ML20082C268
| ML20082C268 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/03/1995 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20082C265 | List: |
| References | |
| NUDOCS 9504060260 | |
| Download: ML20082C268 (90) | |
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DEFINITIONS
.S.ECTION DEFINITIONS (Continued) PAGE !
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r s' - i 1.25 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-5 1.26 MEMBER (S) 0F THE PUBLIC...................................... 1-5 1.27 MINIMUM CRITICAL POWER RATI0................................. 1-5 MANUAL.............................. 1.28 0FFSITE DOSE CALCULATION 1-5 1.29 OPERABLE - OPERABILITY....................................... 1-6 1.30 OPERATIONAL CONDITION - CONDITION............................ 1-6 - 1.31 PHYSICS TESTS................................................ 1-6 l'.32 PRESSURE BOUNDARY LEAKAGE.................................... 1-6 1.33 PRIMARY CONTAINMENT INTEGRITY................................ 1-6
- 1. 34 PROCESS CONTRO L PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 1.35 PURGE - PURGING.............................................. 1-7 1.36 RATED THERMAL P0WER.......................................... 1-7 1.37 REACTOR PROTECTION SYSTEM RESPONSE TIME...................... 1-7 .
1.38 REPORTABLE EVENT............................................. 1-7 1.39 ROD DENSITY.................................................. 1-7 1 1.40 SECONDARY CONTAINMENT INTEGRITY.............................. 1-7 l 1.41 SHUTDOWN MARGIN............. ................................ 1-8 I 1.42 SITE B0VNDARY................................................ 1-8 (beim+ce : 1.43".~'."'"'."n..'~n'...............................................
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1 1.45 STAGGERED TEST BASIS......................................... 1-8 1.46 THERMAL P0WER................................................ 1-9 9504060260 950403 PDR ADOCK 05000440 P PDR PERRY - UNIT 1 11
Py-c cr/so ng-Ib55L ' A ff o cA ment a ' Pacc 2 of F7 DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.47 TURBINE BYPASS SYSTEM RESPONSE TIME.......................... 1-9 1.48 UNIDENTIFIED LEAKAGE......................................... 1-9 1.49 UNRESTRICTED AREA............................................ 1-9 ( Det.ete.4) 1.50 .,m..r,, S ru ..
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1.51 VENTING...................................................... 1-9 Table 1.1, Surveillance Frequency Notation........................ 1-10 Table 1.2, Operational Conditions................................. 1-11 s i I i l PERRY - UNIT 1 iii
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U./ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ! INSTRUMENTATION (Continued) Meteorological Monitoring Instrumentation............... 3/4 3-70 Table 3.3.7.3-1 Meteorological Monitoring ' Instrumentation................... 3/4 3-71 Table 4.3.7.3-1 Meteorological Monitoring Instrumentation Surveillance Requirements...................... 3/4 3-72 ; Remote Shutdown System Instrumentation and Controls..... 3/4 3-73 i Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls...... 3/4 3-74 ; Table 4.3.7.4-1 Remote Shutdown System ! Instrumentation Surveillance Requirements...................... 3/4 3-76 Accident Moni toring Ins trumentation. . . . . . . . . . . . . . . . . . . . . 3/4 3-77 Table 3.3.7.5-1 Accident Monitoring Instrumen- t tation............................ 3/4 3-78 ? Table 4.3.7.5-1 Accident Monitoring Instrumenta- , tion Surveillance Requirements.... 3/4 3-80 i Source Range Monitors................................... 3/4 3-81 Traversing In-Core Probe System......................... 3/4 3-82 Loose-Part Detection System............................. 3/4 3-83 R di ::tive Liquid Eff?;:nt M: nit:rin; !a;tr_; n t: tion .. ... . .... ......... ................ 3/4304" T:b!: 3.3.7.9-1 ":di:::tiv Liquid Effluent " M:ni toring In;tra; nt:ti:n. . . . . . . . 3/4305 ' Tobic '.3.7.0-1 ":di ::tiv: Liquid Ef'le:nt 4-M: nit: ring Instra;;nt ti:n "- 5;rveills,:: " quire; nts. ....... 3/4 3 07 t-s PERRY - UNIT 1 ix
PY - CE.E/N/?R- I655 L , 4 #ttcA m d A FAgc 4 o f 67 "? LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PA,G,,E INSTRUMENTATION (Continued) ma ;., Gn/enser Off a R:di ::ti :ys &cofascsf Sy.s+cm C::::= Ef'lur.t SpicsiveInstrumen-Monitoring Ges tation.................................................. 3/4 3 f 4 Tel: 3. 3. ' . '. 0- 1 R: dies:tive Cn::= Ef'? : t #-
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rvei? ? =:: ":;ui. nnts........ 2/4203"- 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...................... 3/4 3-96 3/4.3.9 PLANT SYSTEM ACTUATION INSTRUMENTATION................... 3/4 3-98 Table 3.3.9-1 Plant Systems Actuation Instrumentation..................... 3/4 3-100 Table 3.3.9-2 Plant Systems Actuation Instrumen-tation Setpoints.................... 3/4 3-101 Table 4.3.9.1-1 Plant Systems Actuation Instrumentation Surveillance Requirements...................... 3/4 3-102 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops..................................... 3/4 4-1 Figure 3.4.1.1-1 Thermal Power versus Core F10w............................. 3/4 4-3 Jet Pumps............................................... 3/4 4-4 Recirculation Loop F1ow................................. 3/4 4-5 l Idle Recirculation Loop Startup.... .,.................. 3/4 4-6 3/4.4.2 SAFETY / RELIEF VALVES Safety / Relief Va1ves.................................... 3/4 4-7 Safety / Relief Valves Low-Low Set Function. . . . . . . . . . . . . . . 3/4 4-8 PERRY - UNIT 1 x
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h LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS , SECTION PAGE 6 3/4.~10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY........................... 3/4 10-1 , 3/4.10.2 R00 PATTERN CONTROL SYSTEM.............................. 3/4 10-2 l 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS ......................... 3/4 10-3 l 3/4.10.4 RECIRCULATION L00PS..................................... 3/4 10-4 3/4.10.5 TRAINING STARTUPS....................................... 3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS q A 11 e T an e w ss s-P P n e ncanvr a
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FY- CEZ/NTlR - /GSS L Ada.c.hMcd A Pye lo of f7 l ( LIMITING CONDITIONS 'FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADI0 ACTIVE EFFLUENTS (Continued) ; 3/4.11.3 SOLID RADWASTE TREATMENT................................ { 3/4 11-18 i 3/4.11.4 TOTAL D0SE.............................................. 3/4 11-20 l 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ' 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4 12-1 f Table 3.12.1-1 Radiological Environin6ntal ! Monitoring Program................. 3/4 12-3 , Table 3.12.1-2 Reporting Levels for Radio-activity Concentrations In , Environmental Samples.............. 3/4 12-9 Table 4.12.1-1 Detection Capabilities For Environmental Sample Analysis < - Lower Limit of Detection...........
- 3/4.12.2 LAND USE CENSUS......................................... 3/412-10lI t 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... 3/4 12-14 /
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l SECTION fAGE i INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION
'i Radiation Monitoring Instrumentation ........... B 3/4 3-4a l Seismic Monitoring Instrumentation ............ B 3/4 3-4b {
Meteorological Monitoring Instrumentation . .. . . . . .. B 3/4 3-4b' l Remote Shutdown System Instrumentation and Controls I
. .. B 3/4 3-5 Accident Monitoring Instrumentation ........... B 3/4 3-5 l Source Range Monitors .................. B 3/4 3-5 l Traversing In-Core Probe System ............. B 3/4 3-5 l Loose-Part Detection System ............... B 3/4 3-6 .j n 3f4 3-$ #e I
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Bases Figure B 3/4 3-1 Reactor Vessel Water ! Level ............. B 3/4 3-8 j 3 /4. 4 REACTOR COOLANT SYSTEM . 3/4.4.1 RECIRCULATION SYSTEM ................... B 3/4 4-1 i 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 ! 0'p erational Leakage ................... B 3/4 4-3a l 3/4.4.4 CHEMISTRY ........................ B 3/4 4-4 i PERRY - UNIT I xix Amendment No. 67
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9 t, l' ADMINISTRATIVE CONTROLS l r SECTION PAGE 6.12 HIGH RADIATION AREA........................................ 6-23 I v 6.13 PROCESS CONTROL PROGRAM (PCP).............................. 6-24 l 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)..................... 6-25
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l'Y U.L/NKR- /bb > L Affacluwed c2 Pye lo of 77 DEFINITIONS f LIMITING CONTROL ROD PATTERN 1.22 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR. LINEAR HEAT GENERATION RATE 1.23 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. L!OL" O "^ 0'.!^.ST T" EAT"C"T S'fST " - ( De,1.e.t e.d ) 1.24AThe L!Q"!D AD'c!^.STE TRE!*"' SYSTE" h :ny pr:== :r centr:? quip:=t
=:d to r: dun the rernt er cencentr: tier :f liquid r:dhntiv: = t: r hi s --
prier t; thei- dinh:rge te LTREST ICTED ^REAS. It i;=h : 211 the in:t:1 k d
- nd :=ibbh 'iquid rii!=t : = :;::=t ty',t= : quip =t, = w li :.s their-
- ntreh , p:=r %tru:=titien, =d :er"he th:t =k: th: :y:t= h =th =1.6 I
LOGIC SYSTEM FUNCTIONAL TEST 1.25 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 i 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. MEMBER (5) 0F THE PUBLIC ; 1.26 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupa-tionally 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. MINIMUM CRITICAL POWER RATIO 1.27 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core. OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.28 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent moni-toring alarm / trip setpoints, and in the conduct of the 'fadiological fnvironmental fonitoringfrogram.A INSERT Frc. nes+ ge ( PERRY - UNIT 1 1-5
r Py- LEI /N2R-Ib55L wam w+ a P5e H of 77 , INSERT FOR PAGE 1-5 , speJfiut Jn : The ODCH shall also contain (1) the Radio. active Effluent Co ols and ; Radiological Environmental Monitoring Programs required by - n i: 6.8.4 and (2) descriptions of the inforaation that should be included in the Annual i Radiological Environmental Operating and Annual Radioactive Effluent Release , Reports required by Specifications 6.9.1.6 and 6.9.1.7. ! f f k I i 1 i l 1 l
t PY- cEIfpre /b55L l' Alla chmeyrt- 2. f e lh o f '87 OEFINITIONS * ( ;
- 1. Capable of being closed by an OPERABLE containment automatic ;
isolation system, or '
- 2. Closed by at least one manual valve, blind flange, or deactivated -
automatic valve, as applicable secured in its closed position. ;
- b. The containment equipment hatch is closed and sealed and the t shield blocks are installed adjacent to the Shield Building. l
- c. The door in each access to the annulus is closed, except for normal entry and exit.
- d. The sealing mechanism associated with each Shield Building l penetration, e.g., welds, bellows or 0-rings, is OPERABLE.
4
- e. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.6.1.a., except for normal entry and exit to the annulus. ,
- f. The Annulus Exhaust Gas Treatment System is in compliance with the requirements of Specification 3.6.6.2.
I SHUTDOWN MARGIN i 1.41 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; and xenon free. SITE BOUNDARY 1.42 The SITE SOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. 00LI0ITICATION 1 (Deleted 1.43^00LICIi! CATION 3 hall be the G..nr;in ;f .at ant = 'it: : f:= thn l
- = t: :Hpping :nd 5 rf:1 gr: nd 7:;;f r==te.t 00'J"C C:: CX ".-
(Da.Le.t ed) 1.44^^ """CE C"ECM :h:l' 5: th: ;xlit;ti.; ann =.7t ;f ch==1 rupan ch:- th; ch;r=1 nn;r i; s;n;d t; ; nera ;f inrGnd Rdin:ti".it). L STAGGERED TEST BASIS 1.a5 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains or other designated i
- components obtained by dividing the specified test interval into n equal subintervals. PERRY - UNIT 1 1-8 :
1 YI ~ Cf:.L/AIK2-/L'SS L ! 48tte 4menf 6l hope Ib of E7 DEFINITIONS j
.,: I b b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER i 1.46 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TURBINE BYPASS SYSTEM RESPONSE TIME 1.47 The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two separate time intervals: a) time from initial movement of the main turbine stop valve or control valve until 80% of turbine hypass capacity is established, and b) the time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve. Either response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured. UNIDENTIFIED LEAKAGE i 1.48 UNIDENTIFIED LEAKAGE shall be all leakege which is not IDENTIFIED LEAKAGE. UNRESTRICTED AREA , ( 1.49 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 i
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MEMBERS OF THE PUBLIC from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. VENTILATION EXNAUST TREATNENT SYSTEMS t-(DwLet ed) , 1.50*A VENTILAT!0N EX"A"ET TREAT"ENT SYSTE" i: : y ;y;t : d::f;n;d ;nd in:t:11:d i: r:d :: ;:::::: r:d':f: din: :7 :d!:::tiv: ::t:rt:1 ' p:rtice?::: , f:r: #- f'!::nt: by p:::!ng :nt!!:ti:r :r ::nt : h:::t ; ::: thr::;h :h:r:::1 ! cd::rt:r: :rd/:r "EP^ "!t:r: f:r the per;::: ef r;;;;'n; f: din:: :: , p:rti::1:te: 'r : th: ;:::::: :nh:::t :tr::: prf:r t: th: 7:1:::: t: th: crr'r:rt:nt (::ch : cy t:- f: net :: :f dcr:d i: h::: :n; cff :t :n n:ti: ;;; cf'! :nt:). Engin r:d Erf:ty F::tur: (EEF) ater:;h:rf: :1::ne; ty:t::: :r: ; r.;t ::n;fd:r:d t: 5: "E"'IL*'?OM EX"^1'ET TREA'"E"' SYSTE" :: ;:n:n".: pr: fd:d th: ESF y:t r f: n:t et!: d t: tr :t n:r;:1 r !:::::. O_, , VENTING 1.51 VENTING is the controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air er gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. , PERRY - UNIT 1 1-9 !
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U~ WZ/Af/22- /GSG L-Atla c), suenf & INSTRUMENTATI NO# RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4 3.3.7.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7.9-1 shall be OPERABLE with their alars/ trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the OFFSITE DOSE CALCULAlION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
,1
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above f specification, immediately suspend the release of radioactive liquid !
effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.9-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain why this inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.
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! c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS f 4.3.7.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.9-1. 1 l 3/4 3-84 Amendment No. 30, 49 (PERRY-UNIT 1 '
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R. , TABLE 3.3.7.9-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION i MINIMUM U c CHANNELS P
~
INSTRUMENT OPERABLE ACTION k
- 1. GROSS RADI0 ACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE s~
3
- a. Liquid Radwaste Discharge Radiation Monitor - 1 110 +
ESW Discharge ;ii G
- 2. GROSS BETA OR GAMMA RADI0 ACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
- a. ' Emergency Service Water Loop A Radiation Monitor 1 111
, b. Emergency Service Water Loop B Radiation Monitor 1 111 2
Y 3. FLOW RATE MEASUREMENT DEVICES
- a. Radwaste Jischarge Header
- 1. Radwaste High Flow Discharge Header Flow 1 112
{ 2. Radwaste Low Flow Discharge Header Flow 1 112 1
- b. Service Water Discharge Header Flow 1 113l
- c. Unit 1 Emergency Service Water Flow Monitor 1 113
) r% %
J g%^ w. ng i sa
@ I n ,k ..
t %d g Wi. r
i l'Y- WI@lL2-16SS L AAttch wav1t 2 ; Pye /& of 57 TABLE 3.3.7.9-1 (Continued) * -f
\ '
RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, efficent releases from this pathway may continue provided that prior to initiating a release: ;
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
FACTION 111 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at ( least 10 7 microcuries/ml. , ACTION 112 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requiremt.nt, effluent releases via , i this pathway may continue provided the discharge valve position is verified to be consistent with the flow rate provi' ions of the release permit at least once per 4 hours during actual releases. Prior to initiating another release, at least two tr.:hnically qualified members of the Facility Staff shall indervndently verify the discharge line valving and that the discharge valve position corresponds to the desired flow rate. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 113 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump per-formance curves generated in place may be used to estimate flow. _-- _- . J D ele.te e.n+ir e pay PERRY - UNIT 1 0-- 3/a 3-86 2---
TABLE 4.3.7.9-1 .. , 7, , g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS j[ c CHANNEL SOURCE CHANNEL- [
' CHANNEL FUNCTIONAL =n-h INSTRUMENT CHECK CHECK CALIBRATION- TEST
" 1 54
- 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE hi G
- a. Liquid Radwaste Discharge Radiation D 1
Monitor - ESW Discharge P R(3). Q(1)- ]
- 2. GROSS BETA OR GAMMA RADI0 ACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
, a. Emergency Service Water Loop A Radiation D M g Monitor R(3) Q(2)
[ b. Emergency Service Water Loop B Radiation D M u Monitor R(3) Q(2) ,
- 3. FLOW RATE MEASUREMENT DEVICES
- a. Radwaste Discharge Header
- 1. Radwaste High Flow Discharge Header Flow D(4) N.A. R Q- ;
- 2. Radwaste Low Flow Discharge Header Flow D(4) N.A. R Q
- b. Service Water Discharge Header Flow D(4) N. A. R Q D'
A
- c. Unit 1 Emergency Service Water Flow D(4) N.A. R lb g '
i Monitor Q. s :2 A a j g .'- p
%Wi N g
. r
M- C EI/WRrl -/s55 L M chnwnf A Ac)e /8 of g7 TABLE 4.3.7.9-1 (Continued) f ( RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATIUN SURVEILLANCE REQUIREMENTS TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint. l
- 2. Instrument indicates a downscale failure.
I
- 3. Instrument controls not set :n operate mode except in high voltage i position.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarr i annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
f 2. Instrument indicates a downscale failure.
- 3. Instrument controls not set in operate mode, except in high voltage
(, position. (3) The initial CHANNEL CALIRRATION shall be performed using one or more of the reference standards certified by the Naticnal Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days which continuous, periodic or batch releases are made. f DedC e/\ bw'c p C PERRY - UNIT 1 3/4 3-88
i PY- m.I/MM -/455 L AWA cj1mem F 2 \ INSTRUMENTATION \ PA-jC /Y O[ b7 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION e 3.3.7.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.7.10-1 shall be OPERABLE with their alarz A rip setpoints set to ensure that the limit of Specification 3.11.2.1* are not exceeded. The alarm / trip setpoints of applicable chann21s shall be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 3.3.7.10-1 ' ( ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, declare the channel inoperable, or change the setpoint so it is acceptably conservative,
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.10-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain why th'is inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Reoort. .V
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.10 Each radioactive gaseous effluent monitoring instrumentation channel
- ' shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.10-1. ,
/ kgLe- w e 6 Atfa. d CS T bl3 N I "See Specification 3.11.2.6 for the Main Condenser Offgas Hydrogen Monitor (IN64-N012A/B) limit.
PERRY - UNIT 1 4/4 3-59 " Amendment No. 3D,49 3/4 3 - T 'f
i i INSERT FOR SPECIFICATION 3/4.3.7.10 INSTRUMENTATION MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATILI 3.3.7.10 At least one main condenser offgas treatment system explosive gas monitoring instrumentation channel (IN64-N012 A/B) shall be OPERABLE with its alarm / trip setpoint set to ensure that the limits of Specification 3.11.2 are not exceeded. APPLICABILITY: Whenever the main condenser offgas eatment system is in operation. ACTION:
- a. With an explosive gas monitoring instrumentation channel alarm / trip setpoint less conservative than required by Specification 3.11.2, declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. Vith less than one main condenser offgas treatment system explosive gas monitoring instrumentation channel OPERABLE, operation of the main condenser offgas treatment system may continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. If the recombiner temperature remains constant and THERMAL POVER has not changed, the grab sample collection frequency may be changed to at least once per 8 hours. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.
- c. The provisions of Specification 3.0.3 are not applicable.
PERRY - UNIT 1 3/4 3-84
l'Y- cez/ Awe-isss c V M acJar,a f a - hje 9( of 67 SURVEILLANCE REQUIREMENTS 4.3.7.10 The explosive gas monitoring instrumentation channel (s) shall be i demonstrated OPERABLE by performance of a: , I
- a. CHANNEL CHECK at least once per 24 hours,
- b. CHANNEL CALIBRATION at least once per 92 days. The CHANNEL CALIBRATION shall include the use of standard samples containing a nominal: l
- 1. One volume percent hydrogen, balance nitrogen, and
- 2. Four volume percent hydrogen, balance nitrogen.
- c. CHANNEL FUNCTIONAL TEST at least once per 31 days.
i t
)
i a i I 5 i I t i f l PERRY - UNIT 1 3/4 3-85 (The next page is 3/4 3-96) , I
=m.- , . _ _ , , - . - - , - - , . . - - - ,
~
TABLE 3.3.7.10-1 9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4 - b MINIMUM CHANNELS ( INSTRUMENT OPERABLE APPLICABILITY ACTION
~
h'
- 1. OFFGAS VENT RADIATION MONITOR k -
- a. Noble Gas Activity Monitor 1
- 121 Q
+
- b. Iodine Sampler 1 122 3 a
- c. Particulate Sampler 1 122
- d. Effluent System Flow Rate Monitor 1 123 y e. Sampler Flow Rate Monitor 1
- 123 Y 2. UNIT 1 VENT RADIATION MONITOR
- a. Noble Gas Activity Monitor 1 1,2,3 125 4, 5 121
- b. Iodine Sampler 1
- 122
- c. Particulate Sampler 1
- 122
- d. Effluent System Flow Rate Monitor 1
- 123
- e. Sampler Flow Rate Monitor 1
- 123 g i me u
>5 oE Ok
~yxe s
VL r
^ '
v, [ T BLE 3.3.7.10-1 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION E MINIMUM CHANNELS , Z INSTRUMENT OPERABLE APPLICABILITY ACTION ,
- 3. UNIT 2 VENT RADIATION MONITOR k(b Noble Gas Activity Monitor *
- a. 'l 121
[ , { b. Iodine Sampler 1
- 122 (b ;
3
- c. Particulate Sampler 1
- 122 f, !
7
- d. Effluent System Flow Rate Monitor 1
- 123 o w e. Sampler Flow Rate Monitor 1
- 123 T) h .
P-w 4. HEATER BAY / TURBINE BUILDING VENT RADIATION MONITOR (f E O
~ a. Noble Gas Activity Monitor
- 1 121 l
- b. Iodine Sampler 1
- 122 l c. Particulate Sampler 1
- 122
- d. Effluent System Flow Rate Monitor 1
- 123
- e. Sampler Flow Rate Monitor 1
- 123
- 5. MAIN CONDENSER OFF-GAS HYDROGEN MONITOR
- a. Hydrogen Monitor 1 **
124 - 4 l m w .@ . - l
.o S -
N. s N xD , N' F : L.__.____.--_.._ _- _ - _ .._ _ _ __ .. . _ - - - _ . . _ _ _ _._ _
A Y~ CEI/A/Le-ff 55( \ MACM/meaf A Pkje BM of 77 I TABLE 3.3.7.10-1 (Continued) ( ' RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATION
- At all times.
** During main condenser offgas treatment system operation.
ACTION 121 - With the number of channels OPERABLE less than required by the Minmum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least
- once per 12 hours and these samples are analyzed for gross activity w; thin 24 hours.
ACTION 122 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected within 8 hours with auxiliary sampling equipment as required by Table 4.11.2.1.2-1. ACTION 123 - With the number of channels OPERABLE less than required by the { Hinimum Channels OPERABLE requirement, effluent release via thist pathway may continue provided the flow rate is estimated at least once per 4 hours. ACTION 124 - With the number of channels OPERABLE less than required by the ( Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. If the recombiner temperature remains constant and THERMAL POWER has not changed, the grab f sample collection frequency may be changed to at least once per 8 hours. ACTION 125 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, except as a result of a non-conservative setpoint, immediately suspend containment / drywell purge and vent. Prior to resuming containment /drywell purge and vent, ensure compliance with the requirements of Specification 3.11.2.1. If compliance with Specifica-tion 3.11.2.1 is met, containment /drywell purge and vent may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours.
\ htch. enke pge-l 4&RM UNIT M ' -3/4 0 '32 "
i
~
m TABLE 4.3.7.10-1 h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3A E CHANNEL MODES IN WHICH Q CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE g INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED G
- 1. OFFGAS VENT RADIATION MONITOR I
- a. Noble Gas Activity Monitor D M
- R(2) Q(1)
- b. Iodine Sampler W(4) N.A. N.A. N.A. *
- c. Particulate Sampler W(4) N.A. N.A. N.A. *
- d. Effluent System Flow Rate Monitor D N. A. R
- Q
$ e. Sampler Flow Rate Monitor D N.A. R
- Q
- 2. UNIT 1 VENT RADIATION MONITOR
- a. Noble Gas Activity Monitor D M R(2)
- Q(1)
- b. Iodine Sampler W(4) N.A. N.A. N.A. *
- c. Particulate Sampler W(4) N. A. N.A.
- N.A.
- d. Effluent System Flow Rate Monitor D N.A. R
- Q g)
- e. Sampler Flow Rate Monitor D N.A. R Q
- D)
G
< {
j v1
*2
%{H
% S 4
r
TABLE 4.3.7.10-1 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIPEMENTS i c CHANNEL MODES IN WHICH 5 CHANNEL SOURCE. CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED [
- 3. UNIT 2 VENT RADIATION MONITOR
- a. Noble Gas Activity Monitor D M R(2) Q(1) g
- b. Iodine Sampler W(4) N.A. N.A. N.A.
* ^
- c. Particulate Sampler W(4) N.A. N.A. N.A.
G
- d. Effluent System Flow Rate' Monitor D N.A. R Q b
3
- e. Sampler Flow Rate Monitor D N.A. R Q
{ -t) Y P
$ 4. HEATER BAY /TUR8INE BUILDING VENT Q RADIATION MONITOR
- a. Noble Gas Activity Monitor D M R(2) Q(1)
- b. Iodine Sampler W(4) N.A. N.A. N.A.
- c. Particulate Sampler W(4) N.A. N.A. N.A.
- d. Effluent System Flow Rate Monitor D N.A. R Q hA%
- e. Sampler Flow Rate Monitor D N.A. R Q p7 O
- 5. MAIN CONDENSER OFFGAS HYDROGEN MONITOR $th 3N '
- a. Hydrogen Monitor D N.A. Q(3) M **
.gr
} ,
P'& N r
Py- cezjpeg _ jgggt i L A h'itchme f- lL PL 9 e 5L7 o f T7 ' y TABLE 4.3.7.10-1 (Continued) RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION i
~* At all times. !
** During main condenser offgas treatment system operation.
i (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annuciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint. !
- 2. Instrument indicates a downscale failure.
- 3. Instrument controls not set in operate mode.
(2) The initial CHANNEL CALIBkATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that par-ticipate in measurement assuracce activities with NBS. These standards , shall permit calibrating the system over its intended energy and measure-ment 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. One volume percent hydrogen, balance nitrogen, and i
- 2. Four volume percent hydrogen, balance nitrogen.
(4) The iodine cartridges and particulate filters will be changed at least once per 7 days. Performance of this CHANNEL CHECK does not render the system inoperable, and the applicable ACTION statements need not be antered.
\ c De le +c edire ge O
P 4 1ERRY " NIT 1 L -3/12153 -
l'Y- CEI/N2/C - /655 L A Huh mavd 2 Pa-yec RZr o f 87 . /4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS
#6 CONCENTRATION P9E
\
LIMITING CONDITION FOR OPERATION , i 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTEP AREAS (see Figure 5.1.1-1) shall be limited to < the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited - to 2 x 10 4 microcuries/ml total activity. APPLICABILITY: At all times. 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 REQUIREMENTS 4.11.1.1.1 The radioactivity content of each batch of radfor.ctive liquid waste shall be determined prior to release by sampling and analysis in accord-ance with Table 4.11.1.1.1-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1. 4.11.1.1.2 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 4.11.1.1.1-1. The results of the radioactivity analysis 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 Specification 3.11.1.1. 4.11.1.1.3 Continuous releases of radioactive liquid effluents shall be sampled and analyzed in accordance with Table 4.11.1.1.1-1. 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 Specification 3.11.1.1. ;
\
PERRY - UNIT 1 3/4 11-1 _
fY-kEI/N2A ~ /655 L A//aicA rue >1 b al
/ De(E f C Entif G pge / e JL90[ F 7
. TABLE 4.11.1.1.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimus Type of Lower Limit Liquid Release Sampling Analysis Activity of Detection Type Frequency Frequency Analysis (LLD)
(pCi/ml)*
~7 A. Batch Waste P P Principa) Gamma 5x10 Releage Each batch Each Batch Emitters Tanks 1x10
-6 1-131
-5 P M Dissolved and 1x10 One Batch /M Entrained Gases (Gamma emitters) 1x10
-5 P M H-3 b
Each Batch Composite
-7 Gross Alpha 1x16
~0 P Q Sr-89, Sr-90 5x10 b
Each Batch Composite
-6 Fe-55 1x10
~7 B. Continuogs D W Principa 5x10 b
Releases f Grab Sample Composite 'f Emitters} Gamma RHR Heat Exchanger -6 I-131 1x10 ESW Outlet -5 M M Dissolved and 1x10 Grab Sample Entrained Gases (Gamma Emitters)
-5 D M H-3 1x10 b
Grab Sample Composite Gross Alpha 1x10
-8 D Q Sr-89, Sr-90 5x10 b
Grab Sample Composite -6 Fe-55 1x10 RRY - UNIT 1 3/4 11-2
ev- czrnva-ass t. NACJmm[y , l@C 30 ogC 27 TABLE 4.11.1.1 1-1 (Continued) RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM i TABLE NOTATION
- a. The LLD is 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 particular measurement system (which may include radiochemical separation): 4.66 s b LLO = E - V 2.22 x 106 - Y - exp(-Aat) where LLD is the "a' priori" lower limit of detection as defined above (as pCi per unit mass or volume). s is the standard deviation of the background counting rate or of b 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) t 2.22 x 108 is the number of disintegrations per minute per microcurie Y is the fractional radiochemical yield (when spplicable) A is the radioactive decay constant for the particular radionuclide (sec 1) at is the elasped time between sample collection (or end of the sample collection period) and time of counting (sec) 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 measurement system and not as a posteriori (after the fact) limit for a particular measurement. { ; n a, (PERRY-UNIT 1 3/4 11-3
I j f Y~ C6~IlURQ- /tSSL. IIt/ttchmen t A TABLE 4.11.1.1.1-1 (Continu O') C ~5 / o [ P 7 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l TABLE NOTATION (Continued) I
- b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. i
- c. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling,
- d. The principal gamma emitters for which the LLD specification applies j exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, ,
but with an LLD of 5x10 6 This list does not mean that only these * 'I nuclides are to be detected and reported. Other peaks which are measurable and identifiable, tocether with the above nuclides, shall also be identified k and reported in the annual Radioactiva Effluant Ralaaca Rannet norsuant to Specification 6.9.1.7 in the format outlined in Regulatory Guide 1.21, l Appendix B, Revision 1, June 1974.
- e. 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. Sampling / Analysis of RHR Heat Exchanger is only applicable when there is ESW flow thru the RHR Heat Exchanger. l f. Sampling and analysis is required of the RHR heat exchanger ESW outlet every 12 hours when the samples indicate levels greater than LLD. I l l I A Lc+c a ni ve. pege-PERRY - UNIT 1 3/4 11 4 Amendment fio. 49
PY-LEZ/AJA&/655L ANac.hMenf- R r Ae 9 5a d n RADIOACTIVE EFFLUENTS
~
YC CYlhYrc f6. G LIMITING CONDITION FOR OPERATION i 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED j AREAS (see Figure 5.1.1-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to Re total body and to less than or equal to 5 mrems to any organ, and
- b. During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 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 witHn 30 days, pursuant to Specifica-tion 6.9.2, a Special Report which identifies the cause(s) for '
exceeding the limit (s) and :.; fines the corrective actions that have been taken to reduce the releases and the corrective actions to be e taken to ensure that future releases will be in compliance with the above limits.
- b. The provisions of Specification 3.0.3 are not applicable. ,
SURVEILLANCE REQUIREMENTS I 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters of the ODCM at least once per 31 days. i QERRY-UNIT 1 3/4 11-5 Amendment No.30
PY- C6Z/Nfl- /t>55 t AHac-bMen f & ' RADI0 ACTIVE EFFLUENTS WO ( LIQUID RADWASTE TREATMENT SYSTEM b ele +e, en % g _ LIMITING CONDITION FOR OPERATION > 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of the systes shall be used to reduce the release of radioactivity when the projected doses due to the liquid effluent from each reactor unit to - UNRESTRICTED AREAS (see Figure 5.1.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ, in a 31-day period. APPLICABILITY: At all times. ACTION: C
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, and any portion of the liquid radwaste treatment system not in operation, prepare and submit to the Commission, within 30 days pursuant to Specification 6.9.2, a Special Report which includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, and the reason for the inoperability, and
( 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specification 3.0.3 are not applicable.
t SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with method-ology and parameters in the 00CM. 4.11.1.3.2 The installed LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2. pRRY-UNIT 1 3/4 11-6 Amendment No.30
~
I
i P Y- C&Z/Nef--/655c._ l AN' etch men / R 3/% // RADIOACTIVE EFFLUENTS # 3/4.//./ LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1[Thequantityofradioactivematerialcontainedinanyoutsidetempo-rary tank, not including liners for shipping radwaste, shall be limited to ll , less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICAEILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any of the above .
specified tanks e;:.ceeding the above limit, immediately suspend all additions of radioactive material to the tanks and within 48 hours reduce the tank contents to within the limit, and describe the events leading to the condition in the next ~ Annual Radioactive EffluentReleaseReportpursuanttoSpecificationg.^.1.0."
- b. L 6,9,/.7 The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1[The quantity of radioactive material contained in each of the above ll specified tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. O 3/4/ 11-) PERRY - UNIT 1 3/4 11 m7" Amendment NoJ9, 49
P Y- CEI/uet -/6 SS L Machment et RADI0 ACTIVE EFFLUENTS i (, 3/4.11.2 GASEOUS EFFLUENTS 1 DOSE RATE DC[ete qghe p f LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous i effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) 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 to the skin, and
- b. For all iodine-131, iodine-133, tritium and all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
i APPLICABILITY: At all times. ACTION: With the dose rate (s) exceeding the above limits, immediately decrease the release rate (s) to within the above limit (s). _ i SUREVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be , determined to be within the above limits in accordance with the methodology and parameters of the ODCM. t 4.11.2.1.2 The dose rate due to iodine-131, iodine-133, tritium and to radionuclides in particulate form with half lives greater than 8 days in } gaseous effluents shall be determined to be within the above limits in accord- I ance with the methodology and parameters of the ODCM by obtaining representa-tive samples and performing analyses in accordance with the sampling and anal-ysis program specified in Table 4.11.2.1.2-1. PERRY - UNIT 1 3/4 11-8
-[
TABLE 4.11.2.1.2-1 Cf y 9-55 RADI0 ACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM . (0 E LOWER LIMIT OF 5 SAMPLING ANALYSIS
% GASEOUS RELEASE PATH FREQUENCY TYPE OF DETECTION (LLD)(*) 3 FREQUENCY ACTIVITY ANALYSIS (pCi/mL) 2 A. Drywell and Each PURGEID) Each PURGEID)
Primary Containment and VENT and VENT Principal Gamma Emitters (*) 4 y PURGE and VENT Grab Sample 1x10 p M M H-3 -6 Grab Sample 1x10 B. Offgas Vent, Unit 1 M(b) -4 Vent, Unit 2 Vent Grab Sample M( ) Principa Emitters gg 1x10 and Turbine w Building / Heater 1 H-3 1x10
-6 Bay Vent
- C. All Release Paths Continuous (d) y(c) I-131 -12 as listed in 1x10 Charcoal Sample I-133 1x10
-10 B above ContinuousId) W CC) Principal Gamma Emitters (*) -11 1x10 Particulate Sample Continuous (d) M Gross Alpha 1x10 ~11 Composite Par-ticulate Sample )h Continuous (d) Q Sr-89, Sr-90 1x10 -11 wf)f Composite Par-ticulate Sample g }h h e, q
! O d Continuous (d) Noble Gas Noble Gases -6 l 1x10 g9 Monitor (f) Gross Beta or Gamma (Xe-133 equivalent l- gp l q 1 h l $ n
f Y- CEZ/Af/2K- 1655 L ' AMAc)1m eri /- A beteie enhee Pa$ Pacye 37 of 47 TABLE 4.11.2.1.2-1 (Continu ' (' RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM i TABLE NOTATION a. The LLD is 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 l a blank observation represents a "real" signal. ~ for a particular measurement system (which may include radiochemical separation): 4.66 s b LLD = E - V 2.22 x 106 - Y - exp(-Aat) where LLD is the "a priori" lower limit of detection as defined above (as pCi per unit mass or volume). ' s is the standard deviation of the background counting rate or of b 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 108 is the number of disintegrations per minute per microcurie Y is the fractional radiochemical yield (when applicable) A is the radioactive decay constant for the particular radionuclide - (sec 1) at is the elasped time between sample collection (or end of the I sample collection period) and time of counting (sec) Typical values of E, V, Y and at should be 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 measurement system and not as a posteriori (after the fact) limit for a particular measurement. i i l k PERRY - UNIT 1 3/4 11-10 (
- _ _ _ =
r 7 - L&l/NKK-/bSS L A do cA men f ;L ' bete +c en toc POc FeV)6 3df of E7 TABLE 4.11.2.1.2-1 (Continued h
\
RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM , j TABLE NOTATION (Continued) i
- b. Analyses shall also be performed following startup, shutdown, or a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that 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. l Sampling and analyses shall also be performed at least once per 24 hours l for at least 7 days following each shutdown, startup or THERMAL POWER
- change exceeding 15 percent of RATED THERMAL POWER in one hour. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that 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 l
be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.
- e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, i
Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissians. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified t Spe cat on 6 9.1.7 in the foWaYoutlineY fn'R*eguiaIory uiTe*T.Y1, Appendix B, Revision 1, June 1974. I *l Sampling and analysis of gaseous release points shall be performed initially l 1 f. whenever a high alarm setpoint is exceeded or whenever two or more of the alert setpoints are exceeded. If the high alarm setpoint or two or more of the alert setpoints continue to be exceeded, verify at least once per 4 hours via the radiation monitors that plant releases are below the Specification 3.11.2.1 dose rate limits and sampling and analysis shall be performed at least once per 12 hours. PERRY - UNIT 1 3/4 11-11 Amendment No. 49
PY- CEZ/NRR-/655 t-A ditcA naen (- a RADI0 ACTIVE EFFLUENTS Y DOSE - NOBLE GASES Delk enh q
~
LIMITING CONDITION FOR OPERATION -- 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times. ACTION:
- a. With the calculated air dose from the radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be I taken to ensure that future releases will be in compliance with Specification 3.11.2.2.
- b. The provisions of Specification 3.0.3 are not applicable. ,f SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations. Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
i L t 0
l'Y- LEI /N4C-/bSSL ANachmenl' & RADIOACTIVE EFFLUENTS D L d c e d Are }<tcye (,' . DOSE - 10 DINE-131, 10 DINE-133, TRITIUM AND RADIONUCLIDES IN PARTICUL TE FORM LIMITING-CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: 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-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 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 releases and the proposed corrective actions to be taken to ensure that future releases will be in compliance with Specification 3.11.2.3. I
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.3 Dose Calculations. Cumulative dose contributions from iodine-131, iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days for the current calendar quarter and current calendar year shall be determined in accordance with the methodology and parameters in the ODCH at least once per 31 days, i PERRY - UNIT 1 3/4 11-13 Amendment No. 30 I l l
PY- c1z/NRK -/655 L AWa cam en f 3c Py e <!/ of 87 RADI0 ACTIVE EFFLUENTS
~
GASEOUS RADWASTE (OFFGAS) TREATMENT LIMITING CONDITION FOR OPERATION 3 3.11.2.4 The GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM shall be in operation. The Charcoal bypass mode shall not be used unless the offgas post-treatment radiation monitor is OPERABLE as specified in Table 3.3.7.1-1. APPLICABILITY: Whenever the main condenser air ejector evacuation system is in operation. ACTION:
- a. With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 7 consecutive days, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
- 1. Explanation of why gaseous radwaste was being discharged without treatment, identification of the inoperable equipment or subsystems which resulted in gaseous radwaste being discharged without treatment, and the reason for inoperability,
/
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence. .
- b. The provisions of Specification 3.0.3 are not applicable.
j SURVEll. LANCE REQUIREMENTS - 4.11. .4 The readings of relevant instrumentation shall be checked at least once per 12 hours when the main condenser air ejector is in use to ensure that the gaseous radwaste treatment system is functioning. i 1 i PERRY - UNIT 1 3/4 11-14 Amendment No.30
i
)0 't ~ LEZ/A)/2-/655 L AHa chm en f 7 RADIOACTIVE EFFLUENTS
. ,{ VENTILATION EXHAUST TREATMENT SYSTEMS ! LIMITING CONDITION FOR OPERATION c wk+c e twe pay 3.11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEMS shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radio-activity when the projected dose due to gaseous effluent releases from each reactor unit to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) in a 31 day period would exceed 0.3 arem to any organ of a MEMBER OF THE PUBLIC. APPLICABILITY: At all times. FACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
i
- 1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems which resulted in gaseous radwaste being discharged without treatment, and the reason for the inoperability, i
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from each reactor unit to areas at and beyond the SITE BOUNDARY shall be projected at lea t once per 31 days in accordance with the methodology and parameters in the ODCM. 4.11.2.5.2 The installed VENTILATION EXHAUST TREATMENT SYSTEMS shall be demonstrated OPERABLE by meeting Specifications 3.11.2.1 and 3.11.2.3. I i l PERRY - UNIT 'l 3/4'11-15 Amendment No. 30
PY- CLZ/NRK ~/655 L Affachmenf R. RADI0 ACTIVE EFFLUENTS . 3/0//,2 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.[The concentration of hydrogen in the offgas treatment system shall .l be limited to less than or equal to 4% by volume. APPLICABILITY: Whenever the offgas treatment system is in operation. ACTION:
- a. With the concentration of hydrogen in the offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours.
1s t e. M e a.c h e,q 'y s. er.ifier0
- b. With the continuous monitor inoperable,Aeti'i:: ;rd : .ng 8 -
pr::: t r: & m s.pe o FiccJabo 3. .$ . 2. fcb A cz oc>tJ b ,
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.[The concentration of hydrogen in the offgas treatment system shall y be determined to be within the above limits by continuously monitoring the waste gas in the offgas treatment system whenever the main condenser evacuation system is in operation with the hydrogen monitor OPERABLE as required by ' Tat's 2.2.'.10-1 cf Specification 3.3.7.10. a e t t N 3/4 /1 -d PERRY - UNIT 1 -0/4 11-152- Amendment No.30
FY- wI/AIRK-/65S L A8Mhmen f 2 RADI0 ACTIVE EFFLUENTS bf N D[87 I[N4ll.3 MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11 3 q 3.11.2.7 3 ThereleaserateofthesumoftheactivitiesofthenoblegasesKr-85m,l Kr-87, Kr-88, Xe-133, Xe-135, and Xe-138 measured at the main condenser air ejector shall be limited to less than or equal to 358 millicuries /second, after 30 minutes decay. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3* ACTION: With the release rate of the specified noble gases at 'he main condenser air ejector effluent exceeding 358 millicuries /second after 30 minutes decay, restore the release rate to within its limit within 72 hours or be in at least HOT SHUT 00WN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS
- v. //. 3, / 3
{- 4.11.2.MAThe release rate of noble gases at the outlet of the main condenser air ejector shall be continuously monitored in accordance with Specification 3.3.7.1. 9.n.3.2 3 0.11.2.7.2/sThe release rate of the specified noble gases from the main conden-ser air ejector shall be determined to be within the limits of Specification 3 it 3 -Q.11.2. pat the following frequencies ** by performing an isotopic analysis y or a representative sample of gases taken at the discharge (prior to dilution 1 and/or discharge) of the main condenser air ejector.
- a. At least once per 31 days. j i
- b. Within 4 hours following an increase, as indicated by the Offgas i Pretreatment Radiation Monitor, of greater than 50%, after factoring !
out increases due to changes in THERMAL POWER level, in the nominal l steady state fission gas release from the primary coolant. !
*When the main condenser air ejector is in operation.
**The provisions of Specification 4.0.4 are not applicable, j Y
3/4 11-3 PERRY - UNIT 1
N- FEZ /^/M-/6SS L A Muhmen + ;;L fitye 9'S o { 87 RADIOACTIVE EFFLUENTS ~v 3/4.11.3 SOLID RADWASTE TREATMENT gg gQ LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with -1 the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements i during transit, and disposal site requirements when received at the disposal site. ; APPLICABILITY: At all times, i ACTION:
- a. With SOLIDIFICATION or' dewatering not meeting disposal site and '
shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence,
- 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 ensure that it meets burial ground and shipping requirements and (2) take appropriate administrative action to prevent
( recurrence. , L
- c. The provisions of Specification 3.0.3 are not applicable. '
SURVEILLANCE REQUIREMENTS 1 4.11.3.1 If the SOLIDIFICATION method is used, the PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g. , filter sludges, spent resins, evaporator bottoms, and sodium sulfate solutions).
- a. - If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-TION of the batch under test shall be suspended until such time as f ,
additonal test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL i PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION ' of the batch may then be resumed using the alternative SOLIDIFICATION ' parameters determined by the PROCESS CONTROL PROGRAM. F i 5 PERRY - UNIT 1 3/4 11-18 Amendment No.30 _,,,,----*Wu+*r r"" F **" '** * '
ff- CEZ/AllR-(655L AHauhouen t- 2 PAye % of d'7 [RADIDACTIVE' EFFLUENTS c SURVEILLANCE REQUIREMENTS (Continued) l
- b. If the initial test specimen from a batch of waste fails.to verify SOLIDIFICATION, the'PROCE55 CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each I consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.15, to assure SOLIDIFICATION of subsequent batches of waste.
- c. With the installed equipment incapable of meeting Specification 3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide the contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.
J
.\
D ele + c. ens e i 1 1 I l 1 I
,E"RY-t*IT
" N 3/4 11-'N
W- CEZ// JAM' ~/655 L A Ha th m e d- a
,RADI0 ACTIVE EFFLUENTS l
( 3/4.11.4 TOTAL DOSE ehe % <. LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose t.ommitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity an/ radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 areas to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 arems. APPLICABILITY: At all times. ACTION: l a. With the calculated doses from the release of radioactive materials l' in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b. , calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Com-mission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce sub-sequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above i limits. This Special Report, as defined in 10 CFR 20.405c, shall t include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that in-cludes the release (s) covered by this report. It shall also describe - levels of radiation and concentrations of radioactive material in-I volved, and the cause of the exposure levels or concentrations. If l f the estimated dose (s) exceeds the above limits, and if the release ! condition resulting in violation of 40 CFR Part 190 has not already l been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. l
- b. The provisions of Specification 3.0.3 are not apolicable. ,1 SURVEILLANCE REQUIREMENTS i 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.
4.11.4.2 If the cumulative dose contributions exceed the limits defined in 3.11.4, ACTION a, cumulative dose contributions from direct radiation from unit i
/
operation including outside storage tanks shall be determined in accordance ! with the methodology and parameters in the ODCM. j l k PERRY - UNIT 1
/
3/4 11-20 Amend nent No. 30 s _
H ~ CEZ/N&e ~ MSS L A O cA m e n d- ;t e h/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING y
'f 8 cif 87 3/4.12.I MONITORING PROGRAM ggg g, e p n. og LIMITING CONDITION FOR OPERATION '
3.12.1 as specifiedThe radiological environmental monitoring program shall be conducted in Table 3.12.1-1. APPLICABILITY: At all times. ACTION: a. With the radiological environmental monitoring program not being con-ducted as specified in Table 3.12.1-1, prepare and submit to the Com-mission, in the Annual Radiological Environmental Operating Report per Specification 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. l b. With the level of radioactivity as the result of plant effluents in , an environmental sampiing medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose of to a MEMBER OF THE PUBLIC is less than the calendar year limits Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if: concentration (1) reporting level (1)
- concentration (2) reporting level (2) + .. Q 1.0 When radionuclides other than those in Table 3.12.1-2 are detected and are the result of plant effluents, this report shall be submitted l if the potential annual dose" to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 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 Specification 6.9.1.6. Environmental Operating Report required by j
- c. With milk or broad leaf vegetation samples unavailable from one or more of the sample locations required by Table 3.12.1-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The specific locations from which samples were I unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.14, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM includ-ing a revised figure (s) and table for the ODCM reflecting the new li location (s) with supporting informaticin identifying the cause of the unavailability of samples and justifying the selection of the new l location (s) for obtaining samples. I
- d. The provisions of Specification 3.0.4 are not applicable.
*The methodology and parameters used to estimate the potential annual dose to a j/
MEMBER OF THE PUBLIC shall be indicated in this report. / PERRY - UNIT 1 3/4 12-1 Amendment No.N , 49
lY~ WZ/NRK - /g,55( A Wach men b g RADIOLOGICAL ENVIRO MENTAL MONITORING fye 't9 of 67 , l t, i l SURVEILLANCE REQUIREMENTS l 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figures in the ODCM and shall be analyzed pursuant to the requirements of \ le 3.12.1-1 and the detection capabilities required by Table 4.12.1-1. J ( D ele + e nk e pp c s e i P3" - U"IT 1 * -3/4 12 2 A
^
.m TABLE 3.12.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
- g Number of Samples y Exposure Pathway and Sampling and g and/or Sample Sample locations (3) Type and Frequency Collection Frequency of Analysis
- 1. Direction Twenty eight routine monitoring Quarterly.
Radiation f2) stations either with two or more Gamma dose quarterly. dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows: An inner ring of stations, one in i each meteorological sector, other y than those sectors entirely over
- water (N, NNE, NNW, NW, W. WNW),
in the general area of the SITE g BOUNDARY; An outer ring of stations, one in each meteorological sector, other than those sectors entirely over water (N, NE, NNE, NNW, NW, W, WNW), Q' in the 6- to 8-km range from the g c- - site; and (b The balance of the stations to be placed in special interest A areas such as population centers, nearby residences, schools, and b % in one or two areas to serve as control stations. D h[ , Ihebh t I n Nx km? 4 w _ _ , . _ - . . _ . . . . , , - . . - - - . - - . - - ~ - . ~ . . - . -. - -. - , . . - . . . - . - - . . - , - ~ . . - . . , . . - , - - - . - . . _ - - . . - - - . . _ . - - - . , _ _ _ _ . . . . . _ _ _ _ _ _ __
~'
q TABLE 3.12.1-1 (Continued) A g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM E Number of Samples F Z Exposure Pathway and Sampling and R Type and Frequency + g and/or Sample Sample Locations gy) Collection Frequency of Analysis O
- 2. Airborne 0 5
Radioiodine and Samples from five locations: Continuous sampler _Radioiodine Canister: -- Particulates operation with sample I-131 analysis weekly. ,[ Three samples from close to collection weekly, or i the three SITE BOUNDARY loca- more frequently if tions, in different sectors, of required by dust Particulate Sampler: the highest calculated annual loading. Gross beta radioactivity average ground-level D/Q; analysis following w filter change;( ) and
} One sample from the vicinity gamma isotopic analysis (4) of a community havir.g the highest of composite (by ' h* calculated annual average ground- location) quarterly.
1evel D/Q; and One sample from a control location, as for example 15 to 30 km distant and in the least prevalent wind direction.
- 3. Waterborne
- a. Surface Two samples Composite sample over Gamma isotopic analysisI *)
1-month period.(5) monthly. Composite for gy tritium analysis quarterly. M. J y' e s Fo i H 5 D t-o
% , p *i n
r
+
,q TABLE 3.12.1-1 (Continued) 5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM g
Q C6 g Number of Samples j O Q Exposure Pathway and Samplina and Type and Frequency y y and/or Sample Sample Locations (1) Collection Frequency of Analysis g
- 3. Waterborne (Continued)
- b. Drinking One sample of each of one to Composite sample I-131 analysis on each -t-three of the nearest water over 2-week period (5) composite when the dose 3 supplies that could be when I-131 analysis calculated from the consump- A affected by its discharge. is performed; monthly tion of the water is greater composite otherwise. than 1 mrem per year.( ) Com-One sample from a control posite for gross beta and I C8 "* { n gamma isotopic analyses (4)
R monthly. Composite for
- tritium analysis quarterly.
Y c. Sediment One sample from area with Semiannually. Gamma isotopic analysis I4)
- from existing or potential semiannually.
shoreline recreational value.
- 4. Ingestion
- a. Milk Samples from milking animals Semimonthly when Gamma isotopic ( ) and I-131 in three locations within animals are on analysis semimonthly when 5 km distance having the pasture; monthly at animals are on pasture; highest dose potential. If other times. monthly at other times.
there are none, then one sample from milking animals AN in each of three areas between 1M 5 to 8 km distant where doses P ' are calculated to be greater than 1 mrem per yr.(6) One
${
.1 sample from milking animals g(fuh{
at a control location n 3 15 to 30 km distant and in the T g least prevalent wind direction. m I
TABLE 3.12.1-1 (Continued)
< RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
' 'g.
g ea Number of Samples 5 Q Exposure Pathway and Sampling and +
~ and/or Sample Sample Locations gy) Type and Frequency (b Collection Frequency of Analysis
- 4. Ingestion (Continued)
- b. Fish and One sample of each commercially Inverte- and recreationally important
. Aie in season, or Gamma isotopic analysis (N O "
brates seiniannually if they on edible portions. species in vicinity of plant are not seasonal. . discharge area. t One sample of same species G in areas not influenced by R s plant discharge. M c. Food Samples of three different kinds a Products Monthly during Gamma isotopic (4) and I-131 of broad leaf vegetation growing season. } ' grown nearest each of two analysis. different offsite locations of highest predicted annual average ground level D/Q if milk sampling is not performed. One sample of each of the simi- Monthly during lar broad leaf vegetation grown Gamma isotopicI4) and I-131 growing season. analysis. 15 to 30 km distant in the least prevalent wind direction if milk sampling is not performed. - gg
/ 3
(% > \ S u% y a
;5 Y
P Q & W n
& CEZ/AM'A ~l655 L A k eA ment g ,
TABLE 3.12.1-1 (Continued)
,9ge SY o p g7 ,
l RADIOLOGICAL ENVIR0le4 ENTAL M NITORING PROGRAM TABLE NOTATIONS Sample locations are given on the figure and the table in the ODCM. (1) Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be - provided for each and every sample location in Table 3.12-1 in a table { and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radio-logical Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and salfunction of automatic sampling equipment. If specimens are unobtainable due to , sampling equipment malfunction, effort shall be made to complete correc-tive action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environ-mental Monitoring Procram civen in the ODCM. Pursuant to Specification 6.14, submit in the next Annual Radioactive Effluent Release Report documen-l tation for a change in the ODCM, including a revised figure (s) anc table for the ODCM reflecting the new location (s) with supporting information { identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location (s) for obtaining samples . (2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in adcit.on to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. (The frequency of analysis or readout for TLO systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.) (3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon 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. [f cLe+ e emh m Ptge PERRY - UNIT 1 3/4 12-7 Amendment No. 49
D~ CEI/NAR ~ /655 L , AHa cAmen + e TABLE 3.12.1-1(Continued)) C TABLE NOTATIONS (Continued) ( (4) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. (5) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. (6) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.
~,
./
\
Dele 4e ew&c page E t
^
e l 5
TABLE 3.12.1-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels
=
o w Water Airborne Par i:ulate Fish Milk Food Products - [ Analysis (pCi/1) nr Coasts sC)/m3 ) (pCi/Kg, wet) (pCi/1) (pCi/kg, wet) 4 a + H-3 2 x 10 NA NA NA NA g, 4 Mn-54 1 x 10 3 NA 3 x 10 NA NA 2 4 Fe-59 4 x 10 NA 1 x 10 NA NA 3 4 Co-58 1 x 10 NA 3 x 10 M NA 2 U[ Co-60 3 x 10 NA 1 x 10 4 NA NA 2 Zn-65 3 x 10 NA 2 x 10 NA NA 2 Zr-Nb-95 4 x 10 NA NA NA NA I-131 2 0.9 NA 3 1 x 10 2 Cs-134 30 10 1 x 10 3 60 1 x 10 3-3 2 x 10 3
^
Cs-137 50 20 2 x 10 70 2 2 Ba-La-140 2 x 10 NA NA 3 x 10 NA a For drinking water samples. This is a 40 CFR Part 141 value. O k
& .Is O ?
-b T
?P{.
W r
- ~
_ . . ~ . _ _ _ _ TABLE 4.12.1-1 l A MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)(a), (b), (c) E y IN ENVIRONMENTAL SAMPLES
- h. Airborne Particulate Broad Leaf ) ty
- or Gas Fish Vegetation Water Milk Sediment f_.
~
Analysis (pCi/1) (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pC1/kg, wet) (pCi/kg. dry A y 3 Gross beta 4 1 x 10 -2 NA NA NA NA A h H-3 2000* NA NA NA NA NA f e f f2- 54 15 NA 130 NA NA NA Fe-59 30 NA 260 NA NA NA u, Co-58,60 15 NA 130 NA NA NA D
~ Zn-65 30 NA 260 NA NA NA 7
a; Z r-95 30 NA NA NA NA NA nb-95 15 NA NA NA NA NA I-131 1** 7 x 10 ~2 NA 1 60 NA Cs-134 15 5 x 10 -2 130 15 60 150 Cs-137 18 6 x 10
-2 150 18 80 180 t
i Ba-140 60 NA NA 60 NA NA gAN
-M La-140 15 NA NA 15 NA NA p I l
l I
*If no drinking water pathway exists, a value of 3000 pCi/1 may be used.
q$ l **If no drinking water pathway exists, a value of 15 pCi/1 may be used. 43 o a i N
~
b m
.tA m
0 l
. -----.~ . - .- - - ~. - . -. --
l' Y- WZ/A/LL -/6SS L. - ANtLChmcnF 9, WC *{'*
& P e Sg eng g7 ,
TABLE 4.12.1-1 (Continued) h* MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) TABLE NOTATION
" Acceptable detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13. ;
b Table 4.12-1 indicates acceptable detection capabilities for radioac ive materials in environmental samples. These detection capabilities are tabulated ) ! in terms of the lower limits of detection (LLDs). The LLD is defined, for i purposes of this guide,-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. i' { For a particular measurement system (which may include radiochemical separation): \' t 4.66 s ' D= E - V 2.22 - Y - exp(-AAt) where ! .( LLD is the "a priori" lower limit of detection as defined above (as pCi per unit mass or volume). s b is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute) , E is the counting efficiency (as counts per disintegration) < V is the sample size (in units of mass or volume) . 2.22 is the number of disintegrations per minute per picoeurie Y is the fractional radiochemical yield (when applicable) A is the radioactive decay constant for the particular radionuclide ' l At is the elasped time between sample collection (or end of the sample collection period) and time of counting
}
The value of sb used in the calculation of the LLD for a particular t measurement sy5 tem should be based on the actual observed variance of the ! background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicated variance.
~
i ' 4 a PERRY - UNIT 1 3/4 12-11 i
f'Y- CEI/NM-ff55 L A Hachman f 9 _ fage 57 c} 87 TABLE 4.12.1-1 (Continued) j. MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) i TABLE NOTATION (Continued) r ; 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 i fact) limit representing the capability of a measuremeiit system and not as a posteriori (after the fact) limit for a particular measurement. Occasionally background fluctuations, unavoidable small sample size, the presence of inter- i fering nuclides, or other uncontrollable circumstances may render these LLDs i unachievable. In such cases, the contributing factors should be identified ! and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. C This list does not mean that only these nuclides are to be considered. Other ! peaks that are identifiable, together with those of the above nuclides, shall , also be analyzed and reported in the Annual Radiological Environmental Oper- . ating Report pursuant to Specification 6.9.1.6. l Delete enke ge i ( s k i t f r i i 0
-P[RRY - U"!T-1 -3/4 12 12 l
_ _ _ _ _ _ _ _ . _5
PY- ce.t/un-/sss c Nch m er7f- % RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS e LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a dis-tance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden
- of greater than 50 m2 (500 ft 2) producing broad leaf vegetation. i APPLICABILITY: At all times.
ACTION:
- a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently ,
being calculated in Specification 4.11.2.3, identify the new location (s)* in the next Annual Radioactive Effluent Release l Report, pursuant to Specification 6.9.1.7.
- b. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which milk and/or broad leaf vegetation samples are currently being obtained in accordarce with Specification 3.12.1, add the new location (s) to the radiological environmental monitoring program within 30 days. If no milk anJ/or broad leaf vegetation samples are identified in the new sector with the highest D/Q value, then the next sector with the highest 0/Q value will be considered and so on until a sampling location can be established. The sampling location (s): excluding the control station location, having the lowest calculated 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 land use census was conducted.* Identify the new location (s) in the next Annual Radioactive Effluent Release Report and also include in the report a ravised figure (s) and table (s) for the ODCM reflecting the new location (s).
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at ( least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of tha land use census shall be , included in the Annual Radiological Environmental Operating Report pursuant to l Specification 6.9.1.6. l
- 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 sec- ! tors with the highest predicted D/Qs in lieu of the garden census. Specifica- I tions for broad leaf vegetation sampling in Table 3.12.1-1 shall be followed, including analysis of control samples. PERRY - UNIT 1 3/4 12-13 Amendment No.,39,49
D- MI/NFA-/CnSS L Afl&th men f R. Avje (of o f 87 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 emp;<e INTERLABORATORY COMPARISON p$e PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials that correspond to samples required by Table 3.12.1-1. These saterials are supplied as part of an Interlaboratory Comparison Prograa which has been approved by the t Commission. APPLICABILITY: At all times. ACTION:
- s. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
- b. The provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
e
-pEenv - tin i @ -3/t 12-1 C Amendment No. 30
fY - cEr/ Add-/655 L : As%)vr7W k l P e hot 6F D INSTRUMENTATION (~ SASES , l 1 MONITORING INSTRUMENTATION (Continued) 3/4.3.7.8 LOOSE-PART DETECTION SYSTEN f The OPERA 8ILITY of the loose part detection system ensures that sufficient capability 'is available to detect loose metallic parts in the primary system . .f i and avoid or mitigate damage to primary system components. The allowable out-l of-service times and surveillance requirements are consistent with the recommen- I dations of Regulatory Guide 1.133, " Loose Part Detection Program for the Primary j System of Light-Water-Cooled Reactors," May 1981. c> e at c4 ___
/ WM ,
3/4.3.7.9 AJRADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION
] j fand control, The radioactive ifquid effluent instrumentation is provided to monitor as applicable, the releases of radioactive materials in liquid i
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 alare/ 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 6 '
,6_3, and 64 of Appendix A to 10 CFR Part 50. '
/4.3.7.10 RADIOACTIVE GASEQUS EFFLUENT MONITORING INSTRtMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm ,
setpoints for these instruments shall be calculated in accordance with the i procedures in the ODCM to ensure that the alarm will occur prior to exceeding { the limits of 10 CFR Part 20. This' instrumentation also includes provisions ' for monitoring the concentrations of potentially explosive gas mixtures in the ) GASEOUS RADWASTE TREATMENT SYSTEM. TheOPERASILITYanduseofthisinstrumen-I tationisconsistentwiththerequirementsofGeneralDesignCriteria60,63] and 64 of Appendix A to 10 CFR Part 50.
\
3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM Re.ILu gmc + n ahD i
)
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 PLANT SYSTEMS ACTUATION INSTRUMENTATION t The plant systems actuation instrumentation is provided to initiate action l of the containment spray system in the event of a LOCA with high containment ' PERRY - UNIT 1 8 3/4 3-6
. . . . . . . ~ . - . .- .-. . .. .
PV- Cer/NPR-f6SSL AWa.cA neet f x l fye 65 of 87 l INSERT FOR PAGE B 3/4 3-6 l 3/4.3.7.10 MAIN CONDENSER OFFGAS TREATHENT SYSTEM EXPLOSIVE GAS MONITORING l INSTRUMENTATION The main condenser offgas treatment system explosive gas monitoring , instrumentation is provided to monitor the concentrations of potentially ; explosive gas mixtures in the GASEOUS RADVASTE TREATMENT SYSTEM to ensure the j concentration is maintained below the flammability limits of hydrogen. ' The OPERABILITY and use of this instrumentation to maintain the concentration l of hydrogen below its flammability limit provides assurance that releases of ! radioactive material vill be controlled in conformance with the requirements .i' of General Design Criterion 60 and 63 of Appendix A to 10 CFR Part 50. f i i f t i i I
PY - wr/Nff-/055 L '
- Blac Amen f DL -
Pye 6Y 9 f 57 1 RADI0 ACTIVE EFFLUENTS
-Q g@q MC
[ BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification 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 II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will ! result in exposures within (1) the Section II.A design objectives of Appendix I, ; 10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope , and its MPC in air (submersion) was converted to an equivalent concentration , in water using the methods described in International Commission on Radiological P lrotection(ICRP) Publication This specification applied to the release2.of radioactive materials in liquid effluents from all units at the site. The required detection capabilities for radioactive materals in liquid l
$ waste samples are tabulated in terms of the lower limits of detection (LLDs). l Detailed discussion of the LLD, and other detection limits, can be found in:
{ o (1) Currie, L. A.. , " Lower Limit of Detection: Definition and Elabora-tion of a Proposed Position for Radiological Effluent and Environmental
}
Measurements," NUREG/CR-4007 (September,1984). (2) HASL Procedures Manual, HASL-300 (revised annually). . 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II. A, l III.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for i Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I which assure that the releases of radioactive material in liquid effluents to UNRESTRICTED i AREAS 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 4 operations, there is reasonable assurance that the operation of the facility i will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses ! PERRY - UNIT 1 B 3/4 11-1
' l
\
PY- CEZ/ Add-/655'L ANac Amen l 9-Rupe b 5 c> f' 2 7 3/4.Il RADI0 ACTIVE EFFLUENTS BASES h/4.11.1.2 DOSE (Continued)
~
%MN '
due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Cal-culation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dis-persion of Effluent from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever l' quid effluents require treatment prior to release to the environment. l The requirement that the appropriate portions of this system be used when
; specified provides assurance that the releases of radioactive materials in i
liquid effluents will be kept "as low as is reasonably achievable." This speci-f fication implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limit governing the use of appropriate portions of the liquid radwaste treatment system were speci-fled as a suitable fraction of the dose design objectives set forth in I Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. This specification applies to the release of liquid effluents from each
, reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing the system.
3/4.11.1 [ LIQUID HOLDUP TANKS l The tanks listed in this specification include all those outdoor tanks con-taining radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and that do not have 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, I the resulting concentrations would be less than the limits of 10 CFR Part 20, ! Appendix B, Table II, Column 2, at the nearest potable water supply and the ' nearest surface water supply in an UNRESTRICTED AREA. PERRY - UNIT 1 D 3f h -g__ ' B ~5/4 ll-1 l _ - l
N - (WMA- /bSS L m/aened- e Qe 66 of 67 ' RADI0 ACTIVE EFFLUENTS gg
+' BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE s This specification is provided to ensure that the dose any time at and
\
beyond the SITE BOUNDARY from gaseous effluents from all units on the site will - be within the annual dose limits of 10 CFR Part 20 for UNRESTRICTED AREAS. The annual dose rate limits are those associated with the concentrations of those MPCs as described in Regulatory Guide 1.109. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE B0UNDARY, to annual average concentrations exceeding the Ifmits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to ! compensate for any increase in the atmospheric diffusion factor above that for I the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with 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 BOUNDARY to less than or equal to 500 mrems/ year to the total body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also l 7 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 arems/ year. This specification applies to the release of radioactive materials in gaseous effluents from all reactors at the site. The required detection capa-bilities for radioactive material in gaseous waste samples are tabulated in l terms of the lower limit of detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in: (1) Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984). (2) HASL Procedures Manual, HASL-300 (revised annually). 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Sections II.B. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix : to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calcu-lational procedures based on models and data such that the actual exposure of PERRY - UNIT 'l B 3/4 11-3
W (M /A/ff-/&Sn Mausme,& g. Re b? of 87 ( RADI0 ACTIVE EFFLUENTS D el d e e M W e prtc p BASES 3/4.11.2.2 DOSE - NOBLE GASES (Continued) a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revi-sion 1, July 1977. The ODCH equations provided for determining the air doses at and beyond the SITE BOUNDARY are made using meteorological conditions con-current with the time of release of radioactive materials in gaseous effluents or are based upon the historical average atmospheric conditions. This specification applies to the release of radioactive materials in gaseous effluents from each reactor at the site. For units with shared rad-waste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3/4.11.2.3 DOSE - 10 DINE 131, IODINE-133, TRITIUM AND RADIONUCLIDES IN
- i. PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting j Conditions for Operation are the guides set forth in Section II.C of Appendix I.
The ACTION statements provide the required operating flexibility and at the I same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements imple-ment the requirements in Section III. A. of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a HEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regu-latory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1 July 1977. These equations also provide for determining the actual doses using meteorological conditions concurrent with the time of' release of radio-l active materials in gaseous effluents or are based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium and radionuclides in particulate form are dependent on the existing radionuclide pathway to man in the areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these PERRY - UNIT 1 B 3/4 11-4 -
N ~ WUNO - /655 L ;
- kNtcAmegf87 e Mo RADI0 ACTIVE EFFLUENTS
/ l BASES
\ ,
3/4.11.2.3 DOSE - 100lNE-131 IODINE-133. TRITIUM AND RADIONUCLIDES IN l PARTICULATE FORM (Continued) } ; calculations were: (1) individual inhalation of airborne radionuclides, j (2) deposition of radionuclides onto green leafy vegetation with subsequent > consumption by man, (3) deposition onto grassy areas where milk animals and j meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man. 1 This specification applies to the release of radioactive materials in f i gaseous effluents from each reactor at the site. For units with shared rad-waste treatment systems, the gaseous effluents from the shared system are . I proportioned among the units sharing that system. t k 3/4.11.2.4 and 3/4.11.2.5 GASEOUS RADWASTE TREATMENT (0FFGAS) SYSTEM AND
; VENTILATION EXHAUST TREATMENT SYSTEMS '
i The OPERABILITY of the GASEOUS RADWASTE TREATNENT (0FFGAS) SYSTEM and the
'. VENTILATION EXHAUST TREATMENT SYSTEMS ensures that the systems will be available
{ for use whenever gaseous effluents require treatment prior to release to the '
; environment. The requirement that the appropriate portions of the systems be
) used, when specified, provides reasonable assurance that the releases of i radioactive materials in gaseous effluents will be kept "as low as is reason-(- I ably achievable." This specification implements the requirements of 10 CFR
! Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and
, the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.
lThespecifiedlimitsgoverningtheuseofappropriateportionsofthesystemswe
; in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.
This specification applies to the release of radioactive materials in j
- i 1
' gaseous effluents from each reactor at the site. For units with shared rad- '
waste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.
)
[3/4.11. EXPLOSIVEGASMIXTURE} This specification is provided to ensure that the concentration of ! : potentially explosive gas mixtures contained in the offgas holdup system is maintained below the flammability limits of hydrogen. Maintaining the concentra-tion 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. f 3N. fl. 3 ; 3/'.11.2.SMAINCONDENSER > Restricting the gross radioactivity rate of noble gases from the main f condenser provides reasonable assurance that the total body exposure to a i i MEMBER OF THE PUBLIC at and beyond the SITE BOUNDARY will not exceed a small N fraction of the limits of 10 CFR Part 100 in the event this effluent is inad-vertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and k64 of Appendix A to 10 CFR Part 50. . ,,, ,, m l
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BASES T 3/4.11.3 SOLID RADIOACTIVE WASTE ' This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, { mixing and curing times. 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the result-ant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside stor-age tanks, etc. are kept small. The Special Report will describe a course of ( action that should result in the limitation of the annual dose to a HEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Spe-
) cial Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions f 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 con-sidered to be a timely request and fulfills the requirements of 40 CFR Part 190 until tRC taff action is completed. The variance only relates to the limits of 40 .W Part 190, and does not apply in any way to the other requirements for dose li mitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which h2/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
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e '70 o f 6'7 ' 3/4.12 ' RADIOLOGICAL ENVIR0 ENTAL MONITORING - (l. N BASES , 3/4.12.1 MONITORING PROGRAM The Raotological Environmental Monitoring Program required by this specifi-cation provides representative measurements of radiation and of radioactive ' materials in those exposure pathways and for those radionuclides that lead to i the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Section IV.8.2 of . Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent ( Monitoring Program by verifying that the measurable concentrations of radio- ! ' i active materials and levels of radiation are not higher.than expected on the l basis of the effluent measurements and the modeling of the environmental exposure pathways.
'f'AssessmentBranchTechnicalPositiononEnvironmentalMonitoring, November 1979.
' The initially specified monitoring program will be effective for ,
i at least the first 3 years of commercial operation. Following this period, pro-gram changes may be initiated based on operational experience. 7 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 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined ( as an a priori (before the fact) limit representing the capability of a measure- , ment system and not as an a posteriori (after the fact) limit for a particular measurement. , Detailed discussion of the LLD, and other detection limits: ean be found in: l (1) Currie, L. A. " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984). (2) HASL Procedure Manual, HASL-300 (revised annually). ! 3/4.12.2 LAND USE CENSUS l I This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program given in the ODCM are made if !
' i required by the results of the census. The best information from door-to-door !
' survey, visual or aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Sec-tion IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gar-dens of greater than 50 mz provides assurance that significant exposure path-ways via leafy vegetables will be identified and monitored since a garden of '
i this size is the minimum required to produce the quantity (26.kg/ year) of leafy { I vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To i
- determine this minimum garden size, the following assumptions were made
- ;
(1) 20% of the garden was used for growing broad leaf vegetation (i.e. , sini- l lar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, i PERRY - UNIT 1 8 3/4 12-1
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' RADIOLOGICAL ENVIR0letENTALevdine BASES ps. E MONITORING (Delef <
3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accu-racy 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 valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. y
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I FERRY UNIT 1 S 3/' 12-2 "
PY- WZ///gg- /6 55 L AWa cAmen l- :;L faxye ~72 o f 67 1 5.0 DESIGN FEATURES
'h 5.1 SITE
. 5.1.1. EXCLUSION AREA, UNRESTRICTED AREA FOR LIQUID EFFLUENTS, AND SITE l BOUNDARY FOR GASEOUS EFFLUENTS Figure 5.1.1-1 shows the PNPP site area, including the meteorological tower. The exclusion area boundary is 2900 feet from the center line of the reactor. . All land within the exclusion area is . jointly owned by the CAPC0 Group Companies. CEI controls the exclusion area; controls include mineral rights for oil, gas, and salt. In addition, the U.S. Coast Guard provides control over the Lake Erie portion of the exclusion area. A railroad spur serves the 4 plant, heading in an east-north easterly direction from the railroad company ' right-of-way to the plant site. CEI owns the tracks and only railroad cars consigned to the PNPP are brought onto the site over this spur. Figure 5.1.1-1 also shows the liquid and gaseous effluent discharge locations as well as the plant SITE BOUNDARY for gaseous releases and the UNRESTRICTED AREA for liquid effluent releases. The in: nt: a d da n du te r;.dien tive !
= teri:1: re?::::d " ;;:::::: ef'1 : t: 'rx th: :it: t r:n t =dt;ynd ;
th: SITE "^U""*"Y :h:11 5: li;ited en ;n..t te Opnificet'er. 3.11.2.1,
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rel:n:d ir lia:id :f'l =t: t: U""!!T"!CTE0 '"5^.5 : hell 5: li;it:d pr;;=t te Spai finti= 3.11.1.1. 4 LOW POPULATION ZONE ' 5.1. 2 The low population zone shall be as shown in Figure 5.1.2-1. 5.; CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a vertical right cylinder and an ellipsoidal dome. Inside and at the bottom of the primary con-tainment is a reinforced concrete drywell composed of a vertical right cylinder and a steel head which contains an approximately 18'3" deep water filled sup-pression pool connected to the drywell through a series of horizontal vents. The primary containment has a minimum net free air volume of 1,160,000 cubic feet. The drywell has a minimum net free air volume of 276,500 cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 The containment and drywell are designed and shall be maintained for:
- a. Maximum internal pressure:
Drywell 30 psig. 6@. 1.
- 2. Containment 15 psig.
PERRY - UNIT 1 5-1 I l
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.-ADMINISTRATIVE CONTROLS A1ta d e t 2. - j A 7 73427' ;
6.8 PROCEDURES / INSTRUCTIONS AND PROGRAMS (Continued)
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;b. In-Plant' Radiation Monitorina _ l
_.. i A program'which will ensure the capability to accurately determine ' the airborne' iodine concentration in vital ~ areas'under accident conditions. This program'shall include the following:
- t. l.. Training of personnel;
- 2. Procedures for. monitoring, and 3
- 3. Provisions for maintenance of sampling and analysis equipment.- ,
lc. Post-accident S = lina i A program which will ensure the capability.to obtain and analyze reactor coolant, radioactive iodines and particulates in. plant gaseous effluents, and containment" atmosphere' samples under accident ! The program shall include the following: conditions. i i
- 1. Training of personnel, '
- 2. Procedures for sampling and analysis, and
- 3. Provisions for maintenance of sampling and analysis equipment. ,
"IN s E.A T' hTT A c HE-D . T_% s. A, Q c._ , ((- i; 6.9 REPORTING REOUIREMENTS ROUTINE REPORTS 6.9.1 In addition.to the applicable reporting requirements of Title 10, Code ,
of Federal Regulations, the following reports shall be submitted to the i Nuclear Regulatory Commission pursuant to 10 CFR 50.4 unless otherwise noted. _l STARTUP REPORT j 6.9.1.1 A summary report of plant startup and power escalation testing shall i t be submitted following (1)' receipt of an Operating License, (2) amendment to ' the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have'significantly 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 the Final Safety Analysis Report Subsection 14.2.12.2 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 design predictions and specifications. Any corrective actions'that were required to , obtain satisfactory operation shall also be described. Any. additional specific details require.d 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 com- ; pletion of the startup test program, (2) fr0 days following resumption or commencement 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. PERRY - UNIT 1 6-17 Amendment No. 77,66
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na' ckment a Qe 7tt of 87 INSERT FOR PACE 6-17, SECTION 6.8.4
- 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 MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) thall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology +
in the ODCM,
- 2) Lf 'tations on the concentrations of radioactive material rr , sed in liquid effluents to UNRESTRICTED AREAS conforming t< 3 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 ODCH.
- 4) Limitations on the Annual and qJarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming 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,
- 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 commitment conforming to Appendix I to 10 CFR Part 50,
') Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at and beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, T 61e II, Column 1,
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^A & cA m e g 7 b)e 75 of 87 INSERT FOR PAGE 6-17, SECTION 6.8.4 (CONTIllUED)
- 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
- 9) Liinitations 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 at and beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR I
Part 50, l
- 10) Limitations on the annual dose or dose commitment to any MEH3ER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
- e. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall l provide (1) representative measurements of radioactivity in the l highest potential exposure pathways, and (2) verification of the ;
accuracy of the effluent monitoring program and modeling of I environmental exposure pathways. The program shall (1) be contained l ! in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following: 1
- 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,
- 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and
- 3) Participation in an 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.
PY- Cer/AILL -/655 L Mahmnf&L c e '76 of 87 ADMINISTRATIVE CONTROLS A 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 I 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 dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the indi-vidual total dose need not be accounted for. In the aggregate, at least 80% o' the total whole-body dose received from external sources should be assigned to specific major work functions; and
- b. Documentation of all challenges to safety / relief valves.
I
- c. Annual reports shall also include the results of specific activity analysis in which the primary coolant exceeded the limits of Specifi-cation 3.4.5. The following information shall be included: (1) Re-actor power history starting 48 hours prior to the first samp12 in which the limit was exceeded; (2) Results of the last isotopic analy-sis for radioiodine 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 radiciodine l concentrations; (3) Clean-up 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 radiciodine isotope concen-tration in microcuries per gram as a function of time for the dura-tion of the specific activity above the steady-state level; and (5)
The time duration when the specific activity of the primary coolant exceeded the radioiodine limit. ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT gga dA akcJd T^)sEeT I . 9.1. 6 6 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted A A single submittal may be made for a multiple unit station. The submittal na should combine those sections that are common to all units at the station. This tabulation supplements the requirements of S20.407 of 10 CFR Part 20. PERRY - UNIT 1 6-18
. -. . - - - =. . . . .. . -. _ . . . .. -
f Y- CE.r/Det-/655 L i b & CI1M2Mf 7 l
/qe 77 o F 87 i INSERT FOR PAGE 6-18 i 6.9.1.6 The Annual Radiological Env'ranmental Operating Report covering the operation of the unit during the pre ious calendar year shall be submitted before May 1 of each year. The report shall include summaries, j 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 ODCH and ( (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. , i i i I l l 1 I P l i i
' ADMINISTRATIVE CONTROLS r ? - CEZ/M'E-/b5% ,
A Arae1,m eaf- 1
-/ENUALRADIOLOGICALENVIRONMENTALOPERATINGREPORT(Continued}' &#)O ? # 07 L '
prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial' criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of the unit for at least two years prior to initial criticality in addition to the following. The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological fp g environmental surveillance activities for the report period, including a com-parison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specifica-tion 3.12.2. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all loca- ' tions specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical ' Position, Revision 1, November 1979. In the event that some individual results ; are not available for inclusion with the report, the report shall be submitted ! noting and explaining the reasons for the missing results. The missing data j shall be submitted as soon as possible in a supplementary report. i The reports shall also include the following: a summary description of i the Radiological Environmental Monitoring Program; at least two legible maps
- l covering all sampling locations keyed to a table giving distances and direc- i tions from the centerline of one reactor; the results of licensee participation !
in the Interlaboratory Comparison Program and the corrective action taken if j the specified program is not being performed as required by Specification 3.12.3; I reasons for not conducting the Radiological Environmental Monitoring Program as , required by Specification 3.12.1, and discussion of all deviations from the l l sampling schedule of Table 3.12.1-1; ciscussion of environmental sample measure- ' ments that exceed the reporting levels of Table 3.12.1-2 but are not the result ! of plant ef fluents, pursuant to ACTION b of Specification 3.12.1; and discus- ! ; Ision t e' all analyses in which the LLD required by Table 4.12.1-1 was not achievable, e ANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT ggg ggg ; f6.9.1.7 Routine radioactive release reports covering the operation of tlEeN f unit during the previous calendar year shall be submitted annually. The [ submission of reports must be no longer than 12 months.reportmustbesubmitteda "One map shall cover stations near the SITE BOUNDARY; a second shall include
,the more distant stations. ,
PERRY - UNIT 1 6-19 Amendment No. 49
P'r - CEI/N/22 -/655L AHacAmen F ;L fye 79 of E7 INSERT FOR PAGE 6-19 6.9.1.7 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted annually. The report must be submitted as specified in 10 CFR 50.4, and the time between submission of reports must be no longer than 12 months. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid vaste 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 I to 10 CFR Part 50. 1
ADMINISTRATIVE ChuusHDLS 3 r 7 - wI/A/#& /655 L A M i r_A m m + J L l 1fue / 80 of E7 tI AdNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) ( The . Annual Radioactive Effluent Release Report shall include a sum-mary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nu-clear Power Plants." Revision 1, June 1974, with data summarized on a quartarly basis following the format of Appendix 8 thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g. , LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g. , cement, urea fomaldehyde). The Annual Radioactive Effluent Release Report submitted each year shall : include a summary of hourly meteorological data collected over the previous calendar year. This annual sumary may be either in the form of an hour-by- l hour listing on magnetic tape of wind speed, wind direction, atmospheric f stability, and precipitation (if measured), or in the form of joint frequency . cistributions of wind speed, wind direction, and atmospheric stability.* This l report shall include an assessment of the radiation doses due to the l radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This report shall also include an assessment of the radiation doses from radioactive liquid and gaseous l BOUNDARY to effluents MEMBERS OF THE PUBLIC due to their activities inside the SITE (Figure 5.1.1.1-1) during the report period. All assumptions used in l 6 j making these assessments, i.e., specific activity, exposure time, and f location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used
, for determining the gaseous pathway doses. The assessment of radiation doses i
shall be performed in accordance with the methodology and parameters in the ; 0FFSITE DOSE CALCULATION MANUAL (ODCM). l 6 i Tto Annual Radioactive Effluent Release Report submitted each year shall ! also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBL]C from reactor releases and other nearby uranium fuel cycle l i
! sources, including doses from primary effluent pathways and direct radiation, l
, for the previous calendar year to show conformance with 40 CFR Part 190, i
" Environmental Radiation Protection Standards for Nu,: lear Power Operation."
Acceptable methods for calculating the dose contribution '/ rom liquid and ! gaseous effluents are given in Regulatory Guide 1.109, Rev.1, October 1977. ' t The Annual Radioactive Effluent Release Report. shall include a list l and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
\
i a
' *In lieu of submission with the Annual Radioactive Effluent Release Report.. '
the licensee has the option of retaining this summary of required meteorological 3 data on site in a file that shall be provided to the NRC upon request, f j PERRY - UNIT I 6-20 Amendment No. 49 t bew e A c-9
i M - W ^"C N r L-ADMINISTRATIVE CONTROLS A ttr 4 m W Z ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (continued) p g; g g7 The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the pt, i 0FFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifications 6.13 and 6.14, L respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.15. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2. k The Annual Radioactive Effluent Release Report shall also include the following: an explanation as to why the inoperabiitty of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.7.9 or 3.3.7.10, respectively; and description of the events hading to liquid holdup tanks exceeding the limits of Specification 3.11.1.4. MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience shall be submitted to the Nuclear Regulatory Commission pursuant to 10 CFR 50.4 on a monthly basis, with a copy to the Director, Office of Resource Management, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: (1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1. (2) The Minimum Critical Power Ratio (MCPR) for Technical Specification 3.2.2. (3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.3. (4) The Simulated Thermal Power Time Constant for Technical Specification 3.3.1. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor fuel. (The approved revision at the time reload analyses are performed shall be identified in the COLR.) The core operating limits sha.ll be determined so that all . applicable, limits (e.g., fuel thermal-mechanical limits, core' thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be submitted upon issuance for each reload cycle, to the Nuclear regulatory Commission pursuant to 10 CFR 50.4. 6-21 Amendment No. 33, AB, AB,66 PERRY - UNil 1
N- W.L /WK~&% L ! g//acAmeva / . 7 - e 80 o*f' 87 h i ADMINISTRATIVE CONTROLS V'
' RECORD RETENTION (Continued) j
- 1. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records. l 1
- m. Records of analyses required by the radiological. environmental moni- i toring program that would permit evaluation of the accuracy of the analysis at a later date. This would include procedures effective at the specified times and QA records showing that these procedures were ;
followed.- Rwh of reAcas pufeweQ b ~ cm .s &<. fc irke CFFstTE DOSE: n. A C.At.r_u MTIDfJ M AMRAL. AJ the PkoMs CotJTRol pro 6R AM . 6.11 RADIATION PROTECTION PROGRAM j 6.11.1 Procedures for personnel radiation protection shall be prepared consist-ent with the requirements of 10 CFR Part 20 and shall be approved, maintained, ! and adhered to for all operations involving personnel radiation exposure. { f 6.12 HIGH RADIATION AREA ) i 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph ! 20.203(c)(2) of 10 CFR Part 20, each hig*h radiation area in which the intensity of radiation is greater than 100 mrem /hr
- but less than 1000 mrem /hr** shall ,
be barricaded and conspicuously posted as a high radiation area and entrance i thereto shall be controlled by requiring issuance of a Radiation Work Permit j (RWP)*. Any individual or group of individuals permitted to enter such areas j shall be provided with or accompanied by one or more of the following: l l
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area. j t
- b. A radiation monitoring device which continuously integrates the i radiation dose rate in the area and alarms when a preset integrated ;
dose is received. Entry into such areas with this monitoring device l may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
- c. A health physics qualified individual f.e., qualified in accordance I with ANSI N18.1-1971, with a radiation dose rate monitoring _ device l who is responsible for providing positive control over the activi- l ties within the area and shall perform periodic radiation surveil- i lance at the frequency specified by the Plant Health Physicist. j i
I
" 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 follow-ing plant radiation protection procedures for entry into high radiation areas.
** Measurement made at 18 inches from source of radioactivity.
PERRY - UNIT l' 6-23
t' Y~ LEr/M2-/6 ss L \ A do-cant evt f 2 ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) 6.12.2 In addition to the requirements of Specification 6.12.1, areas access-ible to personnel with radiation levels such that a major portion of the body could receive in 1 bour a dose greater than 1000 arem* shall be provided with uniquely keyed locked doors or continuously guarded to prevent unauthorized entry, and the keys shall be maintained under the achsinistrative control of the Shif t Supervisor or the Plant Health Physicist. cept during periods of access by personnel under anDoors shallRWP. approved remain locked ex-For individ-ual areas accessible to personnel with radiation levels such that a major por-tion of the body could receive in I hour a dose in excess of 1000 ares
- that are located within large areas, such as the containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped device.
a warning off, conspicuously posted, and a flashing light shall be activated as 6.12.3 In addition to the requirements of Specifications 6.12.1 and 6.12.2, for individual areas accessible to personnel such that a major portion of the body could receive in I hour a dose in excess of 3000 mrem *, entry shall require an approved RWP which will specify dose rate levels in the immediate work area and the maximum allowable stay time for individrals in that area. 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 qual-ified in radiation protection procedures to provide positive exposure control over activities within the area. 6.13 PROCESS CONTROL PROGRAM (PCP) g p[dLC UI: b k A f cic1 C 6.13.1 IM56TT The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee initiated changes to the PCP: 1.
} Shall be submitted to the Commission in the Annual Radioactive Ef fluent Release Report for the period in which the change (s) was made.
This submitt 1 shall contain:
, a.
Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental , information; , b. k A determination that the change did not reduce the overall
\ conformance of the solidified waste product to existing criteria for solid wastes; and i l
s c. Documentation of the fact that the change has been reviewed and
\ found acceptable by the PORC. j ' Measurement made at 18 inches from source of radioactivity.
4 PERRY - UNIT 1 6-24 Amendment No. 49
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INSERT FOR PAGE 6-24 Changes to the PCP. i
- a. Shall be documented and records of reviews performed shall be l retained as required by Specification 6.10.3.n. This documentation -
shall contain:
- 1) Sufficient information to support the change together with the :
appropriate analyses or evaluations justifying the change (s) : and i i
- 2) A determination that the change vill maintain the overall ,
conformance of the solidified vaste product to existing ! requirements of Federal, State, or other applicable regulations.
- b. Shall become effective after review and acceptance by the-PORC and the approval of the Plant Manager.
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l'Y - c.GI/NM -(G 55 L. Mla.thenen b * ~ \ ADMINISTRATIVE CONTROLS #
^^^C!!! C^""^L ^^^C"A" ("C") (0:nt i nu;d) L
- 2. Sh:P i;;;;; ;";;tiv; up;n review :nd ::::pten:: by th: ^^" C P 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) u.c g gg ggp
/ 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM: t
- 1. \ '
Shall be submitted to the Commission in the Annual Radinactiva Ef fluent Release Report for the period in which the change (s) was { made. This submittal shall contain:
- a. Sufficiently detailed information to totally support the i rationale for the change without benefit of additional or supplemental information. Information submitted should consist ,
of a package of those pages of the ODCM to be changed with each ; page numbered and provided with an approval and data box, together ' with appropriate analyses or evaluations justifying the change (s); b. f A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PORC.
- 2. Shall become effective upon review and acceptance by the PORC.
6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS
- 6.15.1 Licensee initiated major changes to the radioactive waste systems, liquid, gaseous and solid:
- 1. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was l reviewed by the PORC. The discussion of each change shall contain:
Q
- a. A summary of the evaluation that led to the detemination that
, the change could be made in accordance wth 10 CFR 50.59;
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
" Licensee may choose to submit the information called for in this Specification as part of the annual FSAR update.
L PERRY - UNIT 1 6-25 Amendraent No 49 Dde+c
W ~ ME/NA't- /655 L Mlachorreryf L Pye 86 cf 87 INSERT FOR PAGE 6-25 Changes to the ODCH:
- a. 'Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3.*. . 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 vill maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR l Part 190, 10 CFR 50.36a, the Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent,
~
dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the PORC 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 Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCH vas made. Each change shall be identified by markings in the margin of the affected pages, i clearly indicating the area of the page that was changed, and shall .
indicate the date (e.g., month / year) the change was implemented. I
FY - LEZ///R&-/(,5S L ANackMenf c1;L f u) e 6 7 o f 2 7
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ADMINISTRATIVE CONTROLS o _ MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Continued)
- d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or ;
quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;
- e. An evaluation of the change which shows the expected maximum exposures to MEMBERS OF THE PUBLIC in the UNRESTRICTED AREA ,
and to the general population that differ from those previously estimated in the license application and amendments thereto;
- f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
- g. An estimate of the exposure to plant operating personnel as a result of the change; and
- h. Documer.tation of the fact that the change was reviewed and found l acceptable by the PORC.
l
- 2. Shall become effective upon revie.< and acceptance by the PORC.
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5
- "ERRY Uni 1 L -{-20
PY-CEI/NRR-1655 L Attcchm:nt 3 Page 1 of 2 Significant Hazards Consideration The standards used to arrive at a determination that a request for amendment involves no significant hazard considerations are included in the Commission's regulations, 10 CFR 50.92, which state that the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed change has been reviewed with respect to these three factors and it has been determined that the proposed changes do not involve a significant hazard consideration because:
- 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and alter only the format and location of programmatic controls and procedural details relative to radioactive effluents, radiological environmental monitoring, solid radioactive vastes, and associated reporting requirements. Compliance with applicable regulatory requirements vill continue to be maintained. In addition, the proposed changes do not alter the conditions or assumptions in any of the Updated Safety Analysis Report (USAR) accident analyses. Since the USAR accident analyses remain bounding, the radiological consequences previously evaluated are not adversely affected by the proposed changes. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed changes do not involve any changes to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there vill be no change in types or increase in the amounts of any radioactive effluents released offsite. Therefore, it can be concluded that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed change does not involve a significant reduction in the margin of safety.
The proposed changes do not involve any actual change in the methodology used in the control of radioactive effluents, solid radioactive vastes, or radiological environmental monitoring. These changes are considered administrative in nature, provide for the relocation of procedural details ousside the Technical Specifications, and add appropriate
PY-CEI/NRR-1655 L Attechnent 3 , Page 2 of 2 administrative controls in the Technical Specifications to provide , continued assurance of compliance with applicable regulatory requirements. These proposed changes also comply with the guidance contained in-Generic Letter 89-01.: Therefore, it can be concluded that ; the proposed changes do not involve a significant reduction in a margin of safety. l r i l I r i I h i i h l l l l I 1
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. PY-CEI/NRR-1655 L Enclosures !
1 4 i s 3 4 1 1 i l I 0FFSITE DOSE CALCULATION MANUAL And i i I PROCESS CONTROL PROGRAM .
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