUNDARY locations (in more frequently if required by dust Particulate Sampler:
different sectors) of the Gross beta radio highest calculated annual loading.c activity analysis average groundlevel x/Q I sample from the vicinity followi
- b. change;ggfilter community having one of the Gamma isotopic highest calculated annual analysis
- of highest groundlevel x/Q.
composite (by location) at least quarterly. Engr.75'8 H. - P.hZ Date W .4 1-52 Rev. 7.s'
s;Pg 3/* ac-,vus... . . .. TABLE 13.4-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER Of REPRESENTATIVE TYPE AND FREQUENCY SAMPLING AND SAMPLES AND Of ANALYSIS XPOSURE PATHWAV COLLECTION FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS (a)
- 2. AIRBORNE (Continued) 1 sample from a control loca-tion, as for example 15-30 km distant and in the least pre-valent wind direction.
3-. WATERBORNE Composite sample ove'r Gamma isotopic analysis *
- a. Surface f I sample upstream 1-month period.9 monthly. Composite for 1_ sample downstream tritium analysis quarter 1,y.
Quarterly. Gamma isotopic" and tritium
- b. Ground Samples from 1 or 2 sources analysis quarterly, only if likely to be affected.h Moathly. (Composite) 1-131 analysis on each
- c. Drinking 1 sample of each of 1 to 3 composite when the dose of the nearest water supplies calculated for the con-that could be affected by its sumption of the water discharge. is greater than 1 mrem per year.' Composite for gross beta and gamma 1 sample from a control isotopic analyses monthly.
location. Composite for tritium analysis quarterly. Sealannually. Gamma isotopic analysis *
- d. Sediment I sample from downstream area semiannually.
from with existing or potential
. shoreline recreational value.
Engr.J-H. P. Date 70 I-53 Rev,
s;Pg 3/4 12-Sodco:/4 - nwu TABLE 13.4-1 (Continued)
- RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM s
huMBER Of REPRESENTATIVE TYPE AND FREQUENCY SAMPLING AND EXPOSURE PATHWAY SAMPLES AND Of ANALYSIS COLLECTION FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS (a)
- 4. INGESTION ,
Gamma isotopic
- and I-131 Samples from milking animals Semimonthly when ani-
- a. Milk a. mais are on pasture, analysis semimonthly when in 3 locations within 5 km monthly at other times. animals are on pasture; distance having the highest monthly at other times.
dose potential. If there are none, then, I sample from milking animals in each of 3 areas
- between S- to 8-km distance where doses are calculated to be .
greater than 1 mrem per yr.I
- 1 sample from milking animals at
. a control location (15-30 km
~
km distant and in the least prevalent wind direction). Gamma isotopic *
- b. Fish and a. I sample of two recreationally Sample in season, or semiannually if they analysis on edible Invertebrates important species la vicinity portions.
of plant discharge area. are not seasonal. I sample of same species in area not influenced by plant discharge.
!"SR'H$.
Date ':r " J> 1-54 Rev. 45 V e 4
,h
s;Pg an u-ouuu . ,, ..~ Tr-;tLE 13.4-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM s NUMBER Of REPRESENTATIVE TYPi AND FREQUENCY SAMPLES AND SAMPLING AND EXPOSURE PATHWAY COLLECTION FREQUENCY Of ANALYSIS AND/OR SAMPLE SAMPLE LOCATIONS (a) _ 8
- 4. INGESTION (Continued)
Sag les of 3 different Monthly when available, Gama isotopic" and
- c. food Products if milk sampling is 1-131 analysis.
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