ML20081F242
| ML20081F242 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/03/1991 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20081F239 | List: |
| References | |
| GL-89-01, GL-89-1, NUDOCS 9106070241 | |
| Download: ML20081F242 (269) | |
Text
{{#Wiki_filter:- _ - - - DfFINITIONS SECTION _P AG E. 1.0 DEFINITIONS 1.1 ACTION ............................. 1-1
- 1. 2 ACTUATION LOGIC TEST ....................... ..... ... ................ ..... 1-1
- 1. 3 ANALOG CHANNEL OPERATIONAL TEST ...... 1-1 1.4 AXIAL FLUX DIFFERENCE ................ ........ .........
.................. l' 1
- 1. 5 CHANNEL CALIBRATION ... ......
1-1
- 1. 6 CHANNEL CHECK ...........................................
........ ................ 1-1
- 1. 7 CONTAINMENT INTEGRITY ..........
1-2
- 1. 8 CONTROLLED LEAKAGE .................... ........ ........ 1-2 1.9 CORE ALTERATION ..................... ..................
........ ................ 1-2
- 1.9a CORE OPERATING LIMITS REPORT ... ...... 1-2 1.10 DOSE EQUIVALENT I-131 ...........,,.... .......... ......
... ..... 1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY ................. ....... ....... 1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE T
................ 1-3 1.13 FREQUENCY NOTATION ...................IME ............... 1-3 1.14 GASEOUS RADWASTE TREATMENT SYSTEM ....................... .... 1-3 1.15 IDENTIFIED LEAKAGE ............. 1-3 1.16 MASTER RELAY TEST .............. ........ ........................
............. 1-3 1.17 0FFSITE DOSE CALCULATION MANUAL 1.18 OPERABLE - OPERABILITY .......... . . . . . (00CM) . . .1-4 1-4 1.19 OPERATIONAL MODE - MODE ......
1.20 PHYSICS TESTS ........................................... 1-4 1.21 PRESSURE BOUNDARY LEAKAGE
........................... 1-4
................... 1-4 1.22 PROCESS CONTROL PROGRAM (PCP) ........................... .......... 1.
1.23 PURGE-PURGING ............. ..........................., 1.24 QUADRANT POWER TILT RATIO 1-4 1-5 1.25 RATED THERMAL POWER ............. .......................
.............. 1- 5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME ................ ..... ..... 1-5 1.27 REPORTABLE EVENT ................... 1-5 1.28 SHUTDOWN MARGIN ......................................... 1-5 1.29 SLAVE RELAY TEST ................... .......... ...................
......... 1-5
- 1. 30 00LI0III:".T!0M . . . P.EAM.TE i 1.31 SOURCE CHECK ...................... ......... ........... ---l ...
...... 1-5 1.32 STAGGERED TEST BASIS ..... . .......... ......... ....... 1-6 1.33 THERMAL POWER ............. .................... ........ . . .... 1-6 1.34 TRIP ACTUATING DEVICE OPERATIONAL TEST .. 1-6 1.35 UNIDENTIFIED LEAKAGE ....... ............................ ............... 1-6 1.36 VENTILATION EXHAUST TREATMENT SY 1-6 1.37 VENTING ........................ STEM ....................
......................... 1-6 TABLE 1.1 OPERATIONAL MODES .................................. 1-7 TABLE 1.2 FREQUENCY NOTATION ..... ....... .................. 1-8 SUMMER - UNIT 1 1 Amendment No. g , y 9106070241 910602 PDR ADOCK C6009395 P PDR
,- 4 , s A to Document Control Desk Letter TSP 890004 PROPOSED TECHNICAL SPECIFICATION CHANGE TSP-890004 V. C. SUMMcR NUCLEAR STATION F1 6 AY 2F GANGES SPECIFICATION TITLE DESCRIPTION OF CHANGE Page I DEFINITIONS solidification definition is relocated to the PCP.
~ ~ ~
Pages IV, XI, XVI INDEX Index is revised to reflect change in scope of the ODCM. 1.17 0FFSITE DOSE CALCULATION MANUAL Definition is updated to reflect the change in scope of the ODCM. 1.22 PROCESS CONTROL PROGRAM Definition is updated to reflect the change in scope of jtf.e?CP. 1.30 SOLIDIFICATION Definition is relocated to the PCP. 3/4.3.3.8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING Programmatic controls are included in 6.8.4.e. Item 1). INSTRUMENTATION . Existing specification procedural details are relocated to the ODCM. 3/4.3.3.9 RADI0 ACTIVE GASE0US EFFLUENT Programmatic controls are included in 6.8.4.e. Item 1). MONITORING INSTRUMENTATION Existing specification procedural details are relocated to the ODCM. Existing requirements for explosive gas monitoring instrumentation is retained. 3/4.11.1.1 LIQUID EFFLUENTS: CONCENTRATION Programmatic controls are included in 6.8.4.e. Items 2) and 3). Existing specification procedural details are relocatsj to the ODCM. 3/4.11.1.2 LIQUID EFFLUENTS: DOSE Programmatic controls are included in 6.8.4.e. Items 4) and'5). Existing specification procedural details are relocated to the ODCM. 3/4.11.1.3 LIQUID EFFLUENTS: LIQUID WASTE Programmatic controls are included in 6.8.4.e. Item 6) . TREATMENT Existing specification procedural details are relocated to the ODCM.
Enciv;ure 1 to Document Control Desh Letter TSP 890004 PROPOSED TECHNICAL SPECIFICATION CHANGE TSP-890004 V. C. SUMMER NUCLEAR STATION
SUMMARY
OF CHANGES TITLE DESCRIPTION OF CHANGE SPE C I F IC AT' '" GASEOUS EFFLUENTS: DOSE RATE Programmatic controls are included in 6.8.4.e. Items 3) 3/4.11.2.1 and 7). Existing specification procedural details are relocated to the ODCM. RADI0 ACTIVE EFFLUENTS: DOSE-NOBLE Programmatic controls are included in 6.8.4.e. Items 5) 3/4.11.2.2 and 8). Existing specification procedural details are GASES relocated to the ODCM. RADI0 ACTIVE EFFLUENTS: DOSE-Programmatic controls are included in 6.8.4.e. Items 5) 3/4.11.F.3 RADIOI0 DINES, RADIOACTIVE MATERIAL AND and 9). Existing specification procedural details are TRITIUM relocated to the ODCM. RADI0 ACTIVE EFFLUENTS: GASEOUS Programmatic controls are included in 6.8.4.e. Item 6). 3/4.11.E.4 Existing specification procedural details are relocated RADWASTE TREATMENT to the ODCM. RADI0 ACTIVE EFFLUENTS: SOLID Existing specification procedural details are relocated 3/4.11.3 to tne PCP. RADI0 ACTIVE WASTE RADI0 ACTIVE EFFLUENTS: TOTAL DOSE Programmatic controls are included in 6.8.4.e. Item 10). 3/4.11.4 Existing specification procedural details are relocated to the ODCM. RADIOLOGICAL ENVIRONMENTAL MONITORING: Programmatic controls are included in 6.8.4.f. Item 1). 3/4.12.1 Existing specification procedural details are relocated MONITORING PROGRAM to the ODCH. RADIOLOGICAL ENVIRONMENTAL MONITORING: Programmatic controls are included in 6.8.4.f. Item 2). 3/4.12.2 Existing specification procedural details are relocated LAND USE CENSUS to the ODCH. RADIOLOGICAL ENVIRONMENTAL MONITORING: Programmatic controls are included in 6.8.4.f. Item 3). 3/4.12.3 Existing specification procedural details are relocated INTERLABORATORY COMPARISON PROGRAM to the ODCM.
.y INDEX LIMlilNG COND1110NS FOR OPERA 110N AND-SURVEllLANC SECTION P_ AGE , 3/4.2 -POWER DISTRIBUTION LIMITS 3/4.2.1 -AXIAL FLUX 0!FFERENCE.................................... 3/4 2-1 3/4.2.2 3/4.2.3 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CNANNEL-FACT 0R.........................................3/4 2-8 3/4. 2. 4 QUADRANT POWER TILT 3/4.2.5 RATI0................................ 3/4 2-12 DNB PARAMETERS........................................... 3/4 2-15 3/4.-3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP _ SYSTEM INSTRUMEN_ TAT!0N....................., 3/4 3-1 3/4.3.2
-ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION........................................3/4 3-15 3/4.3.3 MONITORING INSTRUMENTATION Radiation _ Monitoring..................................... 3/4 3-41 Movable Incore Detectors................................. 3/4 3-46
-Seismic Instr 6 mentation..................................
Meteorological 3/4 3 '? Instrumentation........................... 3/4 3-50_ , Remote Shutdown Instrumentation...... ................... 3/4 3-53
--/p
[7- Accident Monitoring Instrumentation......................
.Ra44+asti'!: LMuid-E4f49en 3/4 3 Ins trumentat4on. . . . . . . . c.t-Mont4cr4ng h.t,gT[ ..............................- 4/44W 9:di:::t!W h ;Ef' Nent " crit: rings Ins tremontat4cn.- .. .. ..
. . . . . . . . . . - 3/4 3-73
-Loose-Part Detection Instrumentation.....................3/4-3-80
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- Gl * $ " [ "ST^
[ M 4 v m-1 ,f 7 v q - , ,u .
,v
' SUMMER - UNIT 1 IV Amendment No. 79
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 3/4-11-1
/ Cor,ttttttstion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
60 g (N 7 0 0 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .--314- 1 1 t Liquid "este Treatoent..................................'. 3/4-446-Liquid Holdup Tanks...................................... 3/4 11-7 Settling Pond............................................ 3/4 11-8 3/4.11.2 GASEOUS EFFLUENTS 0cte-Rate................................................ 3/4-11 ,/ 7 m. uv __o_m,_ r ___
. nuu . wu ........................................... ,, , ,n, 3. N,4-g,
- __2
.vose nauiv ivu nin , r.-.._.o utam,. _
one 1 "edici.uclides Other then " obit-6tw i................... 3/4-HH5-Cascotts-hthreste-Ttcetee nt. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/%44=N Explosive Gas Mixture.................................... 3/4 11-17 Gas Storage Tanks........................................ 3/4 11-18 3/4. . -ii.J au ttu MMw...m. iius nois.......................... . 1_ /44+49 3 4.ii.4 TOTAL 005E............................................... 3/^ 11-21 4 s., ,u, . s. .e n. ~m t. ge. v. r..a..i. c e. .u. e. n.n,,u. .u.e. m. .. i. u. .n u. . v. .r. n. o. v. .u..r
.)
3/4.i2.i n0 N I TO R I NG Mt00 RAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44 2H ' w ,%- m,, .m,,,, ,n ws W 1 3 u TT-w/ 7 16.4. wnITU WW6 b6asvs.... ......... ..........................
. . . . . . . . _ . . _ . . . ..., g ...m ..... ..
3 o, 33,3.wo-mJ/9.14.J 1R I C.K IUK) 4Wrnn10Vn rnVu .. . . . . . . . . . . . . . . . . . . . . wi is SUMER; UNIT 1 XI l 1 l
INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.i LIQUID EFFLUENTS........................ ........ B 3/4 ii-i I 3/4.11.2 cAsEcus trFtVEnTs............ ................... s 3/4 11-3 3/4di.3 00H9-RA0iOAsTiVE-VAST E . . . . . . . . . . . . . . . . . . . . . . . . . . . . = r7/4 ii-b-1 /4.11.4 TOTAL D03E....................... ^...................... B1/4-11'6-g - 3 /4 A P-R A010t0G18 At-ENVI RONMENT A H',0HiTORI NG
,y a/4a aa-rownoauunoorx:. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-3/4-12-3
-3/4de-2 LAND-USE-GEK5tl?...................... . ..... ... . .. B M 12-1
--BAA2. 3 IffTMt:ABORATORY-t0MPARis01t-PROGtM. . . . . . . . . . . . . . . . . . . . . . . B 3/4-12*2-9 D
s stMiER-UNIT 1 XVI
. DEFINITIONS I l 0FFSITE DOSE CALCULATION MANUAL (00CM) TIME-OFES1IEJOSE CALCULATION MANUAL shall contain th and I parameters used in the7dTcutatiefs>f-a ue to radioactive gaseou and liquid effluent cu ation of gase nd-44quiIkaf_ fluent m aito ' rm trip setpoints. ' L 5ur 1+ OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). OPERA',IONAL MODE - MODE
- 1. 15, An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive comaination of core reactivity condition, power level and average reactor coclant temperature specified in Table 1.1 PHYiICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and
- 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Comission.
PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. PROCESS CONTROL PROGRAM (PCP) 1 D he PROC contain the samni-%-enaiysis, and li;*"d'Ou"in"*s:".'" ""'"***"'"**"""" TNSMT c---> PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. SUtEER - UNIT 1 1-4
INSERT 1 0FFSITE DOSE CALCULATION MANUAL 1.17 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The 00CM shall also contain (1) the Radioactive Effluent Controls-and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8, INSERT 2 PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive waster based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10CFR20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
" DEFINITIONS I QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated output or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERML POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to ! the reactor coolant of 2775 MWt. ' REACTOR TRIP SYSTEM RESPONSE TIME ! 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coli voltage. REPORTA0LE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is suberitical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shall be the energizatirn of each slave relay and verification of OPERABILITY of each relay. The aLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. GObiOIFICAT40N No c u.> sN l 1.30 40LIC!FICfrTMM-cM1-bs-the-conseshn-sf--ndleastive-wastes from-11guid systems--tounife=1y di s + H h"+ =d . ' t M :, 8-- :9 8 :d ::1 M H +2 t " u + =
- vol=; er,d h
- pc, 5:=d:d by : St 910 Ourf:c: ;f dinke-ettsk:-en-al4-tidos ffteentandkg).
o SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive-source. SUMER - UNIT 1 1-5 Amendment No, 35 l
l INSTRUMENTATION' DI0 ACTIVE LIQUID ETFLUENT HONITORING INSTRUMENTATION L TING CONDITION FOR OPERATION ' s ._ 3.3.3.8 he radioactive liquid effluent monitoring instrumentation channelt shown in le 3.3-12 shall be OPERABLE with their alarm / trip set ints set ensure that he limits of Specification 3.11.1.1 are not exceede . The alar-trip setpoint of these channels shall be determined in accord ce with the OFFSITE DOSE C CULATION MANUAL (00CH). APPLICABILITY: A all times. ACTION:
"a. With a radioa ive liquid effluent monit g instrumentation channel alarm / trip set nt less consery tive a required by the above specification, i s
ediatelyjus[e th r lease of radioactive liquid effluents monitore by the & f e channel or declare the channel inoperable.
- b. With less than the m q. n mber f radioactive liquid effluent monitoring instrumenta i n anne s OPERABLE, take the ACTION shown in Table 3.3-12. Addi o ly/fthisconditionprevailsformore than 30 days, in the ne t m finnual effluent report explain why this condition was not cor ed a timely manner,
- c. The provisiem cifi ation 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each radioactiv iquid effluent monit ing instrumentation channel shall be demonstrated OPERABLE by performance of th CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATIAN and ANALOG CHANNEL OPERAT NAL TEST operations at the frequencies shown h Table 4.3-8.
SU194ER - UNIT 1 3/4 3-67 Amendment No. 35
E . TABLE 3.3-12 4
@ RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRl8ENTATION -
c ~~ i f MINIltsi CHANNEp * ((' INSTRUMENT:f OPERA 8tE g,'
- 1. GROSSRADIDACTh(TY IT R5 PROVIDING ALARM AND AUTOMATIC t TERMINATION OF R SE
\
- a. Liquid Radwaste E 1 _Line - RM-L5 or RM L9 1 31
- b. Nuclear (Processed 51 enerator lowdown Ef f) nt 1 31 Line RM-L7 or RM-L9 \ / l.
- c. Steam Generator Blowdown ff en -tine '
O 4
~-
- 1. Unprocessed th;rtng Power Ope la RM-L1 r AM-L3- 1 32
{ 2. Unprocessed during Startup - 32 : 1 Y d. Turbine Building Sump Effluen ine - RM-48 C 1 33
.. E
- e. Condensate Demineralizer ackwash Effluent t RM-L11 1 36
- 2. FLOW RATE MEASUREMENT DE 1 ES"
- a. Liquid Radwaste ffluent Line - Ianks I and 2 1/ tank 34
- b. Penstocks Hipi um Flow Interlock ** 1 34
- c. Nuclear Blowdown Efffuent Line 34 Steam G derator (Unprocessed) Blowdown Effluent Line 1 34 '*
d.
- 3. TANK L L INDICATING DEVICES .
s Condensate Storage Tank 1 35 f, a .- l_~
.. s'f 3.,
- fjdw' rate for the monitor RM-L9 is determined by adding flow rates for monitors St-L5 and 44-L7. ', f j; *
- Ainimum dilution flow is assured by an interlock terminating Ilquid waste releu r=
'h[p
. .A .
z d
? dilution f t r' is not available.
S
INSTRLMNTATION '
/
_TA8LE 3.3-12__(Continued) TA8Li NOTATION ACTIONQ1- With the number of channels OPERABLE less than requir d by the N Minimum Channels OPERA 8LE requirement, effluent reldses may continue for up to 14 days provided that prior to (nitiating a release:
- a. At least two independent samples are ana red in accords with Specification 4.11.1.1.3, and t
- b. At least two technically qualified abers of the Ft.111ty 5taf f independently verify the rol se rate calculations a' discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 32 - 1 With the numbe of channels OP ABLE less than required by the Minimum Channels OPERABLE req rement, effluent releases via this pathway may ntinue fp/ up to 30 days provided grab samples are analyze for ross radioactivity (beta and gamma) at a limit of de ect at least 10 7 microcuries/ gram:
- a. At least on per hours when the specific activity of the sec ndar polan is greater than 0.01 microcuries/ gram DOSE EQU T I-131.
b. At leagt o ecoolant M*p)nfary per 24 hour is 1 when the specific activity of s than or equal to 0.01 micro-tutties/ tam DOSE EQUIVALENT -131.
)/
ACTION 33 - With the ber of channels OPERABL less than required by the Minimum y arsels OPERABLE requirement, effluent releases via this p thway a.ay continue for up to 30 ys provided that, at least nce per 8 hours, grab samples are ollected and analyzed for ross radioactivity (beta and gamma) a a limit of detection of t least 10 7 microcuries/ gram. ACTION 34 - ith the number of channels OPERABLE less than cuired by the Minimum Channels OPERABLE requirement, effluent eases via this pathway may continue for up to 30 days provi the flow rate is estimated at least once per 4 hours during tual releases. Pump curves may be used to estimate flow. SUPMER - UNIT 1 3/4 3-69
N4TRtMENTATION _ j :p,, w] - /
, TABLE 3.3-12 (Continued)
TABLE NOTATION
\
ACTION 35 -
\ With the number of channels OPERABLE less than requi by the x Minimum Channels OPERABLE requirement, liquid additip6s to this
\ tank may continue for up to 30 days provided the task liquid s' prevent level is overflow.
estimated during all liquid additions to Ihe tank to ACTION 36 Witg the number of channels OPERABLE less tha, required by the minityJe conti channels OPERABLE requirement, effluent releases may in acco for e up to 30 days provided that snaples are analyzed ance with specification 4.11.1.1<1 and 4.11.1.5.
\ .
a\* T
,/
/
/
/
Sd*ER-UNIT 1 3/4 3-70 Amendment No. ")
il
~
I ! l TABLE 4.3-8 c J I RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTR' MENTATION SURVEILLANCE REQUIREMENTS E e c LOG CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL h INSTRUMENT CHECK CHECK CALIBRATI TEST
- 1. ' GROSS BETA OR GA % RADI0 ACTIVITY MONITORS PROVIDING ALARM A UTOMATIC TERMINATION OF RELEASE
- a. Iiquid Radwaste Effluen ~ e - RM-L5, RM-L9 D P R(2) Q(1)
- b. Nuclear Blowdown Effluent Line RM-L7 0/ P R(2) Q(1) l c. Steam Generator Blowdown Effluent Li -
M R(2) Q(1)
, RM-L3, RM-LIO s
[ d. Turbine Building Sump / 4 : Effluent Line - RM-L8 / M R(2) Q(1)-
- e. Condensate Demineralizer Backwash ne RM-L11 D R(2) Q(4)
- 2. FLOW RATE MEASUREMENT DEVICES r
- a. Liquid Radwaste Efflue .Line D(3) N.A. R Q
- b. Penstocks Minimum ow Interlock 0(3) N. A.' R Q l
l g c. Nuclear Blow wn Effluent Line 0(3) N.A. R Q x h d. Steam nerator Blowdown D(3) N.A. R Q g Eff ent Line
,+
z 3. T LEVEL INDICATING DEVICES
- a. Condensate Storage Tanks D N.A. R Q 4
s
INSTRUMENTATf0N TABLE 4.3-8 (Continued) TABLE NOTATION j (1)_ The solation an ANALOG of this pathway CHANNEL and _ control OPERATIONAL room alarm annunciati TEST c s n oc of the following conditions exists: 1. Instrument indicates measured levels above se al m/ trip setp
- 2. Lo of Power (alarm only).
- 3. -Low f (alarm only).
4 Instrume i le ailure larm only).
- 5. Normal ss/Bypa\ndicates switch a downs ti By ass (al m only).
- 6. Other instrumen contro s not set in operate mode.
(2) The initial CHANNEL CAL RATT shall the reference standards performed using one or more of using Qandards'that ha ified en obta by he National Bureau of Standards or in measu gm
- r. assurance)actvitie with ed from suppliers that participate NBS. These stan6.eds shall
' permit calfb tingthefystem ve measurement r' e. Ffr subsequ its intended range of energy and been related t thejinitial ca ib ation shall be used.t CHANNEL CALIBRA
'(3) CHANNEL CHECK sha 1 consist f veri of release. CHANNEL CHEC shall be m ing indication of flow during periods e at least once per 24 hours on days on which continuous periodic, or atch releases are made.
(4) The ANALOG CHANNEL-OP ATIONAL TEST shall isolation of this pa way and local panel a 1so demonstrate that a'utomatic of the following co ditions exists: rm annunciation occurs if any
- 1. Instrument dicates measured 16vels above e alarm / trip setpoint.
2. Loss of P wer (alarm only).
- 3. Low fl w-(alarm-only).
4. Ins rument' indicates a downscale failure (alarm only 5. ormal/ Bypass! switch set in Bypass (alarm only). 6. Other instrument controls not set in operate mode. l N N SUfEER - UNIT 1 3/4 3-72 Amendment Nc N.56
l . INSTRUMENTATION ygft.omv6 GM RAD 10ACiIVC-GAf40'J S EF444NT HONITORING INSTRUMENTATION LIMITING _ CONDITION FOR OPERATION 4.plmv'cyas 3.3.3.9 The radicsotMusws-off4-w-t monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the liraits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the ODCH. APPLICABILITY: As shown in Table 3.3-13 ACTION: c y b ik gaa
- a. With a-radioact4ve-gaseous-4f-f4uent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the s above Specification,,iewediately-suspend-the-release-c hradioactive-gaseous-offluents-nonttorod-by-tho-affected-channel--on declare the channel inoperable,4d Ab 4 &nWnew th Td4 3 5-13
.g g /cs M 9 4
- b. With less than the minimum number of, radioactive-gssenus of fluent monitoring instr ntation channels OPERABLE, take the ACTION shown in Table 3.3-13. dd tt4 ona 14y-44--thi s-con d i t i on -p r o v e t-l e-f o r- mo re than-3 Mays, a t54-next-sem14nneaLeffluent-report,-upb.in why
-thi b c ond iti c n4*s c - no t-c or rac te d--i n-s -timelp manne r.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. !
SURVEILLANCE REQUIREMENTS
~
.tx-Plos We cfit' 4.3.3.9 Each r.edicactive,gaucuc-afment monitoring instrumentatien channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-9. _
,n- ,~r x, - ~r
'# )
Q,fgyc fft. in}cuta Ste. cotabmalaIer? O f ' " k k c/ss b a.eut, sf ars,.wcc RA ff.nc
"*<d ~Je mih yo'cW \
L
\
dept %~ k Gmnuu:m ,/Jaew n r v p.uce;48 4 ,, 6.9. 2 c-4xMa*'n W!!o) Wu'.S i?!7'sia $ 4 & Mta x Ltcc.aw & Q s, , y[m ,f, G' O g ) n/x ann. ,
/
d , d' 'w '
/ ,/
O SUMMER - UNIT 1 3/4 3-73 Amendment No. 35
E m 3 TABLE 3.3-13 em.es.vc cas 7 P.^."!O^.CTIVE C^3EOUS EFFLUEF MONITORING INSTRUMENTATION C 2 U MINIMUM CHANNE'LS
. INSTRUMENT OPERABLE APPLICA8ILITY ACTION
- 1. MSTE G^3 "aLOUP-GYSTEM- (t4ah5*
-- e . Wble Gee ac tivity Writer - Previd!ng
^1: : :nd ^2 r:tte Tecoinett:n f Rele::e 1
- 3S
^^^^10 er *- h 1;f WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Oxygen Monitor 2 **
44 y- b. Hydrogen Monitor 1 ** 42 w I PLANT VENT EXHAUST SYSTEM h a. Noble Gas Qvity Monitor - Providing i Alarm and Automati Termination of 1
- 40 Release from Waste Gas Iduo System j RM-A3 I
- b. Iodine Sampler
- 43
- c. Particulate Sampler I
- 43
- d. Flow R asuring Device 1
- 39
- e. Sampler Flow Rate Measuring Device 1
- E TA3LE 3.3-13'(Continued)
I y RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION e 5 MINIMUM CHANNELS INSTR [ I OPERABLE APPLICABILII ACTION
- 4. REACTOR BUILDING P E SYSTEM
- a. Noble Gas Activity, nitor Providing 1
- 41 Alarm and Automatic Tehination of
' Release - RM-A4 i
l b. Iodine Sampler-
- 43
- c. Particulate Sampler ~
- 43 R d. Flow Rate Measuring Device 1
- 39' Y e. Sampler flow Rate Monitor 1 *
- 39 l
i
1 INSTRUMENTATION TABLE 3.3-13 (Continued) TABLE NOTATION
- 4t-eH-times--duMeg-eeleases-vio-this-pathway, h wu
** During waste gas holdup system operation (treatment for primary system offgases).
ACTION 38 - H ith the number of channels OPERABLE less than required by the
, (b #w tank mum Channels OPERABl.E requirement, the contents of the may be released to the environment for up to 14 d provide that p~rior to initiating the release:
- a. At lea two independent samples of the tank's contents are anal d, and b.
At least two echnically qualified membe- of the facility Staff independe ly verify the release ate calculations and discharge val lineup; Otherwise, suspend release f radioaptive effluents via this pathway. ACTION 39 -
/
Withthenumberofchannels0)PR E less than required by the {$ wyb Minimum Channels OP3ARLE rMuireme effluent releases via this pathway m6y continuyfor up to 3 ys provided the flow rate is estimated at 1 st once per 4 h . ACTION 40 1With
, the numoet of hannels OPERABLE less t Minimuin Channel ERABLE requirement, effluen equired by the
\ lesses via this pathway p continue for up to 30 days provi grab
( tOh wd.) samples arytaken at least once per 8 hours and thes samples are analyfed for gross activity within 24 hours. ACTION 41 - With.the number of channels OPERABLE less than required by
/ Hd e
( * ,#jtimumChannelsOPERABLErequirement,immediatelysuspend 0RGlNG of radioactive effluents via this pathway. ACTION 42 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner. ACTION 43 y 71th-theaumber of channels OPERABLE less than requi,rgdAy-ttfe-~~~~ (p n Minimum ChaEne'15-OPERABLE the af fected pathway may co tinue:fF up to 30 days provided samples are Jco tinuo requirementpent-Nieases via collectedwIDi' aux 4Marnsampling ytipment Ts required in Table 4.11-2. ACTION 44 - _ With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue -f-or- up to P days provided one hydreacn-analya ee-upstream-and-one -hyd r-oge nanalya e r-downstre am-a re GPHABif-or grab samples are taken and analyzed at least once per24 hours,at-the-locat4on-of-the %perable hydrogen analyrer. With both the channels inoperable, be-in-et-leest ancH0f-STANDBY Tartu and anal within C haum epma um imq c<,,rta nu prailM grsab >^m o d
" d cot (gg ence F r 4 boons etwamg epn3 opr* thu SUMMER - UNIT 3)Ihi 3 'Y*
- M
E TABLE 4.3-9 I ' EML ostvE. GAS
"'l'I0ACTI'!C CASC000 Ei[LUE%T MONITORING INSTRUMENTATICH SURVEILLANCE REQUIREMENTS , C
- 5
"' ANALOG CHANNEL MODES IN %dHICH r i
" CHANNEL -500066- CHANNEL OPERATIONAL SURVEILLANCE
!. INSTRUMENT CHECK C"EC"- CALIBRATION TEST REQUIRED
- 1. MSTE C"5 "0LDUP--GYSTEM (rd.\wwl
- a. " ble C : ^ ' b ity S niter -
-RM-Al p - o o nrn ort)
- 1,2' WASTE GA.c ' '/
~
.(STEM EXPLOSIVE
, GAS MONI , TEM w g a. Hydr .or D --N-A Q(4) M ** 4 w
- 4 b. Oxygen. .cor D --M-A - Q(5) M **
u t I m i _ _-_--____-____.-__m__-__-______._______m_
l
.i
'I t :
.}
l E l TABLE 4.3-9 (Continued) j- 9 l , ~ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE 5 R
" ANALOG C . EL MODES IN WHICH " CHANNEL -SOURCE CHANNEL OPERAT INSTRtMENT L' SURVEILLANCE CHECK CHECK CALIBRATION ST l REQUIRED' i
- 3. MAIN PLANT VENT EXHAUST SYSTEM i
- a. Noble Gas Activ Monitor -
RM-A3 D M R( - Q(2) *
- b. Iodine Sampler W .N.A. .A.
- N. A.
t
- c. Particulate Sampler -W N. N.A. N.A. * .I
{ d. . Flow Rate Measuring Device ' D N.A. R Q
- e'. ' Sampler Flow Rate Monitor -D N. A. R
- Q
- 4. REACTOR BUILDING PURGE SYSTEM
- a. Noble Gas , Activity Monitor - D P,M RM-A4 R(3) - Q(1)- *
-j
- b. Iodine Sampler W- N.A. N.A. N.A.
I
- c. Particulate 5 er W N. A. N.A. N.A. *
- d. Flow Rate asuring Device D N. A .. R *
.i
! e. Samp r Flow Rate Monitor D N. A. R
- i -Q e
l .
r INSTRUMENTATION-TABLE 4.3-9 (Continued)
-TABLE NOTATION A t o ' l t ime+r- (WY dst f During waste gas holdup system operation (treatment for primary system offgases).
(1) The OG CHANNEL OPERATIONAL TEST shall c',so des,r : rate that automa V isolation his pathway and control roem alart annunciation oc if any of the folio conditions exists:
- 1. Instrument indicates ured levels aLova,+M arm / trip setpoint.
- 2. Loss of Power (alarm only).
- 3. Low flow (alarm only).
- 4. Instrument ingica es a downscale failure (alarm o
- 5. Norm Sypass switch set in Bypass (alarm only).
Other instrument controls not set in oDerate mode. _
/
(2) The OG CHANNEL OPERATIONAL TEST shall also demonstrate that contr V 4J room alar nnunciation occurs if any of the following condition ists:
- 1. Instrumen i dicates measured levels above the alarm e oint.
i 2. Loss of Power.
- 3. Low flow.
- 4. Instrument indicates a downs a e fa re.
- 5. Instrument controls not set i pera e mode, l
I (3) The initial CHANNEL CALIBjM ON shall be perfor using one or more of ( the reference standajr Vfortified by the National B (p!g M - using standards tp w have been obtained from suppliers tT (u of Standards or partic'ipate l in measuremen nr,urance activities with-NBS. These standar shall permit ca rating the system over its intended range of energy l measu ent range. For subsequent CHANNEL CALIBRATION, sources that ve-b related to the initial calibration shall be used, j j(,4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: L 1. 1500 t 30 ppm hydrogen, balance nitrogen, fnr the outlet hydrogen monitor and
-2. 4 1 0.1 volume percent hydrogen, balonce nitrogen, for the inlet hydrogen monitor.
y.(jb'TheCHANNELCALIBRATIONshallincludetheuseofstandardgassamples containing a nominal:
- 1. 7511.5 ppm oxygen, balance nitrogen, for the outlet oxygen monitor and
- 2. 3.5 1 0.1 volume percent oxygen, balance nitrogen, for the inlet oxygen monitor.
SUMMER - UNIT 1 3/4 3-79 Amendment No. 7 56
3/4.11 RADI0 ACTIVE EFFLUENTS ' 3/4.11.1 LIQUIU EFFLUENTS
- ChNCENTRATION LIM G CONDITION FOR-OPERATION
\
/
3.11.1.1 e concentration of radioactive material released fr the site (see Figure .1-4) shall be limited to the concentrations s 'fc fied in 10 CFR Part 20, Anoe ix B, Table II, Column 2 for radi s t r th6n dissolved or entrai %d no le gases._ For dissolved.grat entr(nucl le ases, the concentration sh be limited to 2 x 10 microc ml otal activity. ; APPLICABILITY: At 11 times. l ACTION: With the concentration of radioactive materi r eased from the site exceeding the above limits, immediatel t re he concentration to within the above limits. SURVEILLANCE REQUIREMENTS
, i a
4.11.1.1.1 The radioactivity conten f each batch of radioactive liquid waste shall be determined prior to ele se by sampling and analysis in accordance with-Table 4.11-1. Th resul of pre-release analyses shall be used with the calculational met ds in the ODCM to assure that the concentration at the point of elease is ma tained within the limits of Specification 3.11.1.1. ' 4.11.1.1.2 - Post-release alyses of samples co osited from batch releases shall be performed in ac ordance with Table 4.11- .The resulta of The previous post-release alyses shall-be used with e calculational .ithuds in the ODCM to assure t t the concentrations at_the po t of re19ase were maintained within t limits of Specification 3.11.1.
-4.11.1.1.3 The dioactivity concentration of liquids d charged from continuous rele se points shall be determined by collectio and analysis of samples-in ac rdance with Table 4.11-1. -The results of th analyses shall.be used with t calculational methods in the ODCM to assure that the concentrat ns at the point of relene are maintained within the limits _of Specific lon_3.11.1.1.
4.11. 1.4 At least one circulating water pump shall be determine to be in. oper ion and providing dilution to-the-discharge structure at least_ nce per . _4- urs whenever dilution is required to meet the site radioactive ef uent -
;c centration limits of Specification 3.11.1.1.
SUP91ER - UNIT 1 3/4 11-1 ____ 2 __ ,_ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ o
-. . . . . - - . - - . . - . ~ - . - . - . - - . = - . ~ . - . --
- .. s TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum kwerLimit Liqui Release- Sampling- Analysis of Detection Type
. Type o A ivity (LLO)
Frequency Frequency sis .(pCi/ml)" A.-Batch W te P
-Release Each Batch Each Batch P
/N ,7 d Prinpipa a 5x10 Tanks 1, Waste
\ @ ters} Ga Monitor h) 131 /
/ 1x10
-6 Tanks P M On Batch /M n' / Dissolfed and
- 2. Condensate W Entrpined Gases 1x10 5 Demineral- (Gafena emitters)
M N, izer Backwash P WI Receiving Tank-Each Bat Composite b f' 3 1x10 5 3, Nuclear ~7 Gross Alpha 1x10
. Blowdown Monitor P Sr-89, St-90 .g Tank Each Batch Co otite Q[ b 5x10 Fe-55 ~1x10
-6 B. Continuous, 0 W ~7 Releases Grab Sample [ Composite Principa} Gamma Emitters 5x10 I-131 1x10
-6
- 1. Steam
. Generator- M M Blowdown GrabJ[ ample Okssolvedand 3 Enttained Gases 1x10
- 2.. Turbine / (GamgaEmitters)
Building 0 M H-3 \ Sump. rab Sample Composite c 1x10 5
- 3. Service Water-Effluent Gross Alpha 1x10 7
0 St-89, Sr-90 -8 Grab Sample Q Composite c \-5x10 Fe-55 -6
\1x10 SUMMER - UNIT 1 3/4 11-2
_ . _ . . _ . , _ , . _ _ _ . . _ _ , , _ . _ _ _ . _ . _ . . _ . . . . . . - . ~ , . - _ _ . _ . _ . . _ . .
_. . . .. . . - - ._ _ _ _ _ _ _ _ . . _ - - ~ _ _ _ . _ . TABLE 4.11-1 (Continued) / TABLE NOTATION a. The LLD is the smallest concentration of radioactive material in a sample that will yield a net count above background that will be detected dith 5% probability with 5% probability of falsely concluding that a lank servation represents a "real" signal. For particular measurement system (which may include radio separ tion): emical 4.66 s E*V 2. 22
- Y - xp (-Mt)
Where: LLD is the "a priori" lower limit of } pCi per uni mass or volume). Current tacti n as defitted above (as t ature defines the LLD as the detection capability for the i }rme ation only and the MCf minimum detact le concentr given instrument rocedu an type of h detection capability for e ample, ss is the standar deviation of e ekground counting rato cr of the c5unting rate of a ank sample as ppropriate (as counts per minute), E is the counting eff tency (as ounts per transformation), Visthesamplesize(inunitsgfmassorvolume), 2.22 is the number of tran f rmations per minute per picocurie, Y is the fractional radio e ical yield (when applicable), A is the radioactive de y con ant for the particular radionuclide, and at is the elapsed tim between m point of sample collection and time of counting (fo plant efflue ts, not environmental samples).
.The value of s u d in the calculat n of the LLD for a detection systemshallbI counting rate sed on the actual o erved variance of the background appropriate) ther the of than counting rate o the blank samples (as on an unverifie theoretically predicted variance. I gamma-ray calculating the LLD for a ra ionuclide determined by ectrometry the background shoul include the typical contribu ons of other Typical alues of E, V, Y, and it shall be useradionuclides normally resent in the samples.
in the calculation. It should e recognizeo that the LLD is defined as an priori (before the-fact) li it representing the capability of a measurement system and not as a post .iori (after the fact) limit for a particular meas ement."
/
"For a 're complete discussion of the LLD, and other detection 1 the f lowing: its, see (1) ASL Procedures Manual, HASL-300.(revised annually).
(2 Currie, L. A.,
" Limits for Qualitative Detection and Quantitativ Determination - Application to Radiochemistry" Anal. Chem. 40, 58 93 (1968).
- 3) Hartwell, J. K. , " Detection Limits for Radioisotopic Counting Techn ues,"
Atlantic Richfield Hanford Company Report ARH-2537 (June 22,1972). \ s SUM ER - UNIT 1 3/4 11-3 \
i 4- # TABLE 4.11-1 (Continued) , _ TABLE NOTATION b A composite sample is one in which the quantity of liquid mpled is proportional to the quantity of liquid waste discharged ap'd in which (Ke method of sampling employed results in a specimen w (ch is re esentative of the liquids released. 4 c. To be opresentative of the quantities and co ce r ions of radios ive materials in liquid effluents, sem e composits d in proportion to the rate of ow f eshall offluent be stream. i Prior to ialyses, cil samples taken for t posite shall be thoroughly ple to be representati ixed-inorderforthecomposite) of the effluent release. c ;
- c. A batch release is the discharge y volume. wastes of a discrete Prior t ysampling for analys{ey each batch shall be-isola and then thoroughly mixed, by a method escribed in the 00CM, to assure representati sampling,
- e. A continuous release i the dische ge of liquid wastes of a nondiscrete volume; e.g. from a olume of system that has an input flow during the continuou rele se.
- f. The principal gamn.a emitters or which the LLD specification applies exclusively are the follow g adionuclides: Mn 54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-1 4, Cs 137. Co-141,;and Ce-144. This IIst "
does not mean that only hese nu ide* are to be detected and reported. Other peaks hich are m asurable and identifiable, together with the above nucli s, shall-also e identified and reported. 1 1 X OUMER - UNIT 1 3/4 n.4
- -.- = - . .- - . . - . -_ - - - . - . . - . - _. - - . --- - _.
/
l RADIDACTIVE EFFLUENTS
',00SE LIMIT!NG CONDITION FOR OPERATION /
/
' /
311.1.2\ lhe dose or dose commitment to an individual from radio / active materials in liquid's,ffluents Figure 5.1-4) shall released, be limited:from each reactor unit, from the,/ite (see
- a. Our any calendar quarter to less than o total hody and to less than or equal t u)to1.5mremtothe o any organ, and
- b. During an calendar year to less th equa to 3 mrem to the total body nd to less than or equal 0 rem to any organ.
APPLICABILITY: At all t mes. ACTION: a. With the calculated ose from the elease of radioactive materials in liquid ef fluents e'veeding an of the above limits, in lieu of any other report regulftd by Specification 6.9.1, prepare and submit to the Commission withinN 0 days, pursuant to Specification 6.9.2, a Spstial Report which iden s the cause(s) for exceeding the linit(s) and defines the cp ective actions to be taken to the releases and the proposed /ac ons to be taken to assure that subsequent releases wil /einNompliancewithSpecification3.11.1.2. b
- b. The provisions of spe ifications 0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11 1.2 Dose Calcy ons. Cumulative dose contri ons from liquid effluents shall be determinett in accordance with the 00C!i at least nce per 31 days. N SUMMER - UNIT 1 3/4 11-5
010 ACTIVE EFFLUENTS j
\
LIQUID WASTE TREATMENT / N
/
/
\ /
LIMITING CONDITION FOR OPERATION / i 3.11.1.3 Th liquid radwaste treatment system shall be OPERABLE. The appro-priate portion of the system shall be use6 to reduce the radion tive materials in liquid waste prior to their discharge when the projected dpfes due to the liquid effluent om the site (see Figure 5.1 4) when averaged over 31 days, would exceed 0.06 em to the total body or 0.2 mre frgan.* j APPLICABILITY: At al times. ACTION:
- 4) l
- a. With the liquid dwaste treatment s 4 inoperable for more than 31 days or with ra ioactive liquid wast being discharged without treatment and in ex ss of the above 1) its, in lieu of any other report required by Sp ification 6.9 i,f prepare and submit to the Commission within 30 d pursuant to Specification 6.9.2 a Special Report which includes t followi information:
1
- 1. Identification of the n rable equipment or subsystems and the reason for inoperab)oity,
- 2. Action (s) taken to re . ore the inoperable equipment to OPERABLE status, an
- 3. Summary descripti n of action ( taken to prevent a recurrence,
- b. The provisions of Specifications 3.0.3 nd 3.0.4 are not applicablo.
SURVEILLANCE REQUIREMENTS \ \ 4.11.1.3.1 Doses d to liquid releases shall be projecte at least once per 31 days, in accordance with the ODCM. 4.11.1.3.2 The iquid radwaste treatment system shall be demo trated OPERABLE by operating e liquid radwaste treatment system equipment for t least 30 minutes at 1 st once.per 92 days-unless the liquid radwaste syst has been utilized t pro:ess radioactive liquid effluents during the previo 92 days. 1 .
- Per reactor l
SUM 4ER - UNIT 1 3/4 11-6 l
RADI0 ACTIVE EFFLUENTS 3 4.11.2 1 GASEOUS EFFLUENTS DOS RATE LIMIT {0NDIT10N FOR OPERATION 3.11.2.1 The ose rate in unrestricted areas due to radioactiv/ materials
/
released in gastous ef fluents from the site (see Figure 5.1-3) shall be limited to the following.
- a. For noble gases: Less than or equal to 50 Ye yr to the total body and I s than or equal to 3000 mr / r the skin, and b.
For all radiolpdines and for all radioAc}ftp materials in particulate formtoand equal 1500triti mr g w/yr to any ofTanith half lives grettW Less than hanor 8 days: APPLICABILITY: At all times. gTION: / With the dose rate (s) exceeding the\above limitt, immediately decrease the releaseratetowithintheabovelimit(7). SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate duej o noble gases gaseous effluents shall be determinedoftothebeODCM, procedures within the Above limits in acc dance with the methods and _ 4.11.2.1.2 The dose ra due to radioiodines, trit m and radioactive materials in particulate form wfi )h half lives greater than 8 d s in gaseous effluents shall be determined tb be within the above limits in a cordance with the methodsandproceddesofthe00CMbyobtainingreprese ative samples and performing analysyi in accordance with the sampling and a alysis program specified in TahTe 4.11-2. SUMMER - UNIT 1 i4 11-10
l .
- '- 3 l
T TABLE 4.11-2 y, C RADI0 ACTIVE GASEOUS WASTE PONITORING AND SAMPLING AND ANALYSIS PROGRAM I rn / 3 Minimum Lowes Limit
- Sampiing Analysis ef Detection i E Gaseous Release Frequency Frequency Type of Activity Analysis y '(LLG)(pC1/ml)*
l 4 A .' Waste Gas Sterage P P /
/
Tank ch Tank Each Tank Principal Gamma Eemitters4 4 1x10 l' Gr Sample
/
81 Reactor Building P . P / . -36" Purge Line Each Purge Each Purge Principal Gamma Emitters 9 1x10 I -6" Purge Line R-3 -6 1x10 l j 82 Reactor Building M D M /
-6" Purge Line Grab Sample ,/ Principal Gamma Emmitters9 1x10
~4 R
(if continuous) ' H-3 -6 w 1x10 r b C. Main Plant Vent MD *' Grab Sample 9 1x10-4 ipa 1 Gasuna Emitters
- H-3 1 p 1x10
-6 D.I. Reactor Building u Contin'9us / I-131 2 g Purge 5, ampler Charcoal Sample
/
I-133 b( 1x10[30 1x10 rs 9 PrincipalGammaEnh -11 c 2. Main Plant Vent / Continu9us Sampler Particulate Sample 1x10 i 1-131, others ! Continupus M Sampler Composite _g Gross Alpha a10 Particulate Sample ' Continupus Q Sampler Composite ,37 Sr-89, Sr-90 1x10 ! Particulate Sample Continuous -6 \ l Noble Gas Noble Gases 2x10 4 Monitor Monitor Gross Beta
l
\ TABLE 4.11-2 (Continued)
TABLE NOTATION a, the LLO is the smallest concentration of radioactive material in a sa th'at will yield a net count above background that will be detected w fh e 95%' probability with 5% probability of falsely concluding that a b1 )nk obser\ation represents a "real" signal. , For a p\ art l separatioh)icularmeasurementsystem(whichmayincluderadioch ical { i
* \ 4.66 s b E*V 2.22 Y exp (-M t) i j
Where: LLO is the "a p ori" lower limit of de t ' pCi per unit masssor volume). Current li as defined above (as the detection capa lity for the instrumen ion only ure defines the LLD as and the MDC minimum detactable c ncentration, as the tection capability for a ; given instrument proc ure and type of s mple. s s is the standard devi c5unting rate of a blank ion of the b kground counting rate or of the mple as a repriate (as counts per minute), E is the counting efficienc (as fc unts per transformation), V is the sample size (in units 4( mass or volume), 2.22 is the number of transfor p a ions per minute per picoeurie, Y is the fractional radiochetical eld (when applicable). A is the radioactive deca constant r the particular radionuclide, and at:is the elapsed time etween midpoint of sample collection and time of counting (for f lant effluents, n environmental samples). The value of s used in the calculation of systemshallbIbpedontheactualobservedhe LLD for a detec' tion ariance of the background i counting rate or 'of the counting rate of the b ank samples (as ' appropriate) r her than~on an unverified theor ically predicted variance. In alculating the LLD for a radionue de determined by ' gamma-ray s ctrometry the background should inclu the typical contributipfis.of other radionuclides norailly presen in the samples. Typical values of E, V, Y, and At.shall be used in th calculation. ' It should recognized that the LLO is defined as an a prior (before the fact)-11m) representing the capability of a measurement system and not'as a poster /ori (after the fact) limit for a particular measurement.
/
"For the afomo 'e complete discussion of the LLD, and other detection limits, s owing:
(1) ASL Procedures Manual, HASL-300 (revised annually). (2) Currie, L. A.,
" Limits for Qualitative Detection and Quantitative Determination
) Hartwell, J. K.',
- Application to Radiochemistry" Anal. Chem. 40, 586 93 (19 ).
" Detection Limits for Radioisntopic Counting Techniques,"
Atlantic Richfield Hanford Company Report ARH-2537 (June 22, 1972). SUMER - UNIT 1 3/4 11-12
\
N TABLE 4.11-2 (Continued)
\ TABLE NOTATION b.
. Analyses shall also be performed within 24 hours following shytdown,
'startup, or a THERMAL POWER change exceeding 15 percent of he RATED THERMAL POWER within a one hour period,
- c. Tritium grab samples shall be taken at least once per f4 hcurs when there(uelingcanal.S flooded. U
- d. Samples'shall be changed at least once per 7 d d analyses shall be completed within 48 hours after changing r er removal from sampler). ' Sampling shall also be performed ast once per 24 hours for at leasts 7 days following each shwte rtup or THERMAL POWER change exceediqg 15 percent of RATED THE WER in one hour and analyses shall be completed within 48 hours ~of changing. When samples collected for 24' hours are analyzed, the rresponding LLD's may be increased by a factor of 10.
\
e. Tritium grab samples's, hall be taken a least once per 7 days from the ventilation exhaus spent fuel is in the sp(ent fuel po 1.from the spgdt fuel pool area, whenever
- f. The ratio of the sample f rate to the sampled stream flow rate shall be known for the time M iod covered by each dose or dose rate calculation made in accordanc with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.
- g. The principal gamma emiM'ers for ich the LLD specification applies exclusively are the f 11owing radio uclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and e-138 for gase s emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65 Mo-99, Cs-134, Cs 137, Ce-141 and Ce-144 for particulateemiss)fns. This list does et mean that only these nuclides are to De detected and reported. Other peaks which are measureable an ' identifiable, together wit the above nuclides, shall also be identified and reported.
~ Sul+1ER - UNIT 1 3/4 11-13
DI0 ACTIVE EFFLUENTS
\
DOSE - NOBLE GASES j.IMITI CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous of ts, from each following: reactor u it, from the site (see Figure 5.1 3) shall be 11 4 ted to the
- a. During a calendar quarter:
radiation Less than or equal o 5 mrad for gamma d less than or equal to 10 mrad fo beta radiation and,
- b. During any calendar year: Less than or eque to 10 mrad for gamma radiation and ) ss than or equal to for beta radiation. !
APPLICABILITY: At all time ' ACTION a. With the calculated air oss fro radioactive noble gases in gaseous effluentsexceedinganyo(icthe report required by Specif fbove limits, in lieu of any other on 6.9.1, prepare and submit to the Commission within 30 days, Report which identifies t c rsuant to Specification 6.9.2, a Special se(s) for exceeding the limit (s) and defines the corrective tions o be taken to releases and the proposed corrective ac ions to b taken to assure that subsequent releases will be in mpliance wi Specification 3.11.2.2.
- b. The-provisions of pecifications 3.0. and 3.0.4 are not applicable.
( SURVEILLANCE REQUIREME)i S i
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4.11.2.2 Dose Cal etions Cumulative dose contribu calendar quarter nd current calendar year shall be dete nsined for the current in accordance with the ODCM least once per 31 days. 5 SUMMER - UNIT 1 3/4 11-14
*-twu-e--tw+e--r-+ , -dt-v----)6-e- sae-'>i6*1F4---e- ww-m w--tar m-se Je e m+w-wu=-sw-==-ph-m+- -m-eT1ip4+-u r+4de- r----,+w--~ w-+---ww-4m,ew<ie=,y.en-w,w w er > i4ien-s+-s -9a we.
-+uJu- Pbm 4w%W 4-wt 4 4mm-'g*W-re+7 vvy'w'ww myF"TW-* w'
RADI0 ACTIVE EFFLUENTS X / DOSE - RADIOI00lNES RADIOACTIVE MATERIALS IN DARTICULATE FORM, AND TRITg LIMITING CONDITION FOR OPERATION /
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3.11.2.3 The dose to an individual from radiofodines tritium a radioactive materials in' particulate form, and radionuclides (other than np61e gases) with half-lives greater than 8 days in gaseous effluents released rom each reactor unit (see Figure 5.1-3) shall be limited to the following:
- a. During a'ny calendar quarter: Less than or equa to 7.5 mrem to any organ and,'
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- b. During any ca'lendar year: Less th al to 15 mrem to any organ. Q)
APPLICABILITY: At all times' c) ACTION: ()
- a. With the calculated dose from t e release of tritium radiciodines, and radioactive materials i rticulate form with half lives greater than 8 days, in gaseous ef ents exceeding any of the above limits, in lieu of any other repo required by Specification 6.9.1, prepare andsubmtttotheCommipionwithin30 days,pursuanttoSpecifica-tion 6.9.2, a Special Report which, identifies the cause(s) for exceeding the limit odd defines th6 corrective actions to be taken s
to releases and the/ proposed actions'No t be taken to assure that subsequent releaspf will be in complid ce with Specification 3.11.2.3,
- 5. The provisions 'o/ f Specifications 3.0.3 a 3.0.4 are not applicable.
SLLRVEILLANCE REQUIREMENTS 4.11.2.3 Dose / Calculations Cumulative dose contributions for he current calendar quarter and current calendar year shall be determined accordance with the 0 H at least once per 31 days. SUMMER - UNIT 1 3/4 11-15
RADIOACTIVE EFFLUENTS CASEOUS RADWASTE TREATMENT LIM'! TING CONDITION FOR OPERATION 3.11.2 4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION li HAUST TREATMENTsSYSTEM shall be OPERABLE. The appropriate portions of tt)/ GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive mat tials in gaseous waste prior to their discharge when the projected gaseo effluent air doses due to' gaseous effluent releases from the site (see figu 5.1-3), when averaged over 3,1 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiatbn. The appropriate portions of the VENTIL TREATMENTSYSTEMshallbeusedtoreduceradioactivemater/IONEXHAU als in gaseous e to gaseous etfluent waste prior to the'1((discharge when the projected dosesreleasesfromthe exceed 0.3 mrem to an) organ.* A APPLICABILITY: At all es. g ACTION: a. With the GASEOUS RADVASTE TREATHENT/tYSTEM and/or the VENTILATION EXHAUST TREA1 MENT SYSTEM inoperab M for more than 31 days or with gaseouswastebeingditshargedw)thouttreatmentandinexcessof theabovelimits,inliegofa other report required by Specifica-tion 6.9.1, prepare and sohm to the Commission within 30 days, pursuant to Soecification 8 .2, a Special Report which includes the following information:
- 1. Identificationof)ieinoparableequipmentorsubsystemsand the reason for i perabilit)
- 2. Action (s) take restore t inoperable equipment to OPERABLE status,and/
- 3. Summary $'r.ription of action (s) t en to prevent a recurrence,
- b. The provisi s of Specifications 3.0.3 and .0.4 are not applicable.
SURVEILLANCE REQUIAEMENTS s 4.11.2.4.1 Dos es/due to gaseous releases from the reactor all be projttted at least once p 31 days, in accordance with the ODCH. 4.11.2.4. The GASEOUS RADWASTE TREATHENT SYSTEM and VENTILATI EXHAUST TREATHENJ SYSTEM shall be demonstrated OPERABLE by operating the SEOUS RADWAST,I TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATM T SYSTEM equippent for at least 30 minutes, at least once per 92 days unless e app opriate system has been utilized to process radioactive gaseous e (luents du7ngtheprevious92 days. [Perreactorunit Sut94ER - UNIT 1 3/4 11-16
xRADI0 ACTIVE EFFLUENTS i 3/4 11.3 SOLIO RA010 ACTIVE WASTE ' l
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LIMITING CONDITION FOR OPERATION
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3.11.3 The lolid radweste system shall be OPERABLE and used, as plicable in accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATI and packaging of radioactive isastes to ensure meeting the requirements of 1 FR Part 20 and of 10 CFR Part 7hprior to shipment of radioactive wastes fr the site. APPLICABILITY: At all times.
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ACTION:
- a. With the packa og requirements of 10 art 20 and/or 10 CFR Part !
71 not satisfied, suspend _ shipments of . ectively packaged solid I radioactive wastes rom the site, b. With the solid radwas y system inop able for more than 31 days, in 1 lieu of any other repor required y Specification 6.9.1, prepare and submit to the Commissio(n Nithin days pursuant to Specification 6.9.2 a Special Report which i ludes the following information: l
- 1. Identification of the } eperable equipment or subsystems and the reason for inoperabill y,
- 2. Action (s) taken to restore t inoperable equipment to OPERABLE status,
- 3. A description f the alternative sed for SOLIDIFCATION and packaging of radioactive wastes, a
- 4. Summary scription of action (s) take to prevent a recurrence.
- c. The provis ns of Specifications 3.0.3 vnd 3. 4 are not opplicable.
I 1-SURVEILLANCE REQA(IREMENTS s
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4.11.3.1 T e solid radwaste system shall be demonstrated OPERAB
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at least once per days by: a Operating the solid redwaste system at least once in the p -vious 92 days-in accordance.with the PROCESS CONTROL PROGRAM, or b. l Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONT OL , PROGRAM. l ! Su m ER - UNIT 1 3/4 11-19 L
RADI0ACV!VE EFFLUENTS
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SURVE)l.LANCEREQUIREMENTS(Coatinued) /
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4.11.3.2 0 CESS CONTROL PROGRAM shall be used to verify the OLIO! FICA-TION of at least ne representative test specimen from at leasyevery tenth batch of each type f wet radioactive waste (e.g., filter slydges, spent resins, evaporator b s, boric acid solution , and sodi ' sulfate solutions),
- a. If any test spe' en fails to veri 50e ! ATION, the SOLIDIFICA-TION of the bate nder test shall ded until such time as additional test spe mens can be o a 3e , alternative SOLIDIFICATION parameterscanbede(teqmined,thedance with the PROCESS CONTROL PROGRAM, and a subsequeh tesl'4 rJ es SOLIDIFICATION. SOLIGIFICATION of the batch may then be(gsumect(Asing the alternative SOLIDIFICATION parameters determined by thesPROCESS CONTROL PROGRAM.
- b. If the initial test speci A a batch of waste fails to verify SOLIDIFICATION, the PRD,0fSS CONTR0 PROGRAM shall provide for the collection and testiptf of represen(taQve test specimens from each consecutive batch pf the same type of waste until at least 3consecutiveiptialtestspecimensde trate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified s, required, as provided in Specific (ion 6.13, to assure SOLIDIFICATI of subsequent batches of waste.
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'N SUMMER - UNIT 1 3/4 11-20
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R 010 ACTIVE EFFLUENTS I \ 3/4.'11.4 TOTAL DOSE LIMITI ONDITION FOR OPERATION
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3.11.4 The ose or dose commitment to any member of the public ofradioactivityandradiation,fromuraniumfuelcyclesources/,shallbe ue to releases limitedtoless1anortaualto25mremtothetotalbodyorgnyorgan the thyroid, which shall be limited to less than or e ual to/75 mrem) ove(except r 12consecutivemont(. APPLICABILITY: At all' times. ACTION: - j a. With the calculate dosesfromtherelep[eofradioactivematerials in liquid or gaseou effluents exceedi tion 3 ll. l.2.a. 3.11 1.2.b, 3.11.2.jV.fg twice the limits a 3.11.2.2.b of Specifica-3.ll.2.3.a. or 3.ll 2.3.b, in lieu of prepare and submit to th ny other rejsort required by Specification 6.9.1, Specification 6.9.2aSpeialRe.p)ortwhichdefinesthecorrectiveCommiss 6n, action to be taken to redu s sequent releases to prevent recut-rence of exceeding the limit of Specification 3.11.4 This Special Report shall include an ana)$ (s which estimates the radiation exposure (dose) to a membe/ of The public from uranium fuel cycle sources (including all eJtluent 12 consecutive month petiod that 1 thways and direct radiation) for a this report. If the p'stimated dose cludes the release (s) covered by
) exceeds the limits of Specifica-tion 3.11.4, and if the release cond tion resulting in violation of 40CFR190hasnotalreadybeencorrecyd,theSpecialReportshall include a reques foravarianceinaccgdancewiththeprovisions of 40 CFR 190 a d including the specifieasinformation of 6 190.11(b).
Submittal of e report is considered a ti ly request, and a variance is granted uftil staff action on the request is complete. The variance op1y relates to the limitt, of 40 CFR 90, and does not apply in /ny way to the requirements for dose I'mitation of 10 CFR Part20/asaddressedinSpecifications3.11.1a, 3.11.2.
- b. The visions of Specifications 3.0.3 and 3.0.4 ar not applicable.
$URVEILLA E REQUIREMENTS
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4.11 Dose Calculations Cumulative dose contributions from liquid a gas, ous effluents shall be determined in accordance with Specifications .11.1.2,
/4 1.2.2, and 4.11.2.3, and in accordance with the 00CM.
SUMMER - UNIT 1 3/4 11-21 i i
I l N '3/4.12 RADIOLOGICAL ENVIRONMENTAL MON!TORING
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3/4,12.1 MONITORING PROGRAM s LIMIT!NG CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program s all be ennducted as specified'in Table 3.1R 1. APPLICABILITY: 'At all times. g
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ACTION:
- a. With the iological environmental monitoring ogram not being conducted as pecified in Table 3.12-1, in lie of any other report required by R5 tcification 6.9.1, prepare an[d f,ubmit to the Commission, in the Annual reasons for not(diological c nducting Operating the program Report. a description as tequired and the of the for plans preventing a recu ence,
- b. With the level of ra (oactivity in ap environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, ihy lieu of y other report required by Specification 6.9.1, prep e an submit to the Commission within 30 days from the end of t aff cted calendar quarter a Special Report.
When more than one of the r fonuclides in Table 3.12-2 are detected in the sampling medium, thi eport shall be submitted if: concentration (1) concent.ation (2) + ...> 1.0 ~ limit level (1) limit le el (2) ther than those f Table 3.12-2 are detected and When radionuclides'lant effluents, thi report aretheresultof/p shall be submitted if the potential a;t.ual dose to an individ is equal to er greater than the cale at year limits of Specift tions 3.11.1.2, 3.11.2.2 t.nd 3.11.2.3 This report is not requirst f the measured level of radioactivjywasnottheresuitofplante luents; however, in such an event, the condition shall be reporte and described in the Annual 84diological Environa. ental Operating Rep rt.
- c. With ilk or fresh leafy vegetable samples unavai ble from one or mor of the sample locations required by Table 3.12 , in lieu of a other report required by Specification 6.9.1, pr re and submit o the Commission within 30 days, pursuant to Specift tion 6.9.2, a SpecialRepertwhichidentifiesthecauseoftheunavaihbilityof samples and identifies locations for obtaining replaceme samples.
The locations from which samples were unavailable may the e deleted from those required by Table 3.12-1, provided the locations rom which the replacement samples were obtained are added to the envir mental monitoring program as replacement locations,
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab e.
SU MER - UNIT 1 3/4 12-1 Amendment No. 35
DicLOGICAL ENVIRONMENTAL HONITORING SURVE!LLANCEREhu!REHENTS , /
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4.12.1 The radiologica 1 onitoring samples shall be collected pursuant to Table 3J2(l entfronmen fro the I akibnt,;;iven in the table and figure in the ODCM and sJtalT'be anaiyn pur ant to t.ie-reg ~uirements of Tables 3.12-1 and 4.12-1 /
/ N s 4
SUMER - UNIT 1 3/4 12-2
i l j.
@ Table 3.12-1
{ Radiological environmental monitoring program Virgil C. Summer Nuclear Station
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f 5 Exposure Phtenay Minimum Number of Sampling and
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- Z and/or Sample Sample Locations and TypeandFrequen6y g Collection Criteria for Selection Frequency of Analy[ sis /
AIR 80RNE I. Particulates A 3 icator samples to Continuous sampler i be ta at locations operation with Gross beta following ' (in dif ent sectors) weekly collec 'on. filter change; Qwrterly beyond but close to composite (by location) i the exclusion for gamma isotopic. 1 undary as practicable e the highest offsite i t* sectoral ground level y concentrations are f
- i. , anticipated.(1) w -
8 1 Indicator saep e to 1 be taken in _ sector ' f> beyond as close to \ { the ex usion boundary as acticable cor e-ing to the t-i residence having the highest anticipated j' offsite ground level concentration or dose.(1) { C 1 Indicator sample to { be taken at the loca-i j tion of one of the j dairies most likely to
- be affected.(1)(2) 1
Table 3.12-1 (continued) I / E ExposureP)thway Minimum Number of Sampling and Type and Frequency / e and/or Sample Sample Locations and Collection of Analysis ' g Criteria for Selection Frequency j/- AIRBORNE, (continued D M ontrol sample to bela(enatalocation at lea
- 10 air miles from the ite and not inthemostkrevalent wind directiert 1)
II. Radiciodine A 3 Indicator samples o Continuous sampler/ Gamma Isotopic be t,.en at two loca- operatien with for I-131 weekly. M tions as given in I.A. weekly cannister above. 11ecti(n. n i 8 1 Indicator :,ampic to be taken at the loca- r tion as given in I.B. above. /t
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C 1 Indicator sampit to f Ed . be taken at the 1cca-tion as giv in I.C. Q above. D 1 Cnatrol sample to be p taken at the location E n I.D. above. a R a
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E Table 3.12-1 (continued)
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[ f ' Expos e Pathway Minimum Number of Sampling and Type and Frequency E and/or pie Sample Locations and Collection of Analysis ' p Criteria for Selection frequency ,
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w AIRBORNE, (conti ed) III. Direct 13 Indicator stations Monthly or y sraa dose monthly ith two or more quarterly. / or quarterly. d .imeters *.o form an ' inne ring of stations - irs the 13 accessible sectors whhin 1-2 miles of the plant. /
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R B 16 Indicator stations with two or rv>re dost Z eters to form an ouJer J, ring of stations,,in the 16 sectors witMn 3 to 7 5 miles of e plant. ' C 8 Stat" ns with two or
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nor dosimeters to be p) d in special ( interest areas such as l population centers, ! nearby residences, schools, and in 2 or 3 areas to serve as con-trol stations.
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E Table 3.12-1 (continued) x Minimum Number of Sampling and Type and Frequency #
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Expos Pathway E and/or le Sagle Locations and Collection of Analysis O Criteria for Selection Frequency /,/ ~ - WATERBORNE IV. Surface s 1 Irdicator sample Time composite Gamma isotopic monthly Water N ownstream to be taken samples with col- /with quarterly composite atNp location which lection every (by location) to be ana-allo k for mixing month (corre- lyzed for tritium.(5) and difb(ion in the spends to U S ultimate receiving continuop[ sam-river. pling te).(3) B 1 Control sample t be taken at a location cA Z the receiving river, E sufficiently far stream such tha o a effects of p- ed storage operation are anticipatId. / ; C 1 ator sample from t> cation immediately O upstream cf the nearest downstream municipal water supply. D 1 Indicator sample to be taken in the upper reservoir of the pumped \ storage facility.
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w Table 3.12-1 (continued) / h \ '
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9 Exposurhfathway Minimum Number of Sampling and e Sample Locations and Type and Frequency and/orSas@le Collection of Analysis ,e
\ Criteria for Selectica Frequency WATERBORNE, (contirtued)
Indicator sample to Grab sampilng As igv1V.A above. b'ex taken in the upper monthly.(3) / rese'htoir's nnn-fluctu- ' ating hcreational area. F
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1 Control sample to be taken at a location on a separated unaffetted - watershed reservoir. V. Ground Water A 2 Indicator samples to r5 uarterly grab Gamma isotopic and O be taken within the sampling _(5) tritium analyses exclusion boundary d \ quarterly _(5) in the direction potentia 11y af) cted , groundwateysupplies.
- 8 1 Contpoi sample from una cted locats,on. ( -
VI. Drinking A Indicator sample Monthly grab
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Monthly (3) gamma isotopic Water - from nearby pubitc sampling.(3) grocad water supply. and gro'4 Beta analyses and quarterly (5) compos-g ite for tritiw analyses.
<a 5 B 1 Indicator (finished Monthly ccm-3 water) sample from the posite sample 3 nearest downstream f water supply.
C 1 Control (finished Monthly com-water) sample from posite sample an unaffected water supply. i
Table 3.12-1 (continued)
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E Exposurkathway Minimus Number of Sampling and TypeandFrequency/ and/or Sample Sample Locations and Collection of Analysis ' E \ Criteria for Selection Frequency
/ $ INGESTION VII. Milk (2) A Samples from milking Semi-monthly when Gasusar/
isotopic and I-131 animals in 3 locations animals are on ap(lysis semi-monthly (6) with(n 5 km distant pasture, (6) Ainimals are on pasture; havinf4he highest dose monthly other monthly (3) at other times. potentia \ If there times.(3) are none th , I sample from milking inals in each of 3 areas tween w 5 to 8 km distant where 12 dosesarecalculatedh g be greatep than 1 area per year. 4 B 1 Control sample be I, taken at the loc 4 tion of /i a dairy > 20,dles dis-tant and not in the most l prevalet/ wind direc- { [s ( t tion. () C Indicator grass (for- Monthly when Gamma Isotopic. age) sample to be taken available.(3) at one of the locations beyond but as close to the exclusion boundary as practicable where the highest offsite sectoral ground level j concentrations are anticipated.(1)
Iable 3.12-1 (continued)
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E Exposurehthway Minimum Ne'eer of Sampling and Type and Frequency e and/orSampth Sample Lot.itions and Collection of Analysis Criteria ter Selection Frequency g K- [ INGESTION, (continued) D ndicator grass Monthly when Gamma sotopic. (fo e) sample to be available.(3) taken b the location of VII A ove when animals are pasture. E 1 Cont r ol grass (forage) sample to b taken at the location { of VII B above. U VIII. Food A 2 Samples of broad leaf Month y when Gamma isotopic d> Products vegetatior. grown in availabi analysis on edible 2 nearest offsite portion. locations of hi est 3 calculated apnual average grp6nd-level D/Q , if milk amplir.g is not s > perfo ed within 3 km or if ilk sampling is not rformed at a location within 5 to 10 km where the doses are calcu-g lated to b I mrea/yr. g greater than a 2 B 1 Control sample for ?+ the same foods taken 2 at a location at least P 10 mi'es distant and m not is. the most preva-
- 1ent wind direction if
u, C m ' Table 3.12-1 (continued) _ \
. Expo e Pathway /
e and/or Minimum Nuder of Sampling and le Sample Locations and Type and Frequency
} Collection of Analysis Criteria for Selection Frequency INGESTION,(cob )
! 8 (Cont'd) ! milk sampling is not
,rformed within 3 km or 1,
11k sampilng is not at a ation within 5 to 8 km ere the doses j are calcu ged to be greater than aren/yr. k w IX. Fish A 1 Indicator samp to ' 3 Semi an 1(7) Gamma isotopic. on edible
~ be taken at a locat collection of the l 7 in the upper reservol . fo)i'owingspecie portions semi-annually.
O s if avail-able: bass, brea, crappie catfish, carp;D ; rage fish (shad). 4 8 1 Indicator e to be taken a location l in the r reservoir [" C 11 cator sample to ' taken at a location t in the upper reservoir's nonfluctuating recrea-tional area. D 1 Control sample to be taken at a location on the receiving river, sufficient?y far up-stream such that no i efftets of pumped storage operation are anticipated. 6
' + ^ * ~ - r-,-- -- - > _ _ _ _ _ _ _ . - _ _ _ _ . _ . - _ _ , _ _ , _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
E Table 3.12-1 (continued) /
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Expo re Pathway Minimum Number of E and/or . Te Sampling and TypeandFrequ[ncy Sample Locations and Collection Z Criteria for Selection of Analysis Frequency ,/ AQUATIC
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X. Sediment 1 Indicator sample to Semi annual grab / Gamma isotopic. e taken at a location sample.(7) / in he upper reservoir. B 1 Indic tor sample to be taken i the upper reservoir's n -fluc-tuating recreats al Ru area. .; Z' C 1 Indicator samp to
~ be taken on th shore-Iine of the ower reservoir -
l
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l D 1 rol sample to be / : en in receiving i l river, sufficiently ~ far upstream such that i no effects of pumped storage operation are l anticipated.
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v, c NOTES N m o (1) c p ample site ented in locations are based on the meteorological analysis for the period of record as Chapters 5 and 6 of the OLER. 5
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(2) Milking pinal and garden survey resnits will be analyzed annually. Should.the survey indicate nhdairying activity, the owners shall be contacted with reg io a contract for supplying sufMcieat samples. If contractual arrangements can be sa added for addit} al milk sampling up to a total of 3 Indicator 1 site (s) will be ions. (3) Not to exeed 35 days. (4) Time composite samples are s les which are collect with equipment capable of collecting an aliquot at time intervals w period. h are short (e.g. doorly) relative to the compositing m (5) At least once per 100 days. 2 }
~ ~ (6) At least once per 18 days. [;-
s (7) At least once per 200 days. ( NOTE: Deviations from thi ampling schedule may occasionally' essary if sample media are unobtainable due hazardous conditions, seasonal unava size, malfunctir ns of automatic sampling or analysis equipm%1ity, insufficient sample reasons. [ specimens are unobtainable due to sampling equipmenengndotherlegitimate effort malfunction, every all be made to complete corrective action prior to the e f the next sampling peri
. Deviations from sampling analysis schedule will be described the annual r ort.
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. n TABLE 3.12-2 REPORTING LEVELS FOR RADIDACTIVTTY CONCENTRATIONS IN ENVIR00 MENTAL SAMPLES Reporting Levels E
a Wab Airborne Particulate Fish Analysis Mil Food Products l (pCi/ or Gases (pCi/a f3 (pCi/Kg, vet) f/1) (pCf/Kg, wet) H-3 . 2 x 10*I*) N.A. M.A. M.A. N.A. 3 N Mn-54 1 x 10 \ A. 3 10 4 M.A. N. A. Fe-59 4 x 10 N. I x 10 N.A. M.A. Co-58 1 x 1( N. A. 3 e 10* N.A. N.A. 2 # Zn-65 3 x 10 N.A. u: 10 M.A. N A. Zr-Mb-95 2
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4 x 10 .A. 2 x 10* N.A. M.A. 1 I-131 2 0.9 N. 3 1 x 10 1 Cs-134 30 10 1 x 10 3 60 1 x 10 CS-137 20 3 2 x 10 70 2 x 10 Ba-La-140 2 x 10 N.A. N.A. 3x 0 2 ,,g, (a) For dri ng water samples. This is 40 CFR Part 141 value.
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TABLE 4.12-1 E
- MAXIMUM kALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a,c E
W Airborne Particulate or Gag Fish Milk
/ Sediment N ater FoodProducts/
O Analysis ( i/1) (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pCi/kg, dry) w gross beta 4 1 X IC' N.A. N.A. .A. N.A. H-3 2000 N.A. N.A. N.A. N.A. N.A. Mn-54 15 N.A. 130 N . N.A. N.A. Fe-59 30 N. . 260 N.A. N.A. N.A. Co-58, 60 15 N. A. 130 N.A. M.A. N.A. 2n-65 30 N.A. 260 N.A. N.A. N.A. ~ \ 7 Z r-95 30 N.A. Nh N.A. N.A. N.A. 5 Nb-95 15 N. A. N.A. N.A. N.A. N.A. D -2 60 N.A. 1-131 1 10 N.A. 1
- Cs-134 15 5 X 10
-2 130 15 C 150
~2 180 Cs-137 18 6 X 10 150 18 %0 Ba-140 6 N.A. N.A. 60 N.A. N.A.
La-140 15 N.A. N.A 15 N.A. N.A. s
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1 TABLE 4.13-1 (Continued) l TABLE NOTATION
- a. T en leonmental 4.12-1 lists detection capabilities for radioactive materia} in samples.
term These detection capabilities are tabulated in purpo of the lower limits of detection (LLDs). The LLD is defined, for materi softhisguide,asthesmallestconcentrationofradipfctive that wil inbea detected sample that will yield a net count (above syst,e'm background) with 95% probability with only 5% pr f alsely co ciuding that a blank observation represents a real" 'pbability of signal. For a particu ar measurement system (which may include radiochemical separation): 4.66 sb \ ,, W LLO = E - V 's 2.22
- Y exp(-Ady Where:
LLD is the "a priori" low per unit mass or volume). limit of detection as defined above (as pCi Current liter'ature defines the LLD as the detectioncapabilftyfortheinstrumen%tiononlj,andtheHDC, detectable concentration, as minimum instrument, procedure, and type detegtioncapabilitjforagiven emple.) s h is the standard deviation of background counting rate or of the cOuntingratecfablanksampleas(appropriate (ascountsperminute), E is tne counting efficiency as count per disintegration), V is the sample size (in nits of mass or olume), 2.22 is the number o sintegrations per mi te per picoeurie, Y is the fractional radiochemical yield (when a licable), A is the radioacpive decay constant for the partic lar radionuclide, and at is the ela sed time between sample collection (or d of the sample collection iod) and time of counting. The H ue y fs used s in the calculation of the LLD for a rticular measi yent syItem should be based on the actual observed v riance of the backg ound counting rate or of the counting rate of the blan samples (as appr riate) rather than on an unverified theoretically predic ed variance.
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SUMMER - UNIT 1 3/4 12-15
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JABLE4.12-1(Coritinued) TABLE NOTATION In cale ating tne LLD for a radionue ide determ d by gamma-ray
.spectrone. , the background should in lu other radion ' typical contributions of ides normally prese t i ampl3s (e.g., potassium-40 in milk samples Typical values M , Y and At shall be used in the-calculations. -
It should be recognized t th LD is defined as an a priori (before the fact) limit representin (fie capability of a measurement system and not as a posteriori (after the ct) limit for a particuler measurement.
- b. LLD for drinking wa samples,
- c. Other peaks p ntially due to reactor ope tions-(fission and activation products)(which radionuc) des inare measurable Table 4.12-1, and shallidentifiab be identif together with the and reported.
SUM 4ER - UNIT 1 3/4 12-16
RADIOLOGICAL ENVIRONMENTAL MON 2TORING (4.12.2 LAND USE CENSUS LI NG CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify th location of the nea. st milk animal, the nearest residence and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables i each of the 16 meteorolog al sectors within a distance of five miles.
APPLICABILITY: As all times. ACTION:
- a. With a land u census identifying a lo at on(s) which yields a calculated dose or dose commitment gr bein0 calculated in Specificat. 4 %. .3, than the values currently in lieu of any other report required b Specification 9) , prepare and submit to the Commission with!n 3 days, pursuant Specification 6.9.2, a Special Report which identif s the new lo tion (s).
- b. With a land use census 1 entify a location (s) which yields a calculated dose or dose mmit nt (via the same exposure pathway) 20 percent greater than at being obtained in accordanc )with Specification 3.12.1, in lieu ofcation from w any other report required ecification 6.9.1, prepare and submit to the Commission within 0 da , pursuant to Specification 6.9.2, a Special Report which id tifies he new location. The new location shall be added to the adiologica environmental monitoring program within 30 days. The ampling loca on, excluding the control station location, having th lowest calculat d dose or dose commitment (via the same exposure,fathway) may be del ed from this monitoring program af ter OcYober 31 of the year in which this land use census was conducted,
- c. The provisi s of Specifications 3.0.3 and 0.4 are not applicable.
SURVEILLANCE REQUI. MENTS
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4.12.2 The 1 nd use census shall be conducted at least once p r 12 months between the ates of June 1 and October 1 using that informatio which will provide th best results, such as by a door-to-door survey, aeri survey, or by consu ing local agriculture authorities.
*Brfad leaf vegetation sampling may be performed at the site boundary i the firectionsectorwiththehighest0/Qinlieuofthegardencensus. /
SUMMER - UNIT 1 3/4 12-17
RADIOLOGICAL ENVIRONME4TAL MONITORING 3/4.12.3 INTERLABORAT0k'Y COMPARISON PROGRAM
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_LIMITI'kG CONDITION FOR OPERATION
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3.12.3~Ana\ lyses shall be performed on radioactive mater s s supplied as part of an Interlaborat ry Comparison Program whic has bee pproved by the Commission. APPLICABILITY: At all t'imes. ACTION: N -
- a. With analyses not being er med as required above, report the corrective-actions taken in the Annual Radiologi 1 prevent a recurrence to the Commission fronmental Operating Report,
- b. The provisions of ', ecifications .0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIRE s 4.12.3 A su ary of the results obtained as part of the a Interlabor ve required ipants ory Comparison Program and in-accordance with th 00CM (or partic-desi the EPA crosscheck program shall provide the EPA pro am code tion forOperating En ...onmental the unit)Report. shall be included in the Annual Radiolo ical l l
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-SUMMER - UNIT 1 3/4 12-18
INSTRUMENTATION l BASES 3/4.3.3.8 ADJQAC71VE LIQU10 EFFLUENT MONITORING INSTRUMENTATIOJ/
- The radioacti 6ef fluent instrumentatip-i nr.lf P' and control, as applicable, the%)erses of radttfactive materials in liquid ded to monitor 1
effluents during actual or potential rhisq of liquid effluents. The alarm / trip setpoints for these instruments shall be h iculated in accordance with the procedures in the ODCW to ensure that the alarm /frtp-will occur prior to exceeding the lp its i 10 CFR Part 20. The OPERABILITY and'Dse4 f this instrumentati6n is consistent with the requirements of General Desig'trCiteria 60 J 3 'and 64 of Appendix A to 10 CFR Part 50. N E X P LC,5IV E. GAS Mm eroR NGs 3/4. 3. 3. 9 -RAMOAGT4VE- CASE 00fr4FFEUEEMlHITORING- INST RUMENTATION dhe-redicact4ve geseous-ef+19ent-insteumentatfon-is-peovided-to-monitor-- and-control, Os-appl 4eabley-4he-releases-of-cadioact4ve-mater 4als-in-gaseous-. c f f kent +-during-aetual-oe-potential-releases-of--gaseous-af41uants. The
-a la rmA eip-se tpo i n t+-fo r-t he se- inst rumen t 5-sha 44-be-calcu l a ted-4 w esor4ance--
with-the- p roc edu re s - i n - t he - 00CM -to-e ns u ee-tha t-t he-el a rm/ t ei p-w i44-occ uc-p Moe-40-exceeding-the-14*14s-of--10 CF9 Part 20. This instrumentation-e4-se- inc1udes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the waste ges holdup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. _ 3/4.3.3.10 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ens."es that sufficient capability is available to detect loose metallic parts in :ne primary system and avoid or mitigat; damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Prirary System of Light-Water-Cooled Reactors," May 1981. SUMMER - UNIT 1 B - 4 3-4 l
l 1 3/4.11 RAD 10 ACTIVE EFFLUENTS EJSES _ 3/4.11,1 LIQUID EFFLUENTS 3/4.11.1K1 CONCENTRATION l
- This ecification is provided to ensure that the concentrati n of radio-activematerklsreleasedin_liquidwasteeffluentsfromthesit will be less than the concentration levels specified in 10 CFR Part 20 Appepdix B, Table l
II, Column 2. Thislimitationprovidesadditionalassurancet,hatthelevels of radioactive mat.erials in bodies of water outside the siteA111 result in exposures within ( l the Section II. A design objectives of Appendix I,10 CFR 50, to an individua and (2) the limitt, of 10 CFR 20,106 e The concentration lim t for dissolved or entrained nobly,(j) gases is based upon to the l theAssumptionthatXe(35isthecontrollingradioiso,topeanditsMPCinair *
- l. (submersion) was converthd to an equivalent concentr tion in water using the l- methods described in Inte) ational Commission o iological Protection l (ICRP) Publication 2.
p 3/4.11.1.2 OOSE V t This specification is provid to ment the requirements of Sections II.A III.A and IV,A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides t orth in Section II,A of Appendix I. The ACTION statements provide the req red operating flexibility and at the same time implement the guides set fM t in Section IV.A of Appendix I to assure that the releases of radioacgive meterial in liquid effluents will be L kept "as low as is reasonably ach)6vable."\ Also, for fresh water sites with l drinking water supplies which c4n be potenttally affected by plant operati_ons, , there is reasonable assurance that the operat\on of the facility will not result in radionuclide concep(rations in the fl ished drinking water that are inexcessoftherequiremenfsof40CFR141. Th dose calculations in the ODCM implement the requirphents in Section III. A Appendix I that confor:rance with the guides of Appendix I be'shown by calculat nal procedures based on models'and data, such dat the actual exposure of an individual through appro-priate pathways is u kely to be substantially undere timated. .The equations i l specifiedinthe00) for calculating the doses due to- ie actual release rates of_ radioactive materials in_ liquid effluents are c sistent with the methodologyprovJdedinRegulatoryGuide1.109,"CalculatinofAnnualDoses to Man from Ro (ine Releases of Reactor Effluents for-the.P pose of Evaluating Compliance wi 10 CFR Part 50, Appendix I," Revision 1, Octo er 1977 and ,- Regulatory ide l'.ll3, " Estimating Aquatic Dispersion of Eff1hnts from L Accidenta nd Routine Reactor Releases for the Purpose of Imple enting Appendi , " April 1977, i- is specification applies to the' release of liquid effluents f m each l- rea or at the site. For units with shared radwaste treatment system the li id effluents from the shared system are proportioned among the unit ring that system. l l SUMMER - UNIT 1 8 3/4 11-1 {- L l'
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 WoT fis.w. - Nm 3/471L 1.3 LIQUID WASTE TREATMENT [ N The OPERA'BIL QY of the liquid radwaste treatment system ensures that this system will be avaiPabl prior to release to the(enMr(nment.for use whenever liquid efflueJtetequire treatment The requirem portions of this system be used whe(specified-Sro'ent'that the appropriate vides assurance that the releases of radioactive . materials in Ticti( effluents will be kept "as low as is reasonably achievable." ThisJ pecTiicatTDn implements x the requirements of 10 CFR Part 50.36a, General,DerTgn Criterion 60 o bAppendix A to 10 CFR Part 50 and the design objective-gTven in Section II.D of Appen' dix I to 10 CFR Part 50. The specified lipit1r governing the use of appropriate portions 4the liquid radwaste treatment system were specified as a suitable fraction ortAdose desi natfjectives set forth in Section II. A of Apper. dix I,10 CFR Part $h Jqti d effluents. 3/4.11.1.4 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides asturance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area. 3/4 11.1.5 SETTLING PONDS The inventory limits of the settling ponds (SP) are based on limiting the consequences of an uncontrolled release of the pond inventory. The expression in Specification 3.11.1.5 assumes the pond inventory is uniformly mixed, that the pond is located in an uncontrolled area as defined in 10 CFR 20, and that the concentration limit in Note 1 to Appendix B of 10 CFR 20 applies. The batch limits of slurry to the chemical treatment ponds assure that radioactive material in the slurry transferred to the SP cre "as low as is reasonably achievable" in accordance with 10 CFR 50.36a. The expression in Specification 4.11.1.5 assures no batch of slurry will be transferred to the SP unless the sum of the ratios of the activity of the radionuclides to their respective concentration limitation is less than the ratio of the 10 CFR 50, Appendix I, Section II.A, total body level to the 10 CFR 20, 105(a), whole body dose limitation, or that: c I , 3 mrem /yr jC 500 mrem /yr = 0.006 3 where c 3 = unrestricted area SP, in microcuries/milliliterradioactive slurry concentration for ra SUMMER - UNIT 1 B 3/4 11-2
RADIOACTIVE EFFLUENTS BASES 3/4 11.1.5 SETTLING PONDS (Continued) Cd = 10 CFR 20, Appendix E, Table II, Column 2, concentration for single radionuclide "j", in microcuries/ milliliter. For the design of filter /demineralizers using powder resin, the slurry wato volume and the weight of resin used per batch is fixed by the cell surface .rea and the slurry volume to resin weight ratio is constant at 100 milliliters / gram of wet, drained resin with a moisture content of approximately 55 to 60% (bulk density of about 58 pounds per cubic feet). The wet drained slurry density is approximately ' gr/mi and the absorption characteristic for gamma radiation is essentially that of water. Therefore, C) 103 ml/gm) < 0.006, and pCi/gm j Cbj < '6 I pCi/ml Where the terms are defined in Specification 4.11.3.5. The batt.h limits provide assurance that activity input to the SP will be minimized, and a means of identifying radioactive material in the inventory limitation of Specification 3.11.1.5. 4til.2 GASEOUS EFFLUENTS ,
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3/4.11 x DOSE RATE ,< This spe eQionisprovidedtoensurethatthedoseat me at the siteboundaryfromgasouseffluentsfromallunitsonthes)te'gny will be within the annual dose limits o 10 CFR Part 20 for unrestricted-areas. The annual dose limits are the doses a sociated with the concent,rafions of 10 CFR Part 20, Appendix B, Table II, Column 1 N hese limits p 'rovide reasonable assurance thatradioactivematerialdischargedQngaseopseffluentswillnotresultin the exposure of an individual in an unre cted area, either within or outside the site boundary, to annual average, con'ce tions exceeding the limits specifiedinAppendixB,TableII,of10CFRPaK20(10CFRPart20.106(b)). For individuals who may at timps'be within the site 4'oundary, the occupancy 3f the individual will be suff.it'lently low to compensate for any increase in the atmospheric diffusion f. actor above that for the site bounda . The specified release rato limits re strict, at all times, the correspondin gamma and beta-dose rates abov -t'ackground to an individual at or beyond the si'te boundary s to less than c ual to 500 mrem / year to the total body or to less tharrqr equal t 3600 mrem / year to the skin. These release rate limits also relt t, at aM imes, the corresponding thyroid dose rate above background to a chi ti
,yh(the inhalation pathway to less than or equal to 1500 mrem / year. \
SUMMER - UNIT 1 B 3/4 11-3
RADI0 ACTIVE EFFLUENTS SASES
\Thisspecificationappliestothereleaseofgaseouseffluents om all reach rs at the site. For units with shared radwaste treatment sy/tems, the ga.eou effluents sharing hat system. from the shared system are proportioned among e units 3/4.11.2.2 SE - NOBLE GASES This spec fication s is provided to implement the req rements of Sections II.B II .A and IV.A of Appendix I, 10 CFR Parg 50. The Limiting ConditionI.for Oper tion implements the guides set forp in Section II.B of Appendix and at the sameTheACQ0Nstatementsprovidetherequiedoperatingflexibility time 1 lement the guides se fr Appendix I to assure t in Section IV.A of l the releases of rad ve material in gaseous effluents will be kept " low as is reas nab. Achievable". i The Surveillance Requirements implement the requirements in S h on III.A of Appendix ! that
'conformance with the guides f Appendix I b own by calculational procedures based on models and data such that the ac ~ exposure of an individual through to appropriate pathways is unlike calculations established in the u ( DCMptantially fo alculating the doses due to the underestimated. The dose actual release rates of radioacti nobi gases in gaseous effluents are -)
consistent with the methodology pro ided, in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine ' leases of Reactor Effluents for the Purpose of Evaluating Compliance wi CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide .111. " Methods for Estimating Atmospheric Transport and Dispersion of Gase Water Cooled Reactors," Revisio 1, July Efflu ts in Routine Releases from Light-1 7. The 00CM equations provided for determining the air doses the site bo dary are based upon the historical average atmospheric conditi . 3/4.11.2.3 DOSE-RADIOldDINES,RADI0ACTIVEMATERALSINPARTICULATEFORM AND TRITIUM.
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SectionsII.C,III./)andIV.AofAppendixI,10CFRPartThis specificat on is p
- 0. The Limiting Conditions for Optration are the guides set forth in Secti II.C of Appendix
- 1. The ACTION st'atements provide the required operating f1 ibility and at thesametimej$plementtheguidessetforthinSectionIV.A f Appendix I to assure that t)e releases of radioactive materials in gaseous e luents will be kept "as log /as is reasonably achievable." The 00CM calculation 1 methods specified h the Surveillance Requirements implement the requireme ts in Sec-tion III.
of Appendix I that conformance with the guides of Append y I be shown by calcy ational procedures based on models and data, such that the a tual exposu le of an individual through appropriate pathways is unlikely to tia11 underestimated. 'The ODCM calculational methods for calculating t esubstan- doses duej o the actual release rates of the subject materials are consistent w th the methodology provided in Regulatory Guide 1.109, " Calculation of Annual toses
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! SUP91ER - UNIT 1 \
B 3/4 11-4
RADIOACTIVE EFFLUENTS BASES to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluatin / liance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Reg tory Guide 1.111, " Methods for Estimating Atmospheric Transport and Disper on of Gaseous Ef fluents in Routine Releases from Light-Water-Copled Reactors, Revision 1, July 1977. These equations also provide for peter-mining the +ual doses based upon the historical average atmospheric conditions. Dr release rate specifications for radiciodines,' radioactive materials in part ulate form and tritium are dependent on the existing radionuclide pathwa s to man, in the unrestricted area. Th( pathways which wereexaminedinthedlvelopmentofthesecalculationswefe: 1) individual inhalation of airborne rdoynuclides, 2) deposition pf' radionuclides onto green leafy vegetation with subsequent consumptio by man, 3) deposition onto grassy areas where milk animalhand meat produci g animals graze with consump-tion of the milk and meat by man d nd 4) dep,psItion on the ground with subsequent exposure of man. 3/4.11.2.4 GASEOUS RADWASTE TREATHENT The OPERABILITY of the JA 005 RADWASTE TRE T NT SYSTEM and the VENTILATION EXHAUST TREATyENT SYSTEM ensures that systems will be available for use whenever gaseoys' effluents require treatment pMor to release to the environment. of these systems be used, when sp The4Iied, reg provides 61rement that the reasonable appropriate assurance tha portidn(t tQe releases of radioactive ma ials in gaseous effluents will be kept "as lowqs is reason-ably achiev3bi ". This specification implements the requirementsN f 10 CFR Part50.3fd,GeneralDesignCriterion60ofAppendixAto10CFRPaN50,and the deg gn objectives given in Section II.D of Appendix I to 10 CFR Par ( 50. The,specified limits governing the use of appropriate portions of the systqms wpfe specified as a suitable fraction of the dose design objectives set fort fin Sections II.B and II.C of Appendix 1, 10 CFR Part 50, for gaseous effluents _ 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammabil1ty limits. These automatic control features include isolation of the source of hydrogen and/or oxygen to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. SUMMER - UNIT 1 B 3/4 11-5 , l
O RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.6 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage .. tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. lhis is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure".
\/4.11.3 SO,LIO RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the syste 11 be available for use whenever solid radwastes require processing and pac aging prior to being shipped offsite. This specification implements the req 61rements of 10 CFR Plirt s 50.36a and General Design Criterion 60 of Appendix / to 10 CFR Part 50. The ptocess parameters included in establishing the P ESS CONTROL waste / liquid /
PROGRAM solidification may agen incibdg, but ratios, catalyst are not limited to waste waste oil type, content, waste wastep pr ,incipal chemical constituents, mixing and curing times. 3/4.11.4 TOTAL DOSE This specification is pro 190. The specification requires edtomeetthe/oselimitationsof40CFR Reportwheneverthecalculatedtbs(eNfromepreparapionandsubmittalofaSpecial Tant radioactive effluents exceed twice the design objective doses of Ap ix I. For sites containing up to 4 reactors, it'is highly unlikely that th resultant dose to a member of the public will exceed the dose limits pf'40 CF 190 if the individual reactors remain within the reporting requjrement leve . The Special Report will describe a course of action wh)th should result n the limitation of dose to a member of the public for 12 , consecutive months to ithin the 40 CFR 190 limits. ForthepurposesoftheSpecialReport,itmay-beahs med that the dose commit-cycle sources is menttothememberoftpfpublicfromotheruraniumf(ue(freqothernuc negligible, with the,ekception that dose contributions cycle facilities aVthe same site or within a radius of 5 m Res must be con-sidered. If thqMose to any member of the public is estimatedsto exceed the requirements pf 40 CFR 190, the Special Report with a request foha variance (provided th6 release conditions resulting in violation of 40 CFR 190 have not already heen corrected), in accordance with the provisions of 40 CFR 0.11, is con fdered to be a timely request and fulfills the requirements of FR 190 til NRC staff action is completed. An individual is not considere er of the public during any period in which he/she is engaged in carrin ut any operation which is part of the nuclear fuel cycle. . SUMMER - UNIT 1 B 3/4 11-6
1 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING p/
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BASES / 3/4.12. x MONITORING PROGRAM The r iological monitoring program required by this specif1 ation provides measurements of radiation and of radioactive materials in those xposure pathways and r those radionuclides, which lead to the highes,t potential radiation expos res of individuals resulting from thA s ti w operation. This monitoring progr if fuent monitoring therebysupplementstheradiologich)(o program by verifyi .g that the measurable concentrati jradioactive materials and levels of radiat'on are not higher than expec e nj the basis of the effluent measurements nd modeling of the environm J exposure pathways. m e 'fective for at least the The initially first three specifie years honitoring of com program will ercial operation. Foll hgthisperiod, program changes may be initiated b sed on operati7na' ex rience. The detection capabilit .s required by Taple 4.12-1 are state-of-the-art forroutineenvironmentalmeasrementsiningdstriallaboratories. It should be recognized that the LLD is d ined as an,/a priori (before the fact) limit representing the capability of a asuremp t system and not as a posteriori (after the fact) limit for a parti lar, measurement. Analyses shall be per-f)rmed in such a manner that the sta p LL0s will be achieved under routine conditions. Occasionally background uctuations, unavoidably small sample sizes, the presence of interferringf uc ides, s or other uncontrollable circum-stances may render these LLDs una tiievabTE. In such cases, the contributing factors will be identified and dpscribed i the Annual Radiological Environ-mental Operating Report. 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that gangesintheuseof unrestricted areas ar,e' identified and that modificatigns to the monitoring program are madefif tequired by the results of this ceh us. The best survey information from the door-to-door, aerial or consulting th local agricultural authorities shal / e b used. Thiscensussatisfie,thereq(rementsofSection IV.B.3 of Appe ix 1 to 10 CFR Part 50. Restricting the census to gardens of greater than 5 0 square feet provides assurance that signifihant exposure p pathways vip leafy vegetables will be identified and monitore since a garden of this site is the minimum required to produce the quantity (2 kg/ year) of leafy vegftables assumed in Regulatory Guide 1.109 for consumptio by a child. To deter'mine r this minimum garden size, the following assumptions w used, 1) that JO% of the garden was used for growing broad leaf vegetation (i ., sim Mar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/sq are meter. \
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SUMMER - UNIT 1 B 3/4 12-1
1 M ADIOLOGICAL ENVIRONMENTAL MONITORING BASES _ s
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3/4.12.3 INTERLABORATOR _MPAR $$N RAM TherequirementforpartiAc isprovidedtoensurethapfraependenon in an Interlaboratory Comparison Program the measurements of radroactive material ecks on the precision nxenvironmental andmatrices sample accuracyareof performed as pagt4f the quality assurance pri m for environmental monitoring in order t Gmonstrate that the results are reaso ly valid. SUMMER - UNIT 1 S 14 12-2
ADMINISTRATIVE CONTROLS
- c. lecondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
(i) Identification of a sampling schedule for the critical variables and control points for these variables. (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, including monitoring the discharge of the condensate pumps for evidence of condenser in-leakage. (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all off-control point chemistry conditions, (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
- d. Postaccident Sampling 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 conditions. The program shall include the following:
(i) Training personnel, (ii) Procedures for sampliag and analysis, (iii) Provisions for maintenance of sampling and analysis equipment. N S6ST b -> SUMMER - UNIT 1 :-12 Amendment No. 13.37,79 l _ _ _ _ _ _ _ _ ~
1 i l INSERT 3
- e. Radioactive Effluent Controls Program A program shall be provided conforming with 10CFR50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall 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 determinations in accordance with the methodology in the ODCM,
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10CFR20, Appendix B, Table II, Column 2;
- 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10CFR20.106 and with the methodology and parameters in the ODCM;
- 4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas conforming to Appendix I to 10CFR 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 or 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 10CFR50; i
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- 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10CFR Part 20 Appendix B. Table II, Column 1;
- 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the site boundary conforming to Appendix 1 to 10CFR50;
- 9) Limitations on the annual and quarterly doses to a member of the public from lodine-131 Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas beyond the site boundary conforming to Appendix 1 to 10CFR50;
- 10) Limita.tions on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40CFR190.
- f. Radiological Environmental Monitorina Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the 00CM, (2) conform to the guidance of Appendix 1 to 10CFR50, and (3) include the following:
- 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the 00CM;
- 2) A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of the census; and
- 3) Participation in an Inter-laboratory Comparison Program to ensure that independent checks on the precision and accuracy of measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
( ADMINISTRATIVE CONTROLS 1 This report shall'also include the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis af ter the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine 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 concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiolodine limit. ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT Thranue( 9 6.9.1.6 4euttne ra.,diological environmental operating reportAf' covering the b N trier to May 1 of each year. operation of the unit during the previous calendar year shall be 4he-4 ni t lebeeport-shal44e-submi tted-p r4 c r-to
-May-bof-the-year-(014owing3MtiM ;riti::1itp-Atautetukypwysom t v,
~
0.0.1, b The annuat-eadioTagical :n dreamen,ta. Operating reports #shall include summaries, interpretationsh and an analysis of trends of the resuJts, of the
\y $
radiological environmental D sveve441ance m4ui'ias for the report %riod .s.
& ding a repan4+on eit..-preoperational-studies,-operational dntroisa(as- fL_ a 1%
.*d4 g [ @-49propr44te), :nd predows-environmental-surve4444nce-reports-and-an y of-et-observed-impec ts-of-the-ple nt-opera t4en-on-the-me trennent . 'h: reports.
-.shalbahn inc.lude the-results-of-.4and-use-censuses-required-by-Specifica--
(3 t[ -440n-b121 If harmfuLef-fects-or-ev4dence-of-irrever&ible-damage-ar4-detected-
*3 h, . by the monitor 4n9r-the-report-sha14-prodde-en-anal
-planned-conse-of-action-to-alled4te the-pcoblem ysis-of-the-problem and ; -
M
-he-annual-eadiological-env4+onmental-oneceting-vepoet54ha11 ~includenumma*&ded-N Q and tabul+ted-results--in-the.4ermat-of-Requistoey'-Guide-4rerDecembee-49M4f-
~>44-ea d4 e leg teel-e nvi ronmen t al-s amp l e s-ta ke n-du ri ng-t he-report-pe riod.---4 n-the--
eent that-some-results-are-not-evai4able-f4 Mnc4us4en-with-the-repor4r-the--- g co' t shalbbe-submitted noting and axp.laining the reasons for the missing _,, estrttt--The-mitting-data-shell be 3tsbeitted-es-soon-as-poss4ble in 0
-trupplementary-report.
j he- repo r ts4ha l4-als o-i nc4 ude4he-fo l40wi ng t-a-s umma ry-d e scr i p t4 eof a the-gy q .cadiological-environmentabmonitoring prograim a may vi aM ramp &w lucations-(g h ug - keyed-to-a-table-giv4ng distances-and-direct 4ons-from-one-reactor; =n'4 +ha
-eesu l t s-o f-14cens ee-pa rt4cip a ti on-i n-the-I nte r4 abora tory-Compari son.hngram,.
,Q
-eequ& red 4y-Spec [f4 cat 4on-MM--
- 6. 9. l . ~7 N o C U.Se ct. .
k{ ' SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT D 6cntU1n%wtJ c./ y \ 6.9.1.8 Routiae. radioactive effluent release reports' covering the operation
+3 g of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The-peM od-of-the-f4est-report-shell begin e th-the4 ate of=2nitial eriticality.
SUMMER - UNIT 1 6-14 Amendment No.79 l
\
l
ADMINISTRAT!VE CONTROLS V
- 0. 0.1. 0 The-radioactive-of(hont-eefease reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste
] released f rom the unit.as-outl4ned-in-Regulatory-Gu4de-141, "Neesuringr {va l ua ti ngr-a nd -R e po r t i ng-R ad i oac t i v i ty-i n-Sol i d -Wa ste s-ano -R el ea s e s -o f- Rad i o-ac t4 ve-Materials-in41 uid-and40seous-E 4 f thents-f rom 44ght+ Water 2 Cooled-- NuclearJower PJants,'LRevisjon_pne_1974, with data 4ii-marized-ona-
-quar-ter4y-bas 4 6-404404ng thc feemat-of- A (a) et-nsiMwitn ik e/ pes oad HGwaasawrera4 w x>ppend
~ /.w m xi ee B m -ocm t heWreoW fL. "T/* * %Md42
'.n 1 4dd /* *'#45 N_ gr I /:> w( e~radioacthe-eHhent-reteate report-to-be-submitted-withie40-days-after- d Ja ary 1 of each year shall include an annual summary of hourly meteorologica/
8 data ollected over the previous year, This annual summary may be either in
;he fo'qn of an hour-by-hour listing of wind speed, wind direction, and atmy,/
sphericNtability, and precipitation (if measured) on magnetic tape, or jft the form of jh nt frequency distributions of wind speed, wind direction, aryd itmospheric tability. This same report shall include an assessment df the adiation dos. due to the radioactive liquid and gaseous effluentyj released from the unit o station during the previous calendar year. Thip'same report shall also inclu an assessment of the radiation doses from r461oactive liquid and gaseous fluents to members of the public due to ,their activities inside the site boun ry (Figures 5.1-3 and 5.1-4) during the report. All assumptions used in ma 'ng these assessments (i.e. , specific activity, exposure time and location) shall Historical annual average meteorology or met eincludedinthesereports/. rological conditions conc rent with the time of release of radioactive mater is in gaseous effluent ( (as determined by sampling frequencyandmeasurement)sha1beusedfordeter/iningthegaseouspathway doses. The assessment of radiat on doses shall,be performed in accordance sith the 0FFSITE DOSE CALCULATION NVAL(00C3).
/
The radioactive effluent release repo to te submitted within 60 days af ter January 1 ofexposed each year shallofalso 'an assessment inclu (br c from reactorof radiation doses toand other releases the likely most member the pd nearby uranium fuel cycle sources (ipdludin doses from primary effluent pathways and direct radiation) forfthe previo 12 consecutive months to show conformance with 40 CFR 190, Environmental Radi tion Protection Standards for Nuclear Power Operation. Accept'able methods for alculating the dose contribu-tion from liquid and gaseous Affluents are given (i Regulatory Guide 1.109, Rev. 1. / The radioactive effluents release shall include the foll ing information for each type of solid wa,5(e shipped offsite during the report eriod:
- a. Container / volume,
/'
- b. Totpl curie quantity (specify whether determined by mea rement or c./e orPrincipal t'imate), radionuclides (specify whether determined by measu ment estimate),
/,
'd. Type of waste (e.g., spent resin, compacted dry waste, evaporator
/ bottoms),
v SUMMER - UNIT 1 6-15 Amendment No. H , 79 l __.__ I Y N__*-~_i_t b _-_ r
ADHfNISTRATIVE CONTROLS vq~SoTTtH
- f. T pe of tion container (e.g. , LSA, Type A, Type B.ad' Large Quant agent (e.g., cement, urea formaldeh The radioactive effluent te aV orts shal ne site to unrestricted effluents on a quarterl areas of radi 3 Actt we e unplanned releases from aC ials in gaseous and liquid The radi
,. Swn. e fluent release reports shall include
- f. 4 . l.sC4Control Prooram (PCP) made durino the reportino oeriod.S.the Not:r w 4 N"
~
MONTHLY OPERATING REPORT 6.9.1.10 cluding documentation of all challenges to the PORV' , in-be submitted on a monthly basis to the Director, Office of U.S. Nuclear Regulatory Commission, Washington, D.C. Resource M
, shall anagement, month following the calendar month covered .
of each by the rep Any changes to the OFFSITE DOSE CALCULATIONeMANUAL with the(s) shall be Monthly Operating Report within 90 days in which the change In addition, a report of any major changes to thewas radioactive made effective. ment waste tre systems shall be submitted with the Monthly Operating Report for the pe in which the evaluation was reviewed and accepted as set above. forth in 6 5 CORE OPERATING LIMITS REPORT . 6.9.1.11 CORE OPERATING LIMITS REPORT prior to each reload remaining a. portion of a reload cycle, for the following: , or prior to any Moderator Temperature Coefficient BOL and EOL Limits and 300 p surveillance limit for Specification 3/4.1.1.3, b. c. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5 ,
- d. Control Bank Insertion Limits for Specification 3/4.1.3.6, Axial Flux Difference Limits, targu band, and APL HD for Specification 3/4.2.1, e.
Heat Flux Hot Channel for Specification 3/4.2.2,Factor, FhTP,K(I),W(I),APUO and W(Z)gt f. Nuclear Enthalpy Rise Hot Channel Factor, F and Power Factor Multiplier, PF3g, limits for Specification 3/4.2.3. The analytical methods used to determine the core operating limits shall be those previously described reviewed in the following and approved by the NRC, specifically those documents: a. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHO July 1985 (W Proprietary). , SUMMER - UNIT 1 i-16
, Amendment No. 35, 4 , 73 79, 88 ;
s _ I ADMINISTRAllVE CONVR0tS { e. Records of transient or operational cycles for those unit components identified in Table 5. 7-1.
- f. Records of reactor tests and experiments.
g. Records unit of staff, training and qualification for current members of the h. Records ofSpecifications. Technical in service inspections performed pursuant to these 1. Records of Quality Assurance activities as specified in she NRC's approved SCE&G position on Regulatory Guide 1.88, Rev. 2, October 1976. j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59. k. Records of meetings of the PSRC and the NSRC. 1. Records of the service lives of all hydraulic and mechanical snubbers defined in Section 3.7.7 including the date at which the service life commences and associated installation and maintenance records.
- m. Records of secondary wa+
sampling and water quality. n. Records of analysis required by the radiological environmental monitoring program,
- o. b.,As .4 etw p;., d b <6y
,6_.11 RADIATION PROTECTION PROGRAM N
- p
- de. h h cFF5vrE.tio 5E LAttwLMiod
'd h h ctss c.unoL Pe.o uA m.
Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater tnan 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
" Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they otherwise comply with approved radiation protection procedures for entry into high radiation areas.
SUMMER - UNIT 1 6-18 Amendment No 13 , H ,/ 5
'9
~ -
ADMINISTDTIVE CONTROLS b. I A radiation monitorin[ device which continuously integrates the radiation dose dose rate in the area and alarms when a preset integrated is received. may be made after the dose rate level in the area has beenEntry established and personnel have been made knowledgeable of them, c. A health physics Qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsibis for providing positive control over the activities witt.in the area and shall perform periodic radiation Physicist in the Radiation Work Permit. surveillance at the frequenc 6.12.2 with radiation levels such that a major portionn of the body co one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the admin-istrative control of the Shift Foreman on duty and/or health physics supervisi Doors shall remain locked except during periods of access by on. personnel under approved RWP which shall specify the dose rate levels in ths immediate work area The to prior maximum entry. allowable stay time for individual, in that areaseshall be establi levels such that a majorFor individual areas accessible to personnel with radiation in excess of 1000 mrem ** portion of the body could receive in one hour a dose-that are located within large areas, such as PWR can be reasonably constructed around the individual area bewarning a ropeddevice. off, conspicuously posted and a flashing light shall be activated as In lieu of the stay time specification of the RWP direct or remote be po; 1e by(suchpersonnel as use of closed qualified circuit in radiation TV cameras) protection procedures continuous to provid surveil ve expohure control over the activities within the area. e 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation. 4 6.13.2 N censee initiated changes to the PCP: Mg?- q --> 1. N Effluent Release Report for the period made. in whicShall be' submitted h d hange(s) was This submitt e 4 all contain:
- a. Sufficiently detailed a ion to totally support the rationale for Jt a ange withou nefit of additional or supplement 4 W nformation; g
determination that the change did not reduce overall conformance of the solidified waste product to exist for solid wastes; and , riteria
"" Measurement made at 18" from source of radioactivity.
SUMMER - UNIT 1 6 19 Amendment No.25, 79
i INSERT 4 b.13 PRJCf SS CONTROL PROGRAM 6.13.2 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.o. 1! 's documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s); and
- 2) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of federal, State, or other applicable regulations,
- b. Shall become effective after review and acceptance by the PSRC and approval of the General Manager, Nuclear Plant Operations.
I
ADMINISTRATIVE CONTROLS found accepta ntation of the fact that the change has 3 Ea m vt M th C 2. Shall C.s w e. becama-affective upon Niew and acce tineo-41 s% 6.14 0FFSITE 005E CALCULATION MANUAL (00CH) 6.14.1 The 00CM shall be approved by the Commission prior to implementation. D1 2 Licensee initiated changes to the 00CM: ' 1. with'In 40 days of the date the change (s) was vtC n made effecha11 be sub This submittaTM all contain: g56 a. Sufficient detailed information to tot d y support the rationale g for the chan information.geNithout benefit of additTonal or surelemental InforInation submit hould consist s a package of those pages of the O changed with each page numbered and provided with an ap priateanalysesore1diiluationslosb fying and date box, together the change (s); with appro-b. AdeterminJaic t the change will reliabi ty of dose calculations or setpo educe the accuracy or - eterminations, and
- c. Do mentation of the fact that the change has beeb viewed and found acceptable by the PSRC.
Shall become effective upon review and acceptance as set forth i 6.5 above. 1 15 and' Solid) MAJOR CHANGES TO RADIOACTIVE WASTE TREA NENT SYSTEMS (Liquid, G -
,/
6,15.1 (liquid, gaseoirkand solid). Licensee initiated major changes to the radioactive waste syste[ 1. N Shall be repo (ted to the Commission in the MonthI rating Report for the period T Nwhich the evaluation was rev by the Plant Safety Review Committee. The discussion fof e h change shall contain: a. AsummaryoftheehNationthat,dancewith10CFR5059 the change could be made- lee to the deternination that ccor . ; b. Sufficient detailed in,fermati M o totally support the reason forthechangewityh. benefit of4 dditional or supplemental information -
- c. A detailJd description of the equipment, com ents and proc n'es involved and tne interfaces with other ant systems; d.
An evaluation of the che le which shows the predicted r htases of radioactive material- .n liquid and gaseous effluents and/or-quantity of solid waste that differs from those previously predicted in the license application and amendments thereto; SUMMER - UNIT 1 6-20 Amendment No. U .79 f
_ _ _ i-si-l ADMINISTRA?!VE CONTROLS
- e. An evaluation of the change which shows the expected maxim exposures to individual in the unrestricted area and tMie general i
population that differ from those previousJy estimated elicenseapplicationandamendmentsthert6{
- f. A comparisD of the predicted releasspof radioactive materials, in % uid and gaseous $ffluents and in solid watte, to the actual rele'ha for thed sriod prior to when the changes are to be made; / I
- g. An estimate of t y t ure to 14 t operating personnel as a result of the-change; and
- h. Documen n of the fact that the change vihreviewed and Ed acceptable by thi PSRC.
2p /5 hall become effective upon 49fef and acceptance as set for bLn f 6.5 above. A M 5UM ER - UNIT 1 6-21 Amencment No. 33.79
l INSERT 5 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.14.2 Changes to the 00CM:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.o. This documentation shall contain- I l
- 1) Sufficient information to support the change together with I the appropriate analyses or evaluations justifying the change (s); and
- 2) A determination that the change will maintain the level of radioactive ef flut nt control required' by 10CFR20.106, 40CfR190, 10CFR50.36a, and Appendix ! to 10CFR50 and not adversely impact the accuracy or reliability of effluent dose or setpoint calculations.
- b. Shall become effective after review and acceptance by the PSRC and the approval of the General Manager, Nuclear Plant Operations,
- c. Shall be submitted to the Commission in the form of a complete legible copy of the entire 00CM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of'the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the _
change was implemented. t l
Enclosure 2 to Document Control Desk Letter TSP 890004-0 Page 1 of'1 , PROPOSED TECHNICAL SPECIFICATION CHANGE - TSP 890004-0 VIRGIL C. SUMMER NUCLEAR STATION DESCRIPTION AND SAFETY EVALUATION 1 DESCRIPT10H OF CHANGE ThisproposedchangetorevisetheRadiologicalEffluhntTechnical Specifications (RETS) is in accordance with Generic Letter 89-01. This proposed change will relocate the existing procedural details of the current RETS to the Offsite Dose Calculation Manual (00CM) and procedural details for the solid radioactive wastes to the Process Control Program (PCP). The proposed change will (1) incorporate programmatic controls in the Administrative Controls section of the Technical Specifications (TS) that ; satisfy the requirements of 10CFR20.106, 40CfR190, 10CFR50.36a, and Appendix I to 10CFR50; (2) relocate the existing procedural details in current specifications involving radioactive effluent monitoring instrumentation, the control of liquid and gaseous effluents, equipment requirements for liquid and gaseous effluents, radiological environmental monitoring, and radiological reporting details from the TS to the 00CM; (3) relocate the definition of solidification and existing procedural details in the current specification on solid radioactive wastes to the PCP; (4) simplify the associated reporting requirements; (5) simplify the administrative controls for. changes to the ODCM and PCP; (6) add record retention requirements for changes to the ODCM and PCP; and (7) update the definitions of the ODCM and PCP consistent with these changes. SAFETY EVALUATION This change is administrative in nature and is requested in conformance with Generic Letter 89-01 as part of the line-item TS improvement program. Relocating the procedural details of the current RETS to the ODCM and PCP will not reduce the level of radiological effluent control. This change provides programmatic controls for RETS consistent with regulatory requirements and allows future changes to these procedural details to be controlled by the controls for changes to the ODCM and PCP, included in the Administrative Controls section of the TS. These procedural details are not required to be included in the TS by 10CFR50.36a. The removal of procedural details from the TS has no impact upon the plant operation or safety. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. This proposed change is administrative and does not affect the level of. radiological
-effluent control.
Pursuant to the above information, this amendment request does not adversely affect or endanger the health or safety of the general public and does not involve an unreviewed safety question.
. to r ssument C attui ...s Letter TSP 890004-0 Page 1 of 2 PROPOSED TECHNICAL SPECIFICATION CHANGE - ISP 890004-0 VIRGIL C. SUMMER NUCLEAR STATION NO SIGNIFICANT HAZARDS EVALUATION FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS DESCRIPTION OF CHANGE This proposed change to revise the Radiological Effluent Technical Specifications (REIS) is in accordance with Generic letter 89-01. This -
proposed change will relocate the existing procedural details of the current RETS to the Offsite Dose Calculation Manual (0DCM) and procedural details for the solid radioactive wastes to the Process Control Program (PCP). The proposed change will (1) incorporate programmatic controls in the Administrative Controls section of the Technical Speciications (TS) that satisfy the requirements of 10CFR20.106, 40CFR190, 10CFR50.36a, and Appendix 1 to 10CFR50; (2) relocate the existing procedural details in current specifications involving radioactive effluent monitoring instrumentation, the control of liquid and gaseous effluents, equipment requirements for liquid and gaseous effluents, radiological environmental monitoring, and radiological reporting details from the TS to the ODCM; (3) relocate the definition of solidification and existing procedural details in the current specification on solid radioactive wastes to the PCP; (4) simplify the associated reporting requirements; (5) simplify the administrative controls for changes to the ODCM and PCP; (6) add record retention requirements for changes to the ODCM and PCP; and (7) update the defin.itions of the ODCM and PCP consistent with these changes. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERA710N The Commission has provided certain examples (51 FR 7744) of actions likely to involve no significant hazards considerations. The proposed amendment is - consistent with example (i) which states, "A purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature." Therefore, the Licensee has determined that a no significant hazards evaluation is justified and that should this request be implemented it will not:
- 1. Involve a significant increase in the probability or consequences of any accident previously evaluated because no plant equipment has been changed. The proposed TS change does not reduce the level of radiological effluent control, it will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the procedural details of the current RETS to the ODCM and PCP. These procedural details are not required to be in the TS by 10CFR50.36a.
Future changes to these procedural details will be controlled by the ODCM and PCP Administrative Controls sections of the TS. Records of reviews performed for changes made to the ODCM and PCP will be documented and retained for the duration of the operating license.
Enclosure 3 to Document Control Desk letter TSP 890004-0 Page 2 of 2 I Therefore, it is concluded that operation of the facility in accordance with this proposed change will not 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 previously evaluated. This proposed amendment does not modify the configuration of the facility or its mode of operation. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously identified.
- 3. Involve a significant reduction in a margin of safety. The proposed change does not affect the operation of the facility nor modify any method of radiological effluent monitoring or analysis. Therefore, it is concluded that operation of the facility in accordance with this proposed change will not involve a significant reduction in a margin of safety.
l
.j 4
Control Copy No, b h iYiI$$!$,7;'$ $ TIrf[$1 i lb 0d55 NLibh ['k h l'Q)d h,r$P'fl ikj ? 7 OFFSITE DOSE CALCULATION MANUAL , FOR SOUTH CAROLINA ELECTRIC AND GAS COMPANY l VIRGIL C. SUMMER NUCLEAR STATION ' l l PSRC Approval 04t\ &L uh /3/J'/9/ Qf '( ' Date Revision 15 February 1991 Reviewed by: at,.w n w / 2 9/ Date Approved by:' [ ~
~ '
c La . / 3-8-7/ Date k _
LIST OF EFFECTIVE PAGES Pane Revision Page Revnion i 15 1.0 35 15 ii 15 1.0 36 15 iii 15 1.0 37 15 iv 15 1.0 38 15 v 15 1.0-39 15 vi 13 1.0 40 15 vil 13 'i.0 41 15
- viii 13 1.0 42 15 ix 15 1.0-43 15 x 15 1044 15 xi 15 1.0 45 15 xii 15 1.04s 15 1.0 47 15 1.0 48 15 1.0 1 13 1.0 49 15 1.0 2 13 1.0 50 15 1.0 3 13 1.0 51 15 1.0 4 13 1.0-52 15 1.0 5 13 1.0 53 15 1.0 6 13 1.0-54 15 1.0-7 13 1.0 55 15 1.0 8 13 1.0 56 15 1.0 9 13 1.0 57 15 1.0 10 13 1.0 11 13 1.0 12 13 2.0 1 13 1.0 13 13 2.0 2 13 1.0-14 13 2.0 3 13 1.0 15 13 2.0-4 13 1.0 16 13 2.0 5 13 1.0 17 13 2.0 6 13 1.0 18 15 2.0 7 13 1.0-19 15 2.0-8 13 1.0 20 15 2.0 9 13 1.0 21 15 2.0 10 13 1.0 22 15 2.0 11 13 1.0 23 15 2.0-12 13 1.0 24 15 2.0 13 15 1.0.25 15 2.0 14 13 1.0 26 15 2.0-15 13 1.0-27 15 2.0 16 15 1.0 28 15 2.0-17 13 1.0 29 15 2.0 18 13 1.0-30 15 2.0 19 13 1.0 31 15 2.0 20 13 1.0-32 15 2.0 21 13 1.0 33 15 2.0 22 13 1.0 34 15 ODCM, V.C. Summer /SCE &G: Revision 15 (February 1991) i
LIST OF EFFECTIVE PAGE5jcontinued) Page Revision P_ age Revnion 2.0 23 13 3.0 29 13 2.0 24 13 3.0 30 13 2025 13 3.0-31 13 2.0 26 13 3.0 32 13 2.0 27 13 3.0 33 13 2.0 28 13 3.0 34 13 2.0 29 13 3.0 35 13 2.0 30 13 3.0 36 13 2.0 31 13 3.0 37 14 2.0 32 13 3.0 38 14 2.0 33 13 3.0 39 14 2.0 34 13 3.0 40 13 2.0 35 15 3.0 41 13 2.0 36 15 3.0 42 13 2.0 37 15 3.0 43 14 2.0 38 15 3.0 44 13 2.0 39 15 3.0 45 13 3.0 46 13 ' 3.0 47 '13 3.0 1 13 3.0 48 13 3.0 2 13 3.0 49 13 3.0 3 13 3.0 50 13 304 13 3.0 51 13 3.0 5 13 3.0 52 13 3.0-6 13 3.0 7 13 4.0 1 13 3.0 8 13 4.0 2 13 3.0-9 13 4.0 3 13 3.0 10 13 4.0 4 13 3.0 11 13 4.0-5 13 3.0 12 15 4.0 6 13 3.0 13 14 4.0 7 13 3.0 14 13 4.0 8 13 3.0-15 13 4.0 9 13 3.0 16 14 4.0 10 13 3.0 17 13 4.0-11 13 3.0 18 13 4.0 12 13 3.0 19 13 4.0 13 13 3.0 20 13 3.0 21 13 3.0 22 13 A1 13 3.0 23 13 A-2 13 3.0 24 13 A-3 13 3.0 25 13 A4 13 3.0 26 13 A5 13 3.0-27 13 A6 13 3.0 28 13 A-7 13 ODCM, V.C. Summer /SCE &G: Revision 15 (February 1991) il
1.1ST OF EFFECTIVE PAGES (continued) Page Revision A8 13 A9 13 A 10 13 A 11 13 A 12 13 A 13 14 A 14 14 . A 15 13 A 16 13 ODCM, V.C. Summer /SCE&G: Revision 15 (February 1991) lii
i CONTROLLED COPY DISTRIBUTION Person Copy # General Manager, Nuclear Plant Operations 1 , Manager, CHP& EP 2 Manager, Nuclear Licensing 3 Associate Manager, Health Physics 4 Senior Staff Health Physicist 5 Senior Staf f Health Physicist (2nd Copy) Sa Supervisor, RadWogical Analytical Services 6 Supervisor, Count Room 7 . Manager, Chemistry and Health Physics 8 Vice President, Nuclear 07erations 9 Supervisor, Count Room (2nd Copy) 10 Document Control & Records 11 Manager, Technical Oversight 12 Associate Manager, Quality Assurance 13 Supervisor, Environmental Programs 14 Resident NRC Inspector 15 Manager, Operations 16 Manager, Operations (2nd Copy) 17 Manager, Operations (3rd Copy) 18 I ODCM, V.C. Summer /SCE &G: Revision 15 (February 1990) iv
Table of Contentt MGE List of Effective Pages . . . . . . . . . . . . . i Controlled Copy Distribution List . . iv Table of Contents . . . . . . . . . . v List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . vii List of Figures .... . . . . . . . . . . . . . . . . . . . . viii References . .... ... . . . . . . . . . . . . . . . u. . . ix Introduction . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . x Responsibilities . . . . . . . . . .. . . . . . . .. . . . . , , . . . . . . xi 1.0 SPECIFICATION OF LIMITING CONDITIONS FOR OPE R ATION 1.1 Liquid Ef fluents . . . . . . . ...... . . . . . . . 1.0 1 1.1.1 Radioactive Liquid Effluent Monitoring Instrumentation . . .... . . . . . . 1.0 1 1.1.2 Liquid Effluents: Concentration . . . . . . . . . . 1.0 8 1.1.3 Li.1uid Effluents: Dose . . . . . . . . . . . 1.0 14
.1.1.4 Liquid Waste Treatment . . . . . . . . , 1.0 15 1.2 Gaseous E f fluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.0 17 l 1.2,1 Radioactive Gaseous Ef fluent Monitoring l Instrumentation . . .. .. . . , . . . . . . . . 1.0 17 1.2.2 Gaseous Effluents: Dose Rate . . . . . . . . . . . . . 1.0 23 1.2.3 Gaseous Effluents: Dose Noble Gas . . 4 . 1.0 26 1.2.4 Gascous Effluents: Dose Radioiodines, Tritium and Radioactive Materials in Particulate Form . . 1.0 27 1.2.5 Gaseous Radwaste Treatment -
. , . . . . . . . . . . . . . 1.0 28 1.3 Radioactive Effluents: To t al Dose . . . . . . . . . . . . , , . . 1.0 30 1.4 Radioloaical Environmental Monitorina . . . . . . . . . . , 1.0 32 1.4.1 Monitoring Prog ram . . . . . . . . . . . . . . . . . . . . 1.0 32 1.4.2 Land Use Census . . . . . . . . . . . . , , , , , , , . . 1.0-42 -
1.4.3 Interlaboratory Comparison Program . . . . . 1.0 44
- 1.5 Bases ...................... ............. . . . . . . . 1.0 45
-1,6 Reportina Requirements . . . . . . . . . . . . . . . . . . . , . . . . . 1.0 50 1.6.1 Annual Radiological Environmental Operating Report ................... ................. 1.0 50-1.6.2 Semiannual Radioactive Ef fluent Release Report 1.0 51 1.6.3 Changes to the ODCM . . . . . . . . . . . . . . . . . . . 1.0-53 1 6.4 Major Changes to Radioactive Waste Treatment
. stem (Liquid and Gaseous) . . . . . . . . . . . . . . 1.0 54 1.7 Defir%0ns . .............. . . . . . . . . . . . . . . . . . 1.0 56 ODCM, V.C. Summer /SCE AG: Revision 15 (February 1991) v F'yvy''* -tyWWV-'w-99*784"'W37-' 9 yrw kM_ .@s1 -'PrW==*=-Yid +t+4 P'1"**w-W' %'*e s F------_--'----"%A----
l ^ 2.0 llOUlO EFFLUENT 2.1 Liquid Effluent Monitor Setpoint Calculation 201 211 Liquid Effluent Monitor Setpoint Calculation Parameters . 202 2 1.2 Liquid Radwaste Effluent Line Monitors 2.0 6 2 1.3 Liquid Radwaste Discharge Via Industnal and Sanitary Waste System 2.0 14 2.1.4 Steam Generator Blowdown, Turbine Building Sump, and Condensate Demineralizer Backwash Ef fluent Lines . 2.0 15, 2.2 Dose Calculation for Liauid Effluents 2031 2.2.1 Liquid Effluent Dose Calculation Parameters 2.0 31 2.2.2 Methodology 2032 3.0 GASEOUS EFFLUENT , 3.0-1 3.1 Gaseous Efiluent Monitor Setpoints . 3.0 1 3.1.1 Gaseous Effluent Monitor Setpoint Calculation Parameters . 301 3.1.2 Station Vent Noble Gas Monitors . 3.0 5 3.1.3 Waste Gas Decay System Monitor 307 3.1.4 Alternative Methodology for Establishing Conservati've Setpoints . 208 3.2 Dose Calculation for Gaseous Effluent 3.0 12 3.2.1 Gaseous Effluent Dose Calculation Parameters 3.0 12 3.2.2 Unrestricted Area Boundary Dose ... 3.0 14 3.2.3 Unrestricted Area Dose to Individual 3.0 15 3,3 Meteoroloaical Model for Dose Calculations 3.0 45 3.3.1 Meteorological Model Parameters .. 3.0 45 3.3.2 Meteorological Model . 3.0 46 4.0 R ADIOLOGICAL ENVIRONMENTAL MONITORING 4.0 1 Appendix A Worked Examples of Monitor Setcoint Calculations and Dose Calculation A. RM L5, RM L7 and RM L9 . , A1 B. RM L3, RM L8, RM L10 and RM L11 . A6 C. RM A3 and RM A4 . . .. . A10 D. RM A10 . . A14 E, Alternate Methodology for Establishing Conservative Setpoints . A15 ODCM, V.C. Summer /SCE &G: Revision 13 (June 1990) vi
LIST OF TABLES Tgble No Pa g e_ No 1.1-1 Radioactive Liquid Effluent Monitoring instrumentation 1.0 2 112 Radioactive Liquid Effluent Monitoring instrumentation 105 Surveillance Requirements 113 Frequency Notation 10-7 1.1 4 Radioactive Liquid Waste Sampling and Analysis Program 1010 1.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1018 1.2 2 Radioactive Gaseous Effluent Monitoring l Instrumentation Surveillance Requirements 1021 { 1.2 3 Radioactive Gaseous Waste Samphng and Analys6 ! Program 1025 1.41 Radiological Environmental Monitoring Program 1.0 35 142 Reporting Levels for Radioactivity Concentrations 1040 in Environmental Samples Reporting Levels 1 4.3 Maximum Values for the Lower Limits of Detection (LLD)u Reporting Levels 1041 2.2 1 Broaccumulation Factors 2.0 34 2.2 2 Adult Ingestion Dose Factors 2.0-35 2.2 3 Site Related Ingestion Dose Commitment Factor (A,1) 2.0 37 311 Dose Factors for Exposure to a Semi Infinite Cloud of Noble Gases . . 3.0 4 3.2-1 Pathway Dose Factors for Section 3.2.2 2 (P,) 3018 322 Pathway Dose Factors for Section 3 2 3 2 (R,) 3021 3.2 3 Pathway Dose Factors for Section 3.2.3.3 (R,)(Inf ant) 3.0 24 3.2 4 Pathway Dose Factors for Section 3 2 3.3 (R,)(Child) 3.0 27 3.2 5 Pathway Dose Factors for Section 3.2.3.3 (R,)(Teenager) 3.0 30 3.2 6 Pathway Dose Factors for Section 3.2.3.3 (R,)(Adult) 3033 327 Controlling Receptors, Locations, and Pathways 3037 3.2 8 Atmospheric Dispersion Parameters for Controlling Receptor Locations ., . 3.0 39 3.2.9 Parameters Used in Dose Factor Calculations 3.0 40 4.0 1 Radiological Environmental Monitoring Program 402 ODCM, V.C. Su mmer/SCE & G: Revision 13 (June 1990) vii
Ll5T OF FIGURES Figurc tfo_ Page No 2.1-1 Example Liquid Monitor Calibration Curve 2.0 30 221 Liquid Radwaste Treatment System 2.0 39 3.1 1 Example Noble Gas Monitor Calibration Curve 3,0-11. 321 Gaseous Radwaste Treatment System . 3.0 44 3.3 1 Plume Depletion Effect for Ground Level Releases (6) . . 3.0 49 332 Vertical 5tandard Deviation of Materialin a Plume (ys) 3050 3.3 3 Relative Deposition for Ground Level Releases (0 9) 30 51 - 334 Open Terrain Recirculation Factor 3 0-52 4.0 1 Radiological Environmental 5ampling Locations (Local) 40 12 4.0 2 Radiological Environmental Sampling Locations (Remote) 4 0-13 O DCM, V.C. Summer /SCE &G: Revision 13 (June 1990) viis
REFERENCES
- 1. Boegli, T.S., R.R. Bellamy, W.L. Britz, and R.L Waterfield, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants" NUREG 0133 (October 1978).
- 2. " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR 50, Appendix I", U.S. NRC Regulatory Guide 1.109 (March 1976).
- 3. " Calculation of Annual Doses to Man frorn Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR 50, Appqndix 1", U.S. NRC Regulatory Guide 1.109, Rev.1 (October 1977)
- 4. " Final Safety Analysis Report", South Carolina Electric and Gas Company, Virgil C. Summer Nuclear Station.
- 5. " Operating License Environmental Report", South Carolina Electric and Gas Company, Virgil C. Summer Nuclear Station.
- 6. Wahlig, B.G., " Estimation of the Radioactivity Release Rate /Ecguilibrium Concentration Relationship for the Parr Pumped Storage System , Applied.
Physical Technology, Inc., February 1981. .
- 7. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases f rom Light Water - Cooled Reactors", U.S. NRC Regulatory Guide 1.111 (March 1976).
- 8. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors", U.S. NRC Regulatory Guide 1.111, Rev.1 (July 1977).
- 9. Slade, D.H.,(editor), " Meteorology and Atomic Energy"; U.S. Atomic Energy Commission, AEC TID 24190,1968.
- 10. " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants". U.S. NRC Regulatory Guide 1.21, Rev.1 (June 1974).
- 11. " Standard Radiological Effluent Technical 5pecifications for Pressurized Water Reactors", NUREG 0472, Revision 3 (January 1983).
- 12. " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment", USNRC Regulatory Guide 4.15, Revision 1 (February 1979).
- 13. " Age 5pecific Radiation Dose Commitment Factors for a One Year Chronic Intake" NUREG-0172 (November 1977).
ODCM, V.C. Summer /SCE &G: Reviscon 15 (February 1991) 1 ix
INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM)is an implementing and supporting document of the RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS). In accordance with USNRC Generit Letter 89 01, entitled " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program", the procedural details for implementing the Radiological Limiting Conditions for Operation have been incorporated into the ODCM.
- The ODCM describes the methodology and parameters to be used in the calculation of uffsite doses due to radioactive liquid and gaseous elfluents and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm / trip setpoints. The ODCM contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program. Configurations of the liquid and gaseous radwaste treatment systems an also included.
The ODCM will be maintained at the Station as the reference which details the Radiological Effluent Limiting Conditions for Operation of tM V. C. Summer Nuclear Station. Additionally the ODCM will Le maintained as the guide for accepted calculational methodologies. Changes a calculation methods or parameters will be incorporated into the ODCM in order to ensure that the ODCM represents the current methodology in all applicable areas. Computer software to perform described calculations will be maintained current witn this ODCM. ODCM, V.C. Summer /SCE &G: Revision 15 (February 1991) X
RESPONSIBluTIES I The ODCM contains the radiological effluent limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and their bases. Plant procedures implement responsibilities for compliance with the ODCM that include: The Operations group is responsible for;
- Declaring radioactive liquid and gaseous effluent monitor channels operable or inoperable.
- Ensuring the minimum number of operable channels for radioactive I quid and gaseous effluent monitors.
- Notifying the responsible group to implement appropriate action if less than the minimum number of radioactive liquid and gaseous effluent monitor channels are operable.
- Initiating an Off Normal Occurence Report in accordance with SAP 132, when less than the minimum number of channels operable condition prevails for more than 30 days.
e Restoring to within limits, the concentration of liquid radioactive material exceeding ODCM limits released from the site.
- Ensuring radioactive liquid and gaseous effluent monitor setpoints are set as prescnbed in the effluent release permit.
- Suspending release if radioactive liquid and gaseous effluent monitor setpoints are less conservative than ODCM requirements.
- Declaring liquid and gaseous radwaste treatment systems operable or inoperable, e Ensuring operability of gaseous and liquid radwaste treatment systems and ventilation exhaust treatment system.
- Ensuring appropriate portions of the gaseous and liquid radwaste treatment systems are used to reduce the radioactive materials in liquid and gaseous waste prior to their discharge when the projected doses exceed. limits specified by the ODCM.
- Initiating an Off Normal Occurrence Report in accordance with SAP 132, when liquid or gaseous radwaste system is inoperable for more than 31 days.
- Performing channel check and source check at the frequencies shown in Tables 1.12 and 1.2-2 for each radioactive liquid and gaseous of fluent monitoring instrumentation channel.
ODCM, V.C. Summer /SCE &G: Revision 15 (February 1991) xi
Instrumentation and Controls group is responsible for;
- Performing channel calibration and analog channel operational test at the frequencies shown in Tables 1.12 and 1.2 2 for each radioactive liquid and gaseous effluent monitoring instrumentation channel.
- Informing the Operations group of surveillance test results.
The Health Physics group is responsible for: i e Establishing setpoints for radioactive liquid and gaseous effluent monitors, consistent with ODCM methodology, and providing setpoint information to Operations. ' e Implementing remedial actions as requested by Operations. These actions include grab sampling and analysis and providing the results to Operations. e t'erforming periodic radioactive effluent monitor checks to determine backgrounds, normal indications and verifying monitor correlation graphs, and providing this information as necessary to Operations. e implementing radioactive gaseous and liquid waste sampling and analysis program in accordance with ODCM Tables 1.14 and 1.2 3. e informing Operations when at least one Circulating Water Pump or the Circulating Water Jockey Pump is required to provide dilution to the discharge structure.
- Calculating cumulative dose contributions and performing dose projections from liquid and gaseous effluents in accordance with the ODCM and providing the information to Operations.
e initiating an Off Normal Occurrence Report in accordance with SAP-132, when calculated dose from the discharge of radioactive materials in liquid or gaseous effluents are in excess of the limits specified by ODCM sections 1.1.3.1 or 1.2.3.1. e Initiating an Off Normal Occurrence Report in accordance with SAP 132, when liquid or gaseous waste is discharged without treatment and is in excess of the limits specified by ODCM sections 1.1.4.1 or 1.2.3.1. e initiating an Off Normal Occurrence Report in accordance with SAP 132, when the dose or dose commitment to any member of the public due to releases of radioactivity and radiation is in excess of 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 censecutive months. The Corporate Health Physics and Environmental Programs group is responsible for: e implementing the Radiological Environmental Monitoring Program as specified in Section 1.4 of the ODCM. e initiating an Off Normal Occurrence Report in accordance with SAP 132, when the Radiological Environmental Monitoring Program limiting conditions for
- operation are exceeded.
ODCM, V.C. Summer /SCE&G: Revision 15 (February 1991) xii
1.0 SPECIFIC ATION OF LIMITING CONDITIC N5 FOR OPE R ATION l 1.1 LIQUID Ef FLUENTS 1.1.1 RadioactivelLquid Ef fluent Monitorino Instrumentation llMITING CONDITION FOR OPER ATION 1.1.1.1 The radioactive hquid effluent monitoring instrumentation ghan-nets shown in Table 1.1 1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of ODCM Specification 1.1.21 are not exceeded The alarm / trip setpoints of these channels shall be determmed in accordance with ODCM, Section 2.1. APPLICABLE: At all Times ACTION: .
- a. With a radioactive liquid effluent monitoring anstrumentation channel alarm / trip setpoint less conservative than required by the above specification,immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channelinoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 1,1 1. Additionally if this condition prevails for more than 30 days, in the next semiannual effluent report explain why this condition was not corrected in a timely manner.
- c. The provisions of Technical Specifications 3.0.3 and 3 0.4 are not applicable.
SURVEll. LANCE REQUIREMENTS 1.1.1 2 Each radioactive liquid ef fluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the f requencies shown in Table 1.1-2. ODCM, V C. Summer, SCE & G Revision 13 (June 1990) 1 0-1
T A B L E _1 1 1 R ADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENT ATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1 GROSS RADIOACTIVITY MONITORS PROVID-
- ING ALARM AND AUTOMATIC TERMINA- '
TlON OF RELE ASE
- a. Liquid Radwaste tfiluent Line RM L5 1 1 or RM L9
- b. Nuclear (Processed Steam Generator) 1 1 Blowdown Effluent Line RM L7 or RM-L9
- c. Steam Generator Blowdown Effluent Line
- 1. Unprocessed during Power 1 2 Operation RM L10 or RM L3 '
- 2. Unprocessed dunng Startup RM- 1 2 L3
- d. Turbine Building Sump Ef fluent Line -
RM L8 1 3
- e. Condensate Demineralizer Backwash Effluent Line RM L11 1 6
- 2. FLOW RATE MEASUREMENT DEVICE 5'
- a. Liquid Radwaste Effluent Line Tanks 1 1/ tank 4 and 2
- b. Penstock Minimum Flow Interlock *
- 1 4
- c. Nuclear Blowdown Effluent Line 1 4
- d. Steam Generator (Unprocessed) 1 4 Blowdown Effluent Line
- 3. TANK LEVEL INDICATING DEVICES
- a. Condensate Storage Tank 1 5 In the event that simultaneous releases from both WMT and NBMT are required (which normally will be prevented by procedure) the flow rate for monitor RM L9 will be determined by adding flow rates for monitors RM L5 and RM L7.
Minimum dilution flow is assured by an interlock that terminates liquid waste releases if the minimum dilution flow is not available. ODCM, V C. Summer, SCE&G Revision 13 (June 1990) 10-2
TABLE 1 1 1 (Continuedl TABLE NOTATION ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 14 days provided that prior to initiating a reletse:
- a. At lease two independent samples are analyzed in accordance with ODCM Specification 1.1.2.4 ard
- b. At lease two technically qualified members of the Facihty Staff independently verify the release rate calculations and discharge line valving; ,
Otherwise, suspend release of radioactive effluents via this path-way. ACTION 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactivity (beta and gamma) at a limit of detection of atleast 1E 7 microcuries/ gram:
- a. At least once per 8 hours when the specific activity of the secondary coolant is greater than 0.01 microcuries/ gram DOSE EQUIVALENT l 131.
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/ gram DOSE EQUlVALENT l 131.
ODCM, V.C. Summer, SCE8G Revmon 13 (June 1990) 1.0 3
i I l l l l TABLE 111(Continued) T_A B_L_E NOT ATIO N ACTION 3 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, atleast once per 8 hours, grab samples are collected and analyzed for gross radioactivity (beta and gamma) at a limit of detection of at least 1E 7 microcuries/ gram. ACTION 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate, is estimated at least once per 4 hours during actual releases. Pump curves may be used to estimate flow. ACTION S With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid levelis estimated during all liquid additions to the tank to prevent overflow. ACTION 6 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 30 days provided that samples are analyzed in accordance with ODCM Specification 1.1.2.2 and Technical Specifi-cation 4.11.1.5. ODCM, V.C. Summer, SCE&G. Revision 13 (June 1990) 104
T ABLE 1,1-2 R ADIOACTIVE UQUID EFFLUENT MONITORING INSTRUMENT ATION SURVElLL ANCE REQUIRE ME NTS ANALOG CHANNEL CHANNEL OPERA-CHANNEL SOURCE C AllB R A- TIONAL INSTRUMENT CHECK CHECK TION TEST
- 1. GROSS R ADIO ACTIVITY MONI-TORS PROVIDING ALARM AND AUTOM ATIC TERMINATION OF RELEASE
- a. Liquid Radwaste Effluent D P R(2) Q(1)
Line RM L5 or RM L9
- b. Nuclear Blowdown D P R(2) Q(1)
Effluent Line RM L7
- c. Steam Generator D M R(2) Q(1)
Blowdown Effluent Line - RM L3, RM L10 '
- d. Turbine Building Sump D M R(2) Q(1)
Elfluent Line RM L8
- e. Condensate Demineralizer D M R(2) O(4)
Backwash Ef fluent Line RM-L11
- 2. FLOW RATE MEASUREMENT DEVICES
- a. Liquid Radwaste Effluent D(3) N.A. R O Line
- b. Penstocks Minimum Flow D(3) N.A. R O Interlock
- c. Nuclear Blowdown D(3) N.A. R Q Elfluent Line
- d. Steam Generator D(3) N.A. R Q Blowdown Effluent Line
- 3. TANK LEVEL INDICATING DEVICES
- a. Condensate Storage Tank D N A. R Q See Table 1.13 for explanation of frequency notation.
ODCM, V.C. Summer, SCE&G. Revision 13 (June,1990) 1.0 5
l TABLE 112 (Continued) TABLE NOTATION (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occursif any of the following conditions exists: 1
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
l
- 2. Loss of Power (alarm only). '
i
- 3. Low flow (alarm only).
- 4. Instrument indicates a downscale f ailure (alarm only). ,
- 5. Normal / Bypass switch set in Bypass (alarm only).
- 6. Other instrument controls not set in operate mode.
(2) The initial CHANNTL CAllBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and. Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CAllBRATION, sources that have been related te the initial calibration shall be used. (3) 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 on which continuous, periodic, or batch releases are made, (4) The ANALOG CHANNEL OPERATIONAL TEST thall also demonstrate that automatic isolation of this pathway and local panel alarm annunciation occurs if any of the 'ollowing conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint
- 2. Loss of Power (alarm only).
- 3. Lew flow (alarm only).
- 4. Instrument indicates a downscale failure (alarm only).
- 5. Normal / Bypass switch set in Bypass (alarm only).
- 6. Other instrument controls. not set in operate mode.
ODCM, V.C. Summer, SCE&G Revision 13 (June 1990) 1.0-6
4 TABLE 113 FREQUENCY NOTATION NOTATION FREQUENCY D At least once per 24 hours. W At least once per 7 days. , M At least once per 31 days. O At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. P Completed prior to each release. N.A. Not applicable. C ODCM, V.C. Summer, SCE&G Revmon 13 (June 1990) 1.0 7
1.1 2 Liquid Ef fluents Concentration LIMITING C_ONDITION FOR_OPER ATION 1.1 2.1 The concentration of radioactive material released from the site (see Technical Specification Figure 514) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table 11. Column 2 for radionuchdes other than dissolved or entrained noble gases For dissolved or entraened noble gases, the concentration shall be kmited to 2[. 4 microcuries/mi total activity. APPLIC ABL E . At all Times ACTION: With the concentration of radioactive material released from the site exceeding the above hmits, immediately restore the concentration to within the above limits. SURVEILL ANCE REQUIREMENTS 1.1.2.2 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 1.1-4. The results or pre release analyses shall be used with the calculational methods in ODCM Section 21 to assure that the concentration at the point of release is maintained w; thin the limits of ODCM Specification 1.1.2.1. 1 1.2.3 Post release analyses of samples composited from batch releases shall be performed in accorclance with Table 1.1-4. The results of the previous post-release analyses shall be used with the calculational methods in ODCM Section 2.1 to assure that the concentrations at the point of release were maintained within the limits of ODCM Specification 1.1.2.1. 1.1.2.4 The radioactivity concentration of liquids discharged from continu-ous release points shall be determined by collection and analysis of samples in ODCM, V.C. Summer, SCE &G Revi. on 13 (June 1990) 108
i l accordance with Table 1.14 The results of the analyses shall be used with the calculational methods in ODCM 5ection 21 to assure that the concentrations at the point of release are maintained within the limits of ODCM Specification 1 1 2 1. 1.1 2 S At least one Circulating Water Pump or the Circulating Water Jockey Pump shal' be determined to be in operation and providing dilution to the discharge structure at least once per 4 hours whenever dilution is required to meet the site radioactive effluent concentration limits of ODCM Specification 1 1 2,1 i ODCM, V.C. Summer SCE&G Revision 13 (June 1990) 109
T ABL E 11-4 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis (pCl/mL)* A. Batch Waste P P Release d Each Batch Each Batch Principal Gamma 5X107 . Tanks E mitters'
- 1. Waste I131 1X10 6 Monitor Tanks P M Bissolved and One Batch /M Entrained Gases 1X10 5
- 2. Condensate (Gamma emitters)
Demineralizer Backwash P M H3 1X103 b Receiving Each Batch Composite Tank
- 3. Nuclear Gross Alpha 1X10 7 Blowdown Monitor P Q Sr 89, Sr 90 SX10-8 Tink Each Batch Composite
- E f5 1X10 6 B. Continuous D W Principal Gamma 5X10 7 Release' Grab 5 ample Composite' Emitters'
- 1. Steam I-131 1X10 6 Generator Blowdown M M Dissolved and Grab Sample Entrained Gases 1X10 5 (Gamma emitters)
- 2. Turbine Building D M H3 1X10 5 Sump Grab Sample Composite (
- 3. Service Gross Alpha 1X10 7 Water Effluent D Q 5r89,5r90 SX10 8 Tank Grab Sample Composite (
Fe 55 1X10 6 See Table 113 for explanation of frequency notation. ODCM, V.C. Summer, SCE8G Revision 13 (June 1990) 1010
TABLE 1 14 (Continued) TABLE NOTATION l i
- a. The Lower Limit of Detection (LLD)is the smallest concentration of radioactive !
rnaterial in a sample that will yield a net count above background that will be detected with a 95% probabihty. LLD also yields a 5% probabihty of falsely concluding that a blank observation represents a "real" signal. . For a particular measurement system (which may include radiochemical separation): 4.66s b (ET(v)(2 22)(Y)(exp)( 4 a t) Where: LLD is the "a priori" lower limit of detection as defined above (as pCi. per unit mass or volume) Current literature defines the LLD as the detection capabihty for the instrumentation only and the MDC, the minimum detectable concentration, as the detection capabikty for a given instrument procedure and type of sample. 4.66 is a factor which corrects for the smallest activity that has a probabihty, p, of being detected, and a probability,1-p, of falsely concluding its presence. 4.66 2 24 V i + it,/ s, k = a constant w'rose value depends on the chosen confidence level (NRC recommends a confidence level of 95%)
= 1.6545 at 95% confidence level tb = background time ts = sample time sb is the standard deviation of the background counting rate or the counting rate of blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation), i l ODCM, V.C. Summer, SCE &G. Revision 13 (June 1990) I 1011
l T ABL E 1 14 (Contir)ued) TABLE NOTATION V is the sample size (in units of mass or volume), 2 22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when apphcable), A is the radioactive decay constant for the particular radionuclide, and a t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not enviro 1 mental samples). The value of s b used in the calculation of the LLD for a detection system shall be used on the actuoi observed vanance of the back - ground counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unvenfied theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma ray spectrometry the background should include the typical contributions of other radionuchdes normally present in the samples. Typical values of E, V, Y, and a t shall be used in the calculation. It should be recognized that the LLD is defined as an a pnon (before the fact) limit representing the capability of a measurement system and not as a posteriori(efter the fact) limit for particular measurement?
*For a more complete discussion of the LLD, and other detection limits, see the following:
(1) HASL Procedures Manual, H ASL-300 (revised annually). (2) Currie, L A., I
" Limits for Qualitative Detection and Quantitative Deter-(3) mination J.
Hartwell, Application to Radiochemistry" Anal Chem,40,586-93 (1968). K.,
" Detection Limits for Radioisotopic Counting Techniques."
Atlantic Richfield Handford Company Report ARH 2537 (June 22,1972). ODCM, V.C. Summer, SCE &G Revmon 13 (June 1990) 1012 -
I 1 ABLE 114 (Contmued) TABLE NOTATION b A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and m which the method of sampling employed resultsin a specimen which is representative of the liquids released.
- c. To be 'epresentative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be composited in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the elfluent release.
- d. A batch release is the discharge of hquid wastes of a discrete volurne Prior to; sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in ODCM 5ection 2.0, to assure representative sampling.
- e. A continuous release is the discharge of hquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release,
- f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn 54, Fe-59, Co 58, Co 60, 2n-65, Mo 99, Cs 134, C5137, Ce 141, and Ce 144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measursb!e and identifiable, together with the above nuchdes, shall also be identified and reported.
ODCM, V.C. Summer, SCE &G Revmon 13 (June 1990) 1.0 13
113 Liquid Effluents- Dose LIMITING CONDITION FOR OPER ATl_ON _ _ _ _ 1.1 3 1 The dose or dose commitment to an individual from radioactive materials in liquid effluents released f rom the site (see Technical Specification Figure 514) shall be limited:
- a. During any calendar quarter to less than or equal to t 5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
APPLIC ABLE: At all Times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause (s) for exceeding the limit (s) and defines the corrective actions to be taken to the releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with ODCM 5pecification 1.1.3.1.
b The provisions of Technical Specifications 3 0 3 and 3 0.4 are not applicable. SURVEILLANCE REQUIREMENTS 1.1.3.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with ODCM Section 2 2 at least once per 31 days. ODCM, V C. Summer, Su &G Revision 13 (June 1990) 1.0 14
i 1.1 4 pouid Waste Treatment I LIMITING CONDITION FOR OPER ATION 1.1.4.1 The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (See Technical Specification Figure.5.1-
- 4) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.
APPLICABLE: At all Times. l ACTION:
- a. With the liquid radwaste treatment system inoperable for more than 31 days or with radioactive hquid waste being discharged without treatment and in excess of the above limits,in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPER ABLE status, and
- 3. Summary dest.ription of action (s) taken to prevent a recurrence.
- b. The provisions of Technical Specifications 3.0.3 and 3 0.4 are not applicable.
ODCM, V.C. Summer, SCE&G Revision 13 (June 1990) 1015 _, _ __ _ ._. . _ _ _ . - _ . _ . . _ _ . . .._ _ _ . _ _ . _ . _ _ . , _ _ ~ - -.__ . . _ _ _ .
SURVEILLANCE REQUIREMENTS _ l 1.1.4 2 Doses due to liquid releases shall be projected at least once per 31 ! days,in accordance with ODCM Section 2 2. l i 11 A 3 The liquid f adwaste treatment system shall be demonstrated T"tMA"Li by opersting the liquid radwaste treatment system equipment for Sst 23 minute, at I.ast once per 92 days unless the kquid radwaste system
' as b 'n utilized t', r.'ocest ' .9 c e.tive liquid effluents during the previous 92
)
d ,3 l i I h i ODCM,V C. Summer,SCE&G Revision 13 (June 1990) l 1016 L
1.2 GASEOUS EFFLUENTS Radioactive Gaseous Effluent Monitorina instrumentation ] 1.2.1 LIMITING CONDITION FOR OPER ATION 1.2,1.1 The radioacuve gaseous effluent monitoring instrumentation channels shown in Table 1.2 1 shall be OPERABLE with their alarm / trip E setpoints set to ensure that the limits of ODCM Specification 1.2.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with ODCM Section 3.1. APPLICABLE: As shown in Table 1.2-1 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation; channel alarm /trin setpoint less conservative than required by the above ODCM Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channelinoperable.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 1.2-1. Additionally if this condition prevails for more than 30 days, in the next semiannual effluent report, explain why this condition was not corrected in a timely manner
- b. The provisions of Technical Specifications 3.0.3 and 3.0.4 ore not applicable.
SURVEILLANCE REQUIREMENTS 1.2.1.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and and ANALOG CHANNEL OPERATIONAL TEST operations at the f requencies shown in Table 1.2 2. ODCM, V.C. Summer, SCE &G R eosion 13 (June 1990) 1 0-17
TABLE 1.2-1 3 RADIOACTIVE GASEOUS EFFLUENT MONITOR!NG INSTRUMENTATION MINIMUM CHANNELS APPLICA-INSTRUMENT OPERABLE BILITY ACTION
- 1. WASTE GAS HOLDUP SYSTEM
- a. Noble Gas Activitv Monitor - 1 7 Providing Alarm and Automatic Termination of Release (RM-A10 or RM A3)
- 2. M AIN PLANT VENT EXH AUST SYSTEM
- a. Noble Gas Activity Monitor- 1 9 Providing Alarm and Automr(ic -
Termination of Release from Waste Gas Holdup System (RM-A3) 1 11
- b. lodine Sampler 1 11
- c. Part;culate Sample 1 8
- d. Flow Rate Measuring Device 1 8
- e. Sampler Flow Rate Measuring Device
- 3. RE ACTOR BUILDING PURGE SYSTEM
- a. Noble Gas Activity Monitor 1 10 Providing Alarm and Automatic Termination of Release (RM A4) *
- b. lodine Sampler 1 11
- c. Particulate Sample 1 11
- d. Flow Rate Measuring Device 1 8
- e. Sampler Flow Rate Measuring 1 8 Device ODCM, V.C. Summer, SCE &G . Revision 15 (February 1991) 1.0-18
4 T ABl.F 1.2-1 (Continued) TABLE NOTATION At all times during releases via this pathway. ACTION 7 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release:
- a. At least two independent samples of the tank's contents are analyzed, and
- b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 8 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION 9 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours. I ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1.0-19
TABLE 1.21 (Continued) TABLE NOTATION ACTION 10 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. . ACTION 11 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equip-ment as required in Table 1.2-3. ODCM, V.C. Summer, SCE &G: Revision 15 (February 1991) 1.0-20
TABLE 1.2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL ANALOG WHICH INSTRUMENT CHANNEL SOURCE CAllBRA- C ANNEL SURVEILL CHECK CHECK TlON ANCE RE-TIONAL l TEST QUIRED ;
- 1. WASTE GAS HOLDUP SYSTEM ,
- a. . Noble Gas Activity P 'P
- R(3) Q(1)
Monitor- RM A10 or RM A3
- 2. M AIN PLANT VENT EXHAUST SYSTEM l
- a. Noble Gas Activity D M
- R(3) Q(2) i Monitor - RM.A3 b lodine Sampler W N.A. N.A N.A. *
- c. Particulate Sampler W .N.A. N.A.
- N.A. -
- d. Flow Rate O N.A. R Q
- Measuring Device
- e. Sampler Flow Rate D N.A.
- R Q Monitor
- 3. REACTOR BUILDING PURGE SYSTEM
- a. . Noble Gas Activity D P,M
- R(3) Q(1)
Monitor - RM-A4
- b. lodine Sampler W N.A. N.A N.A. *
- c. Particulate Sampler W N.A. N.A. N.A. *
. d. Flow Rate Measur- .D N.A. R Q- -*
ing Device
- e. Sampler Flow Rate -D. N.A. R- Q
- Monitor
- See Table 1.1-3 for explanation of frequency notation.
ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1.0 21
TABLE 1.2 2 (Continuedl l TABLE NOTATION At all times. (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Loss of Power (alarm only).
- 3. Low flow (alarm only).
- 4. Instrument indicates a downscale failure (alarm only).
S. Normal / Bypass switch set in Bypass (alarm only).
- 6. Other instrument controls not set in operate mode.
(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
- 2. Loss of Power.
- 3. Low flow.
- 4. Instrument indicates a downscale f ailure.
S. Instrument controls not set in operate mode. (3) The initial CHANNEL CAllBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CAllBRATION, sources that have been related to the initial calibration shall be used. ODCM, V.C. Summer, SCE&G Revmon 15 (February 1991) 1.0-22
1.2.2 Gaseous Ef tluents: Dose Rate LIMITING _ CONDITION FOR OPERATION _ 1.2 2.1 The dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site (see Technical Specification Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mremlyr to the total body and less than or equal to 3000 mrem /yr to the skin, and '
- b. For all radioiodines and for all radioactive materials in particulate form and tritium with half lives greater than 8 days: Less than or equal to 1500 mremlyr to any organ.
APPLICABLE: At all Times. ACTION: With the dose rate (s) exceeding the above limits, immediately decrease the release rate to within the above limit (s). SURVElLLANCE REQUlf.EMENTS 1.2.2.2 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM. 1.2.2.3 The dose rate due to radiciodines, tritium and radioactive materiais in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of ODCM Section 3.2.2 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 1.243. ODCM, V.C. Summer, SCE&G Revmon 15 (February 1991) 1.0 23
TABLE 1.2 3 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detection Gaseous Release Type Frequency Frequency Analysis (LLD)(pCi/ml)* L. A. Waste Gas Stor- P P age Tank Each Tank Each Tank Principal Gamma 1X104 Grab Sample Emitters 9 B1 Reactor Building P P Principal Gamma
-36" Purge Line Each Purgebd Each Purge b Emitters 9 1X10 4
-6" Purge Line H-3 1X10 6 B2 Reactor Building M* MS Principal Gamma
-6" Purge Line Grab Sample Emitters 9 1X10 4 '
(if continuous) H-3 1X10 6 C, Main Plant Vent MD* MD Principal Gamma Grab Sample Emitters 9 1X10 4 H-3 1 X 10-6 D 1, Reactor Building Continuous W8 l-131 1 X1012 rurge Sampler' Charcoal I-133 1 X1010 Sample
- 2. Main Plant Vent Continuous Wa Principal Gamma 1 X 10-11 Samplerf Particulate Emitters 9 Sample I-131, oth ers Continuous M Sampler' Composite Gross Alpha 1X1011 Particulate Sample Continuous Q Sampler' Composite Sr-89,Sr-90 1X1011 Particulate Samole Continuous Nob e Gas Noble Gases 2X 10-6 Monitor Monitor Gross Beta See Table 1.1-3 for explanation of frequency notation.
ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) ) 1.0-24
l TABLE 1.2-3 (Continued) TABLE NOTATION
- a. See Table 1.1-4 notation (a) for definition of LLD.
- b. Analyses shall be also be performed within 24 hours following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period. -
- c. Tritium grab samples shall be taken at least once per 24 hours when the refueling canalis flooded.
- d. Samples shall be changed at least once per 7 days arid analyses shall be completed within 48 hours after changing (or after removal from sampler).
Sampling shall also be performed at least once per 24 hours for a least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
- e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.
- f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with ODCM Specifications 1.2.2.1,1.2.3.1 and 1.2.4.1.
- g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe 133m, Xe-135, and Xe- 138 for gaseous emissions and Mn 54, Fe-59, Co-58, Co-60,2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported.
Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. ODCM, V.C. summer, SCE&G: Revision 15 (February 1991) 1.0-25
I 1 2.3 Gaseous Ef fluents: Dose - Noble Gas LIMITING CON _DITION FOR OPE _ RATION 1.2.3.1 The air dose due to noble gases released m iscous effluents from the site (see Technical Specification Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,
- b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
APPLICABLE: At all Times. ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report required by ODCM section 1.6, prepara and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with ODCM Specification 1.2.3.1.
- b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1.2.3.2 Dose Calculations Cumulative dose cortributions for the current calendar quarter and current calendar year shall be determined in accordance with ODCM Section 3.2.3 at least once per 31 days. ODCM, V.C. Summer, SCE&G. Revision 15 (February 1991) 1 0-26 _ . _ _ _ _ _ _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . _ _ _ i eJu
1.2.4 Gaseous Effluents: Dose - Radioiodines, Tritium, and Radioactive Materials in Particulate Form. LIMITING CONDIT!ON FOR OPERATION 1.2.4.1 The dose to an individual from radioiodines, tritium, and rad;oactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents (see Technical Specification Figure 5,1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,
- b. During any calendar year: Less than or equal to 15 mrem to any I organ.
APPLICABLE: At all Times. . l ACTION:
- a. With the calculated dose from the release of tritium, radioiodines, and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents exceeding any of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to releases and the proposed actions to be taken to assure that subsequent release will be in compliance with ODCM Specification 1.2.4.1.
- b. The provisio .. of Technical Specifications 3.0.3 and 3.0.4 are rot applicable.
SURVEILLANCE REQUIREMENTS , 1.2.4.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with ODCM section 3.2.3 at least once per 31 days. ODCM, V.C. Summer, SCE&G Revision 15 (February 1991) 1.0 27
A 1.2.5 Gaseous Effluents: Gaseous Radwaste Treatment LIMITING CONDITION FOR OPERATION 1.2.5.1 The GASEOUS RADWASTE TRE ATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to ,their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (See Technical Specification Figure S.1-3), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site when averaged over 31 days would exceed 0.3 mrem to any organ. - APPLICABLE: At all Times. ACTION:
- a. With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than-31 days or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by ODCM section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsyste as 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.
ODCM, V.C. Summer, SCE&G' Revision 15 (February 1991) 1.0 28
- b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1.2.5.2 Doses due to gaseous releases from the reactor shall be projected at least once per 31 days, in accordance with ODCM Section 3.2.2 for air doses and ODCM Section 3.2.3 for organ doses. 1.2.5.3 The GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by opera-ting the GASEOUS RADWASTE TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEM equipment 'or at least 30 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous ?? days. ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1,0 29
1.3 RADIOACTIVE EFFLUENTS: TOT AL DOSE Ll_MITING CONDITION FOR OPER ATION 1.3.1 The dose or dose commitment to any member of the public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 n$ rem) over 12 consecutive months. APPLICABLE: At all Times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of ODCM S p e ci ficati o n 1.1.3.1.a, 1.1.3.1. b, 1.2.3.1.a , 1. 2.3.1. b, 1.2.4.1.a , o r 1.2.4.1.b,in lieu of any other report required and ODCM Section 1.6, prepare and submit to the Commission, within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of ODCM Specification 1.3.1. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of ODCM Specification 1.3.1, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Specia'l Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including information of 5190.11 (b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose ODCM, V.C. Summer,5CE&G: Revision 15 (February 1991) 1.0-30
4 l limitation of 10 CFR Part 20, as addressed in ODCM Specifications 1.1.2 and 1.2.2.
- b. The provisions of Technical Specifications 3 0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1.3.2 Dose _ Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with ODCM Specifications 1.1.3.2,1.2.3.2 and 1.2.4.2. ODCM, V.C. Summer, SCE&G- Revision 15 (February 1991) 1.0 31
4 1.4 RADIOLOGICAL ENVIRONMENTAL MONITORING 1.4.1 Monitorino Prooram LIMITING CONDITION FOR OPERATION 1.4.1.1 The radiological environmental monitoring program shall be con-ducted as specified in Table 1.41. i l APPLICABILITY: At all times. i ACTION:
- a. With the radiological environmental monitoring program not being conducted as specified in Table 1.4-1 in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 1.+2 when averaged over any calendar quarter, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report. When more than one of the radionuclides in Table 1.4-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) + concentration (2) + . R 1.0 limit level (1) limit level (2) When radionuclides other than those in Table 1.4-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of ODCM Specifications 1.1.3.1,1.2.3.1 ODCM, V.C. Summer, SCE&G. Revision 15 (February 1991) 1.0-32
and 1.2.4.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
- c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 1.41,in lieu of any E other report required by ODCM Section 1.6 prepare and subniit to the Comn4ission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause of he unavailability of samples and identifies locations for obtaining repiacement samples. The locations from which samples were unavailable may then be deleted from those required by Table ' A-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.
- d. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1.4.1.2 The radiologicai environmental monitoring samples shall be collected pursuant to Table 1.4-1 and shall be ar,alyzed pursuant to the ; requirements of Tables 1.41 and 1.4-3. ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1.0-33
J Table 1.4-1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path. way and/or Minimum Number of Sample Locations and Sampling and Type & Frequency Sample Criteria for Selection Collection Frequency of Arialysis AmtORM:
- 1. Particulates A) 3 Indicator samples to be tak en at locations (in Continuous sampler Gross beta following filtei i ditferent sectors) beyond but as close to the operation with weekly change; quarterly esclusion boundary as practgable where the coilection composite (by location) for i highest offsste sectorial ground level gamma isotopic l concentrationsore antgipated (t)
- 8) 1 Indicator sample to be tak en in the sector Continuous sampier Gio5s beta following filter beyond but as close to the esclusion boundary as operation with *eekly change, quJrterly practicable corresponding to the residence collection composste ,by location) for having the highest anticipated of fsite ground ga m m a isotopic.
levelconcentrationordose (1) C) I lndicator sample to be tak en at the loc w of Continuous sampler Gross beta following filter one of the dairies most lik ely to be atf t 1) operation with weekly change; quarterly (2) collection composite (by location) for , ga m ma isotopic. 1
- 0) 1 Control sample to be tak en at a loca i at least Continuous sampler Gtoss beta following filter 10 air miles from',ne site and not in tt . most operation with week ly change; quarterly prevalent wind directions (1) collection composite (by location) for gamma ssotc*pic it. Radiciodine A) 3 Indicator samples to be tak en at two locations Continuous sampler Gamma Isotopic for I t 31 as given ni A above operation with weekly weekly canister collection
- 8) 1 Indicator sample to be tak en at the locat.M as Continuous sampler Gamma isotopic for 1 131 given m l 8 above. operation with weekly weekly canister collection C) 1 Indicator sample to be tak en at the location as Continuous sampler Gamma isotopic for 1-131 given in 4.C. above operation with weekly weekly canister collection.
D) 1 Control sample to be tak en at a location as Continuounampler Gamma Isotopic for H 3 t given in i D. above operation with weekly weekly canister collection. lit Direct A) 13 Indicator stations with two or rnore dose- Monthly or quarterly (3.5) Gamma dose monthly or meters to form art mner ring of stations m the 13 quarterly. accessible sectors erithm 1 to 2 miles of the plant.
- 8) 16 Indicator stations with two or more dosi. Monthly or quarterly.(3.5) Gamma dose monthly or meters to form an outer ring of stations in the 16 quarterly.
accessible sectors within 3 to 5 miles of the plant C) B Stations with two or more dosimeters to be Monthly or quarterly (3.5) Gamma dose monthly or placed m special mterest areas such as populd- quarterly. tion centers, nearby residences. Schools a nd m 2 or 3 areas to serve as control stations ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1.0-34
Table 1.41 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path-way and/or Minimum Number of Sample locations and Sampling and Type & Frequency Sample Criteria for Selection Collection Frequency of Analysis WATER 8ORNE: IV Surf a(e Water A) 1 Indicator sample downstream to be tak en at a fime com posite sam ples Gamfn a isotop4 monthly lo(ation which allows for mmng and dilution in with (olie(tion every with quarterly compoute the ultimate rece wing river month (corresponds to (by lo(ation) or monthly USGS continuous sampling sample to be analy red for site) (3) tritium (5)
- 8) 1 Control sample to be tak en at a location on the Time composite sam ples Gamma isotopg monthly receiving river suf ficiently f ar upstream such that with tollection every with quarterly com p00te no effects of pumped storage operation are month ((orresponds to (by lo(ation) or monthf y anticipated USGS continuous sam pling sample to be anaiyavd f or site) (3) tritium (5)
C) 1 indgator sample f rom a location immed.ately Time composite sample: Gamma notopic rnonthly upstream of the nearest downstream municipal with collection every with quarterly compoote water supply month ((orresponds to (by location) or moatt'ly U5GS (ontinuous sampling sampie to be analysed for ute) (3) tritium (5) D) 1 Ind<dtor sample to be tak en in the upper time compoute samples Gamma isotopic monthly reservoir of the pumped storage f acility in the with (oilection every with quarterly comporte plan' dMha'qe canal month (corresponds to (by location)or motsthly USGS continuous sam pling sample to be analyled f or Ote) (3) trit.um (5) E) 1 Indicator sample to be tak en in the upper Grab sampling monthly (3) Gamma notopic monthly reservoir's non. fluctuating recreational area with Quarterly (omriosi'e (by to(ation) or monthly sample to be analysed for tritium (5) F) 1 Control sam ple to be tak en at a location on a Grau samphng monthly (3) Gamma isotopic monthly separate unaf f e(ted w atershed reservoir with quarterly com poot e (by Iv(ation) or monthly sample to be analyJed 'or tritium (5) V Ground Water A) 2 Ind44 tor samples to be tak en within the Quarterly grab sampimg Gamma isotorm and tri-excluton boundarv and in the direction of (%) tium analyses quarterly (5) potentially affected ground water supplies B) 1 Control sample f rom unaffected location Quarterly grab sam pling Gamma notopic and tru (5) tium analyseg quarterly (5) VI Drinking Water A) 1 Indicator sample from a nearby public ground Monthly grab sampling (3) Monthly (3) gam ma water supply source notopic and gross beta analyses and guarterly (5) (ompoute for totium analyses B) 1 Ind<ator (finished water) sample f rom the Monthly composite Monthly (3) gamma nearest downstream water supply sampling isotopic and gross beta analyses and Quarterly (5) compoute for tritium analyses C) 1 Control (finished water) sample f rom the Monthly composite Monthly (3) gamma nearest unaffected pubhc water supply sampling isotopic and gross beta analyses and quarterly (5) compoute for tritium anaiyses ODCM, V.C. Summer, SCE &G : Revision 15 (February 1991) l 1.0-35 l
Table 1.41 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Emposure Path. way and/or Minimum Number of Sample locations and Sampling and Type & Frequency Sample Criteria for Selection Collectior1 Frequency of Analysis INGE $ TION: Vil Mdk(2) A) Samples from milking animals in 3 locations with- Semi monthly when Gamrra isotopic and l i 31 in 5 k m distance having the highest ome poten- anim ais are on pasture. (6) analysis semi-monthly (t>) tial if there are none then i sampie from mdking monthly other times (3) when animal s are on animals in each of 3 areas between 5 to 8 k m pasture, monthly (3) et distance where doses are calculated to be greater other times than 1 mtem per year B) 1 Contr01 sample to be tak en at the lo(ation of a Semi monthly when Gamma isotopic and I- t 31 dary greater than 20 miles destance and not in anim als ar e on pastur e, (6) analysis semi-monthly (6) the most prevalent wind direction (t) monthly othee times (3) when animal s are on pasture, monthly (3) at other times C) 1 indicator grass (forage) sample to be tak en at Monthly wben available t. imma isotopic one of the locations beyond but as cIJ5e to the (3) eiitiusion boundary as practicable where the highest of fsite sectorial ground level concentea+ tions are anticipated (1) D) 1 Indicator grass (f or age) sample to be tan en at G a m m a isotopic the location of Vil(A) above when animais ar 3 on Monthly when available pasture (3) E) I Controi grass (forage) sampie to be tak en at the Ga m m a isotopic location of Vil(B) above Monthly when avadable (3) Vtil Food Products A) 2 Samples of broadleaf vegetation grown in the 2 Monthly when available Gamma isotopic on edible nesrest offsite location of highest calculated (3) portion annuat average ground level O/O if milk sampling is not performed within 3 k m or if milk sampling is not performed at a location within 5-10 k m where the doses are calculated to be greater than I mrem /yr.
- 8) 1 Control sample for the same foods taken at a Monthly when a vailable Gamma isotopic on edible location at least 10 miles distance and not in the (3) portion most prevalent wind direction if mdk sampling is not performed within 3 km or if milk sampling is not at a locatior; within 5 to 8 k m where doses are calcul,sted to be greater than 1 mremlyr IX Fish A) 1 Indicator sample to be tak en at a location in the Semiannual (7) collection Gamma isotc pic on edible upper reservoir. of the following specie portions semia nnually typesif available. bass; bream, cra ppie, c a t fish.
carp, f orage fish (shad) B) 1 Indicator sample to be tak en at a location in the Semiannual (7) cc'ection Gamma isotopic on edible lower reservoir. of the following specie portions semiannually typesif availabte: bass; bream, crappie: catfish, carp; forage fish (shad) C) 1 Indicator sample to be tak en at a location m the semiannual (7) collection Gamma isotop4 on edible upper reservoir's non fluctuating recreational of the following specie portions semiannually area typesif available bass, bream, crappie, catfish. (arp, forage fish (shed) l ODCM, V.C. Summer, SCE&G Revision 15 (February 1991) ! 1.0 36
Table 1 A 1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station i Exposure Path-way and/or Minimum Number of Sample Locations and Sampling and Type & Frequency Sample Criteria f or Selection Collection Frequency of Analysis 1 Control sample to be tak en at a location on the Semiannual (7)(ollection Gamma isotopic on edible IX hsh ((ontinued) D) receiving river suffic ently f ar upstream such that of the f ollowing specie portions semiannually no ef fects of pumped storage c peration are typesif available bass, anticipated bream (rappie, (atfish, (arD forage f 6h hhad) AQUATIC: 1 Indicator sample to I.e tak en at a location in the Semiannual grab sem ple G a m m e isotopic K 5ediment A) ' upper reservoir (7) B) 1 Indicator sampie to be tan en at a location in the semiannual grab sam ple Gamma isotopic upper reservoit's non fluctuating recreational (1) area 1 indicator sampie to be tak en on the shorehne of semiennual grab sa mpie Gamma isotopic C) the lower resef voir (7) 1 Control sample to be tak en at a location on the Semiannual grab sam ple G a m m a isot opic D) re(eiving river sufficiently f ar upstream such that (1) no etfects of pumped storage operation are anticipated I l ODCM, V.C. Summer, SCE8G. Revision 15 (February 1991) 1.0-37
I 1 NOTES
- 1. Sample site locations are based on the meteorological analysis for the period of record as presented in Chapters 5 and 6 of the OLER.
- 2. Milking animal and garden survey results will be analyzed annually. Should the survey indicate new dairying activity, the owners shall be contacted,with regard to a contract for supplying sufficient samples, if contractual arrange-ments can be made, site (s) will be added for additional milk sampling up to a total of 3 Indicator locations.
- 3. Not to exceed 35 days,
- 4. Time composite samples are samples which are collected with equipment capable of collecting an aliquot at time intervals which are short (e.g.,
hourly) relative to the compositing period. S. At least once per 100 days.
- 6. At least once per 18 days.
- 7. At least once per 200 days.
NOTE: Deviations from this sampling schedule may occasionally be necessary if sample media are unobtainable due to hazardous conditions, seasonal unavailability, insufficient sample size, malfunctions of automatic sampling or analysis equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. Deviations from sampling analysis schedules will be described in the annual report. ODCM, V.C. Summer, SCE&G: Revmon 15 (February 1991) I 1.0-38
TABLE 1.4-2 Reporting levels for Radioactivity Concentrations in Environmental 5amples Reporting Levels Airborne Par- Food Water ticulate or Fish Milk Products Analysis (pCi/I) Gases (pCi/m3) (pCiikg, wet) (pCill) (pCi/Kg, we.t) H3 20,000(a) N.A. N.A. N.A. N.A. Mn 54 1,000 N.A. 30,000 N.A. N.A. Fe 59 400 N.A. 10,000 N.A. N.A. Co 58 1,000 N.A. 30,000 N.A. N.A. Co-60 300 N.A. 10,000 N.A. N.A. 2n 65 300 N.A. 20,000 N.A. N.A. Zr-95 400 N.A. N.A. N.A. N.A. Nb-95 400 N.A. N.A. N.A. N.A. 1-131 2 0.9 N.A. 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-140 200 N.A. N.A. 300 N.A. La 140 200 N.A. N.A. 300 N.A. (a) For drinking water samples. This is the 40 CFR Part 141 value. ODCM, V.C. Su mmer, SCE & G : Revision 15 (February 1991) 1.0 39
I TABLE 1.4-3 Maximum Values for the Lower Limits of Detection (LLD)a.c Reporting Levels Airborne Par- Food Water ticulate or Fish Milk Products Sediment Analysis (pCi/l) Gases (pCi/m3) (pCi/kg, wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, d ry Gross Beta 4 1 X 10-2 N.A. N.A. N.A. N.A. H-3 2000(b) N.A. N.A. N.A. N.A. N.A. Mn-54 15 l N A. 130 N.A. N.A. N.A. Fe-59 30 N.A. 260 N.A. N.A. N.A. Co-58 15 N.A. 130 N.A. N.A. N.A. Co-60 15 N.A. 130 N.A. N.A. N.A. 2n 65 30 N.A. 260 N.A. N.A. N.A. Zr-95 30 N.A. N.A. N.A. N.A. N.A. Nb-95 15 N.A. N.A. N.A. N.A. N.A. 1131 lb 7 X 10 2 N.A. 1 60 N.A Cs-134 15 5 X 10 2 130 15 60 150 Cs-137 18 6 X 10-2 150 18 80 180 Ba-140 60 N.A. N.A. 60 N.A. N.A. La-140 15 N.A. N.A. 15 N.A. N.A. ODCM, V.C. Summer, SCE&G. Revision 15 (February 1991) 1.0-40
TABLE 1.4 3 (Continued) I i TABLE NOTATION
- a. Table 1.4 3 lists detection capabilities for radioactive materials in environmental 1
samples. These detection capabilities are tabulated in terms of the lower limits of ; detection (LLDs). See Table 1.14 notation (a) for definition of LLD.
- b. LLD for drinking water samples.
l
- c. Other peaks potentially due to reactor operations (fission and activation products) which are measurable and identifiable, together with the radio-nuclic'es in Table 1.4 3, shall be identified and reported.
ODCM, V.C. Su mmer, SCE & G Revision 15 (February 1991) 1.0 41
, j 1.4.2 Land Ute Census ~ j i
LIMITING CONDITION FOR OPERATION
.1.4.2.1 A land use census shall be conducted and shall identify the location of ,
the neart ' milk animal,the nearest residence and the nearest garden
- of greater than S00 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles.
APPLICABILITY: At all times. ACTION:
- a. With a land use census identifying a location (s) which yields a calculated dose or
. dose commitment greater than the values currently being calculated in ODCM Specification 1.2.4.2, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the new location (s).
- 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 samples are currently being obtained in accordance with ODCM. Specification 1l4.1.1,in lieu of any other report required by ODCM Section 1.6, prepare and s Amit to the Commission with in-30 daysipursuant to Technical Specification 6.9.e, a Special _ Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30; days. The sampling. location,' excluding the control station location.
having the lowest
- c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.
- Broad leaf vegetation sampling may. be performed at the site boundary in the direction-sector.with the highest 57Uin lieu of the garden census, ODCM, V.C. Summer, SCE8G. Revision 15 (February 1991) 1.0 42
calculated dose or dose commitment (via the same exposure pathway) rnay be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. i SURVEILLANCE REQUIREMENTS l 1.4.2.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 using that information which will provide i i the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.
)
ODCM, V.C. Summer, SCE&G. Revision 15 (February 1991) 1.0-43
m 1.4.3 Interlabc atory Comparison Proaram LIMITING CONDITION FOR OPERATION 1.4.3.1 Analyses shall be performed on radioactive materials supplied as part of an interlaboratory Comparison Program which has been approved by the Commission. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report,
- b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.
5_URVEILLANCE REQUIREMENTS _ 1.4.3.2 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 (participants in the EPA crosscheck program shall provide the EPA program code designation for the unit).
>>y ODCM, V.C. Summer, SCE &G: Revision 15 (February 1991) 1.0 44
1.5 BASES i B/1.1 LIQUID E FFLUENTL B/1.1.1 Radioactive Liquid Effluent Monitorino insitumentation The radioactive liquid effluent instrument tici is provided to monitor and control, as applicable, the releases of radioactive ma'.erials in liquid effluents ciuring actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Fart 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criten 60,63 and 64 of Appendix A to 10CFR Part 50. B/1.1.2 Concentration This specification is provided to ensure that concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentrativ levels specified in 10 CFR Part 20, Appendix B, Table 11, Column 2. This limitation provides additional assurance that the levels of rao ; activa materials in l bodies of water outside the site will result in exposures within: (1) the Section ll.A design objectives of Appendix I,10 CFR 50, to an individual, and (2) the limits of 10 CFR 20.106 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the i assumption that Xe 135 the cc,ntrolling rcdioisotope and its MPC in air (submersion) : was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. B/1.1.3 Dose This specification is piovided to implement the requirements of Sections ll.A, Ill. A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section ll.A. of Appendix I. Tne ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix l to assune that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites veith drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radiowclide concentrations in the finished drinking water that are in excess of the requireme.its of 40 CFR 141. The dose calculations in the ODCM implement the requirementsin Section lil.A of Appendix l that conformance with guides of Appendix l be shown by calculational procedures based on models anc' data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in NUREG 0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 4.3. NUREG 0133 implements Regulatory Guide 1.109, Revision 1, October 1977 (section C.1 and Appendix A) and Regulatory Guide 1.113, April 1977. Regulatory Guide 1.109, October ODCM, V.C. Summer, SCE& G: Revision 15 (February 1991) 1.0 45 I
Bases (continued) 1977,is titled " 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" Regulatory Guide 1.113, April 1977,is titled " Estimating Aquatic Dispersion of Effluerits from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix 1" B/1.1.4 Liquid Waste Treatment The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to releate to the environment. The requirement that the appropriate portfons of this system be used when specified provides assurance that the releases of radioamve materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section ll.A of Appendix 1,10 CFR Part 50, for liquid offluents. B/1.2 J G SEOUS EFFLUENTS B/1.2.1 Radioactive Gaseous Effivent Men _i.orino Instrumentation The radioactive gaseous eminent !nstrumentation is provided to monitor and control, as applicable, the releases of re oactive materials in gaseous effluents dunng actual or potential releases of geseot effluents. The alarm / trip setpoir.ts for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABillTY and use of this instrumentation is consistent with the requirements of General Design Criteria 60,63 and 64 of Appendix A to 10 CFR Part 50. B/1.2.2 Dose Rate This specification is provided to ensure that the dose at any time at the site l boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentration of 10 CFR Part 20, Appendix B, Table ll, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individualin an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table 11 of 10 CFR Part 20 (10 CFR Part 20.106 (b)), For individuals who may at times be within the site boundary,the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal 3000 mrem / year to the skin. These release rate ODCM, V.C. Summer, SCE 8 G Revision 15 (February 1991) 1 0-46
Bases (continued) I limits also restrict, at all times, the corresponding thyroid dose rate above background l to a child via the inhalation pathway to less than or equal to 1500 mrem / year. l 1 B/1.2.3 Dose Noble Gases l This specification is provided to implement the requirements of Sections ll.B. Illa and IV.A of Appendix l,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section ll.B of Appendix 1. The ACTION statements provide the required operating flexibility and at tha 9me time implement the guides I set forth in Section IV.A of Appendix I to aswre t":,i the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably ach!evable". The surveillance Requirements implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix i be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated, The 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 NUREG 0133," Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 5.3. NUREG 0133 implements Regulatory Guide 1.109, Revision 1, October 1977 and Regulatory Guide 1.111, Revision 1, July 1977. Regulatory Guide 1.109 is entitled " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR l' art 50, . Appendix 1, " Revision 1, October 1977 and Regulatory Guide 1.111 is entitled
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors, " Revision 1, July 1977. The ODCM equations piovided for determining the air doses at the site boundary are based upon the historical average atmospheric conditions.
B/1.2.4 Dose Radioiodines, Tritium and Radioactive Materials in Particulate Form This specification is provided to implement the requirements of Sections ll.C, Ill.A and IV.A of Appendix l,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section ll.C of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requnements implement the requirements in Section Ill.A of Appendix l that conformance with the guides of Appendix i be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways in 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 NUREG 0133," Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 5.3. NUREG 0133 implements Regulatory Guide 1.109, Revision 1, October 1977 and Regulatory Guide 1.111, Revision 1, July 1977. Regulatory Guide 1.109 is entitled " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1, " Revision 1. October 1977 and Regulatory Guide 1.111 is entithd ODCM, V.C. Summer, SCESG. Revision 15 (February 1991) 1.0 47
Bases (continued) t
" Methods for Estimating Atmospheric Transport and Dispersion of of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors, " Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, tritium, and radioactive materials in particulate form are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1)individualinhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto 9.*assy areas where milk animals and meat producing animals graze with consumption of the milk ahd meat by man, and 4) deposition on the ground with subsequent exposure of man.
B/1.2.5 G_asaous Radwaste Treatment The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 1J CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section ll.D of Appendix i to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections 11.8 and ll,C of Appendix I,10 CFR Part 50, for gaseous effluents. B/1.3 RADIOACTIVE EFFLUENTS: TOTAL DOSE The specification is provided to meet the dose i.mitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the , calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix 1. For sites containing up to 4 reactors,it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the i u!ividual reactors remain within the reporting requirement level. The Special Report wih describe a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to my member of the public is estimated to exceed the requirements of 40 CFR 190, the Sp il Report with a request for a variance (provided the release conditions resulting in vmiation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11,is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle. ODCM, V.C. Summer, SCE & G: Revision 15 (February 1991) 1.0 48
[ Uases(continued) B/1.4.1 Monitorino Prooram The radiological monitoring program required by this specification provides measurements of radiation of radioactive materialsin those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring progtam will be effective for at least the first three jears of commercial operation. Following this period, program changes may be initiated based on operational experience. The detection capabilities required by Table 1.4 3 are state of-the art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of 6 measurement system and not as a pisteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances rnay render these LLDs unachievable. In such cases, the contiibuting factors will be identified and described in the Annual Radiological Environmental Operating Report. B/1.4.2 Land Use Census This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting with iocal agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix 1 to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used,1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter. B/1.4.3 Interlaboratory Comparison Prooram The requirement for participation in an interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are per-formed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid. ODCM, V.C. Summer, SCE8G: Revision 15 (February 1991) 1.0 49
I 1.6 REPORTING REQUIREMENTS 1.6.1 Annual Radiological Environmental Operating Report 1.6.1.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be si'bmitted prior to May 1 of the year following initial I criticality. 1.6.1.2 The annual radiological environmental operating reports shall include summaries, interpretations, and an analysis of tr inds of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by ODCM Specification 1.4.2.1. If harmful effects or evidence of irreversible damage are detecttd by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem. The annual radiological environmental operating reports shall include tummarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period in the event that some results are not available for inclusion with the report, the report shall ue submitted noting and explaining the reasons for missing results. The missing data shall be submitted as soon as possible in a supplementary report. The report shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the resuas of licensee participation in the Interlaboratory Comparison Program, required by ODCM Specification 1.4.3.1. ODCM, V.C. Summer, SCE&G. Revision 15 (February 1991) 1.0 50
4 l 1.6.2 Semiannual Radioactive Effluent Release Report 1.6.2.1 Routine' radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. 1.6.2.2 The radioactive effluent release reports shall include a summ'ary 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 Radioactive Materials in Liquid and Gaseous Effluents from Light Water-Cooled Nuclear Power Plants", Revision 1, June ' 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. ' The radioactive effluent release report to be submitted within 60 days after January 1 of each year shallinclude an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figures 5.13 and 5.14 of the VCSNS Technical Specifications) during the year. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. Historical annual average meteorology or meteorological condition, 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 shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM). ODCM, V.C. Summer, SCE8G. Revision 15 (February 1991) 1.0 51 wwmy-wv+wwr=. e - - - - ' - e- ~ ' ' - - "''fge'-+yWe"9'F-+v'--W'*---TP9*+---
The radioactive effluent release report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1. The radioactive ef fluent release reports shall include unplanned releases from site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. ODCM, V.C. Summer, SCE & G: Revision 15 (February 1991) 1.0 52
9 1.6.3 Chances to the ODCM 1.6.31 Licensee init'ated changes to ODCM:
- 1. Shall be submitted to the Commission in the Monthly Operating Report within 90 days of the date the change (s) was made effective. This submittal shall contain;
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information.
Information submitted should consist of a package of these pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);
- b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PSRC.
- 2. Shall become effective upon review and acceptance as set forth in Technical Specification 6.5.
ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 1.0 53
* ~.
1.6.4 Maior Chanaes To Radioactive Waste Treatment Systems (Liquid and Gaseous) l 1.6.4.1 Licensee initiated major changes to the radioactive waste systems (liquid and gaseous):
- 1. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Plant Safety Review Committee. The discussion of each change shall contain:
- a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; i l l b. Sufficient detailed mformation to totally support the reason for the l change without benefit of additional or supplementalinformation; !
l
- c. A detailed description of the equipment, components and processes i i
involved and the interfaces with other plant systems; l l
- d. An evaluation of the change which shows the predicted releases or radioactive materials in liquid and gaseous effluents that differs from those previously predicted in the license application and amendments thereto;
- e. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;
\
- f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents, 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 i
ODCM, V.C. Summer, SCE 8 G: Revision 15 (February 1991) l 1.0 54
\
- h. Documentation of the fact that the change was reviewed and found acceptable by the PSRC.
- 2. Shall become effective upon review and acceptance as set forth in Technical Specification 6.5.
l
. l l
1 l l l l 1
)
l i I i ODCM, V.C. Summer, SCESG:- Revision 15 (February 1991) 1.0 55
1.7 Definitions ACTION 1.7.1 ACTION shall be that part of a specification which prescribes rneasures required under designated conditions. ANALOG CHANNEL OPERATIONAL TEST 1.7.2 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of *a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy. CHANNEL CAllBRATION - 1.7.3 A CHANNEL CAllBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known valur of input, The CHANNEL CAllBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions, and may be performed by any series of sequential, overlapping or total channel steps such , that the entire channel is calibrated. I CHANNEL CHECK 1.7.4 A CHANNEL CHECKS shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. GASEOUS RADWASTE TREATMENT SYSTEM
- 1.7.5 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to _ reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the-environment.
ODCM, V.C. Summer, SCE &G Revision 15 (February 1991) 1.0 56
OPERABLE OPERABILITY ; 1.7.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). SOURCE CHECK 1,7.7 A SOURCE CHECK shall be the quaktative assessment of channel response when the channel sensor is exposed to a radioactive source. VENTILATION EXHAUST TREATMENT SYSTEM 1.7.8 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal , absorbers and/or HEPA filters for the purpose of removing lodines or particu-lates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). t Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. l I 5 ODCM, V.C. Summer, SCE 8 G: Revision 15 (February 1991) 1057
, . . . - _ _ - - - - _ . ..--.-.....--.-._a._.--..--. - - - .. . - . -l
2.0 LIQUID EFFLUENT 2.1 Liquid Effluent Monitor Setpoint Calculation The Virgil C. Summer Nuclear Station is located on the Monticello Reservoir which provides supply and discharge for the plant circulating water, This reservoir also provides supply and discharge capacity for the Fairfield Pumped Storage Facihte The Parr Reservoir located below the pumped storage faci! sty is formed by the Parr Dam. ' There are tv o analyzed release pathways and sources of dilution for liquid effluents: the circulating water discharge canal and the liquid effluent line to the penstocks of the pumped storage facility. All liquid effluent pathways discharge to one of these release points. Generally speaking, very low concentrations of radioactive waste are discharged to the circulating water discharge while higher concentrations of radioactive waste are, released to the penstocks of the pumped storage f acihty during the generation cycle. The calculated setpoint values will be regarded as upper bounds for the actual setpoint adjustment. That is, setpoint adjustments are not required to be performed if the existing setpoint level corresponds to a lower count rate than the calculated value. Setpoints may be established at values lower than the calculated values,if desired Calculated monitor setpoints may be added to the ambient back-ground count rate. GENERAL NOTE: If no discharge is planned for a specific pathway or if the sum of the effluent concentrations of gamma emitting nuclides equals zero, the monitor setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm should an inadvertent release occur. ODCM, V. C. Summer, SCE&G: Revision 13 (June,1990) 2.0 1
i 2.1.1 Liauid Effluent Monitor Setpoint Calculation Parameters j Term Definition. Sution of Initial Use A = Penstock discharge adjustment factor which will allow 2.1.2 the set point to be established in a convenient manner and to prevent spurious alarms
= f t/fo, B = Steam Generator Blowdown adjustment factor which 2 1.4.1 will allow the set point to be estabhshed in a convenient manner and to preverst spurious alarms.
= fo/ fos C = the efiluent concentration limit (Specification 1.12) 2.1'2 implementing 10CFR 20 for the site,in uCi/ml.
C*
= the effluent concentration of alpha emitting nuclides 2.1.2 observed by gross alpha analysis of the monthly composite sample,in uCi/ml.
C, = the measured concentration of Fe 55 in liquid waste as 2.1.2 determined by analysis of the most recent available quarterly composite sample, in uCi/ml C i
= the effluent concentration of a gamma emitting nuchde, 2.1.2 g, observed by gamma ray spectroscopy of the waste sample, rttuci/ml. .
C. '
= the concentration of nuclide i,in uCumi, as determined 2.1.2 by the analysis of the waste sample.
C., = the concentration of radionuchde i,in uCi/mi,in the 2 1.2 Monticello Reservoir. Inclusion of this term will correct for possible long term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities. C, = the concentration of Sr 89 or Sr 90 in liquid wastes as 2.1.2 determined by analysis of the quarterly composite sample,in uCi/ml. C, = the measured concentration of H 3 in liquid waste as 2.1.2 determined b/ analysis of the monthly composite,in uCi/ml. c = the setpoint,in uCi/ml, oi'he radioactivity monitor 2 1.2 measuring the radioactivit; concentration in the ef-fluent line prior to dilution and subsequent release. This setpoint which is proportional to the volumetric flow to the effluent line and inversely proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value which,if exceed ed, would result in concentrations exceeding the limits of 10CFR 20 in the unrestricted area. c, = the monitor setpoint concentration for RM L7, the 2.1.2 2 Nuclear Blowdown Monitor Tank discharge line monitor, in uCi/ml.
- All concentrations are in units of uCi/ml unless otherwise noted.
ODCM, V. C. Summer, SCE &G Revision 13 (June,1990) 2.0 2
Term Deinition SC " f ni Use c,
= the monitor setpoint concentration for RM L9 the 2.1 2 3 comtened Liquid Waste Processing System and Nuclear Blowdown iystem ef fluent discharge line monitor, in uC /ml.
c '. r the meniior etpoint concentration for RM L11, the 2 1.4 1.3 Condensate l'eminerahzer Backwash discharge kne monitor,in ut. /mi-c,' a the monitor se coint concentrction for RM LS, the Waste 2121 Monitor Tank dacharge line monitor,in uCuml c,* = the monitor setocint concentration for RM L3 the initial 2.1 4.1.1 1 Steam Generator Blowdown Ef fluent line mon,itor,in . uCl/ml. c ."' = the monitor setpoint concentration for RM LIO, the final 2.1 4 1.1 Steam Generator Blowdown Effluent kne monitor,in uCi/ml. c' = the monitor setpoint concentration for RM L8, the 2.1 4.1 2 Turbine Building Sump Effluent kne monitor,in uCi/ml. CF D -= the Condensate Demineralize Backwash Effluent 2.1.4.1 Concentration Factor. CF 3 = the Steam Generator Blowdown Effluent Concentration 2.1 4.1 Factor. " CF, a the Turbine Building Sump Effluent Concentration 2 1.4.1 Factor. DF = the dilution f actor, which is the ratio of the total dilution 2.1 2 flow rate to the effluent stream flow rate (s). F = the dilution water flow setpoint as determined prior to 2.1.2 the release,in volume per unit time. F" = the flow rate of the Circulatina Water System during the 2.1 4.1 time of release of the TurbineT3uilding Sump and/or the Steam Generator Blowdown,in volume per unit time. F* = the dilution flow rate of the Circulating Water System 2.1.4.1 used for effluent monitor setpoint calculations, based on 90 percent o" expectec Circulating Water System flow rate during t ne time ol release and corrected for recir-culated Monticello Reservoir activity,in volume per unit time. F* = t 1e dilution flow rate through the penstock (s) receiving 2.1 2 11e radioactive liquid release upon which the effluent monitor setooint is based, as corrected for any recirculated radioactivity,in volume per unit time.
*(Conservatively this value will be either zero,if no release is to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted.)
ODCM, V. C. Summer, SCE&G. Revision 13 (June,1990) 203
Term Definition Settion of initia~I Use F, a the flow rate of water through the Fairfield Pumped 2.1.2 Storage Station penstock (s) to which radioactive hquids are being discharged during the period of elfluent release.This flow rate is dependent upon operational status of Fairfield Pumped Storage Station,in volume per unit time. i = the effluent line flow setpoint as determined for the 2 1.2 radiation monitor lowtion,in volume per unit time. f, = the maximum permissiole discharge flow rate for re- 2.L4.1 leases to the Circulating Water,in volume per unit time. f*'"'
= the flow rate of the Nuclear Blowdown Monitor Tank 2.1.2 discharge,in volurre per unit time, f, * = the flow rate of a Waste Monitor Tank discharge,in 2.1.2 volume per unit time.
f** = the flow rate of the Steam Generator Blowdown 2.1.4.1 discharge,in volume per unit time, f, a the flow rate of the tank discharge, either fdm or fdb,in 2.1.2 volume per unit time. - f, = the recirculation flow rate used to mix the contents of a 2.1.2 tank,in volume per unit time. i f, = the maximum permissible discharge flow rate for batch 2.1 2 releases to the penstocks,in volume per unit time. MPC = MPC. MPC,, MPC , MPC,, and MPC, = the limiting 2 1.2 conc 6,ntrations of,the appropriate gamma emitting, alpha emitting, and strontium radionuclides, Fe 55, and tritium, respectively, from 10CFR, Part 20, Appendix B, Table ll, Column 2. For gamma emitting noble gas radionuclides, MPC, = 2 x 10 8 uCi/ml. SF = the safety factor, a conservative f actor used to compen- 2.1 2 sate for engineering and measurement uncertainties SF
= 0.5, corresponding to a 100 percent variation.
[ Ci l, = the Lower Limit of Detection (LLD) for radionuclide iin 2.1 3 liquid waste in the Waste Monitor Tank, as determined by the analysis required in ODCM Table 1.14,in uCi/ml ( Ci]y = the concentration of radionuclide iin the waste cr n- 2.1.3 tained within the Waste Monitor Tank sereing as the holding facility for samphng and analysis prior to discharge,in uCi/ml. ODCM, V. C Summer, SCE&G: Revision 13 (June,1990) 2.0 4
Term Definition Section of Initial Use E Cg = the sum of the concentrations Cg of cach measured 212 g gamma emitting nuclide observed by gamma-ray spectroscopy of the waste sample,in uCi/ml. [E Cgln = the gamma isotopic concentrations of the Nuclear 2 1.2 g Blowdown Monitor Tank as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/ml. [E CgJn = the gamme isotopic concentrations of the Condensate 2.1 4.1 g Deminerahzer Backwash effluent (including solids) as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4, in uCi/ml. [E Cgly = the gamma isotopic concentrations of the Waste 2 1.2 g Monitor Tank as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/ml. [E CgJ 3 = the gamma isotopic concentrations of the Steam 2 1 4.1 g Generator Blowdown as obtained from the sum of the measured concentra tions determined by the analysis required in ODCM Table 1.1-4, in uCi/mli - [E Cg ], = the gamma isotopic concentrations of the Turbine 2.1 4.1 g Building Sump as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/ml. [E (CpMPCpJn = the sum of the ratios of the measured con centration of 2 1 4.1 nuclide i to its limiting value MPC, for the Condensate Demineralizer Backwash. [E (C i /MPCp)., .
= the sum of the ratios of the measured concentration of 2 1.4 1 nuclide i to its limiting value MPC, for the Steam Generator Blow down Effluent.
[E (Ci /MPCp]r = the sum of the ratios of the measured con centration of 2.1 4 1 nuclide i to its limiting value MPC, for the Turbine Building Sump Effluent. . [E (Ci/MPCp]' = the sum of the ratios of the measured con centration of 2 1.2 nuclide i to its limiting value MPC for the tank whose contents are being conside.ed for release. For a WMT, X
= M. For the NBMT, X = B.
t, = the mmimum time for recirculating the contents of a 212 tank prior to sampling,in minutes. V = the volume of liquid in a tank to be sampled,in gallons. 2.1.2 ODCM, V. C. Summer, SCE&G: Revision 13 (June,1990) 2.0 5
2.1.2 kquid Radwaste Effluent Line Monitors ' (RM LS, RM L7, RM L9) Liquid Radwaste Effluent Line Monitors provide alarm and auto- { matic termination of release functions prior to exceeding the concentration I limits specified in 10CFR 20, Appendix B. Table 11, Column 2 at the release point to the unrestricted area To meet this specification, the alarm / trip setpoints for liquid effluent monitors and flow measurement devices art set to assure that the following equation is sat.sfied: (1) es,f i l where: . C= the effluent concentration limit (Specification 1.12) implementing, l 10CFR 20 for the site in uComl. c= the setpoint, in uC /ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent kne prior to dilution and subsequent release; the setpoint, which is inversely proportional to the volumetric flow of the effluent line and proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value which,if exceeded, would result in concentrations exceeding the limits of 10CFR 20 in the unrestricted area. F= the dilution watcr flow setpoint as determined prior to the release point,in volume per unit t.me. f= the effluent line flow setpoint as determined at the radiation monitor location,in volume per unit time. At the Virgil C. Summer Nuclear Station the Liquid Waste Processing System (LWPS) and the Nuclear Clowdown System (NBS) both discharge to the penstocks of the Fairfield Pumped Storage (FPS) Facility through a ODCM, V, C. Summer, SCE&G. Revision 13 (June,1990) 2.0 6
common kne The available dilution water flow (F y,) is assumed to be 90 percent of the flow through the FPS penstock (s) to which liquid effluent is bemg discharged and is dependent upon operational status of the FP5 Facility. The waste tank flow rates (f, and f,) and the monitor setpoints (c,,, c, and cc ) are set to meet the condition of equation (1) for a given effluent concentration, C. The three monitor setpoints are determined in accordance with the monitor system configuration for this discharge pathway. The LWP5 discharges through RM LS, which has setpoint yc for alarm / control functions over releases from either Waste Monitor Tanks'1 or 9
- 2. The Nuclear Blowdown discharges through RM L7, which has setpoint c, for alarm / control functions over releases from the Nuclear Blowdown Monitor Tank. These two release pathways merge into a common line monitored by RM L9, which has setpoint cc for control functions over the common effluent line. Although the piping is arranged so that simultaneous batch releases from the two systems could be practiced, operational releases shall be from only one of the two batch systems aj any given. time. The, method by which their setpoints are determined is as follows:
The isotopic concentration for a waste tank to be released is obtained 1) from the sum of tM measured concentrations as determined by the analysis required in Table 1.1-4: 1 1 c, = 1 c, 4 c, + c, 4 r, + cf a e where: C, = the concentration of nuclide i,in uCi/ml, as determined by the analysis of the waste sample. Values for Ca, C , Ct and Cr will be based on most recent available 5 composite sample analyses as required by Table 1.1-4. ODCM, V. C. Summer, SCE&G Revision 13 (June,1990) 2.0 7
ECg = the sum of the concentrations C, of each measured gamma v emitting nuclide observed by gamma ray spectroscopy of the waste sample,in uCi/ml. C,* = the effluent concentration of alpha emitting nuchdes observed by gross alpha analysis of the monthly composite sample,in uCi/ml. C,* = the concentration of St 89 and Sr.90 in liquid waste as determined by analysis of the quarterly composite sample, in UCi/ml. C,* = the measured concentration of H 3 in hquid waste as determined by analysis of the monthly composite sample, in uCi/ml. C,* = the measured concentration of Fe-55 in hquid waste as determined by analysis of the quarterly composite sample,' in uCi/ml. The Cg term will be included in the analysis of each batch; terms for alpha, strontium, Fe 55, and tritium shall be included as appropriate
- lsotopic concentrations for both the Waste Monitor Tank; (WMT) and the Nuclear Blowdown Monitor Tank (NBMT) may be calculated using equation (2).
Prior to being sampled for analysis, the contents of a tank shall be isolated and recirculated The minimum recirculation time shall be: t, = 2V/f, (3) tr = the minimum time for recirculating the contents of a tank prior to sampling. V = the volume of liquid in the tank to be sampled. f, = the recirculation flow rate used to mix the contents of a tank. ODCM, V. C. Summer, SCE & G: Revision 13 (June,1990) 2.0 8
This is done to ensure that a representative sample will be obtained. Mechanical mixers ihall ensure a similar minimum turnover.
- 2) Once isotopic concentrations for either Waste Monitor Tank or the Nuclear Blowdown Monitor Tank have been determined, these values are used to calculate a Dilution Factor, DF, which is the ratio of dilution flow rate to tank flow rate (s) required to assure that the hmiting concentration of 10CFR, Part 20, Appendix B, Table 11, Column 2 are met at the point of discharge for whichever tank is having its contents discharged.
C ' I)F = \ - SF (4)
-- A!PC, '
c c c c e I'Y
- Sgp Sgp ggpc Sgfr
- G[ s
~ NY a e a a a where:
c Y 8
= the sum of the ratios of the measured concentration of
-' AfPC' nuclide i to its hmiting value MPC for the tank whose contents are being considered for release. For a WMT, X = M. For the NBMT, X = B.
MPC, = MPC,, MPC,, MPC,, MPC,, and MPC, = limiting concen-trations of the appropriate gamma emitting, alpha emitting, and strontium radionuchdes, Fe-55, and tritium, respectively, given in 10CFR, Part 20, Appendix B, Table 11, Column 2. For gamma emitting noble gas radionuclides MPC,is to be set equal to 2 x 10 8 pCi/ml, according to the Radiological Effluent Technical Speci-fications. SF = the safety factor; a conservative factor used to_com-pensate for engineering and measurement uncer-tainties.
= 0.5, Corresponding to a 100 percent variation.
ODCM, V. C. Summer, SCE &G: Revision 13 (June,1990) 2.0 9
- 3) The maximum permissible discharge flow rate, f,, may be calculated for the release of either the WMT or NBMT. First the appropriate Dilution Factor is calculated by applying equation (4), using the appropriate concentration ratio term (i e. M or B).
then, f, = u ti r (0) F,, > > t,, where: F, = dilution flow rate to be used in effluent monitor setpoint calculations, based on 90 percent FP5 Station expected flow rate, as corrected for any recirculated radioactivity: c' F,, a m 9 e r, t ! - g,' , ) (7) , where. F, = the flow rate through the Fairfield Pumped Storage Station penstock (s) to which radioactive liquids are being discharged F, should normally fall between 2500 and 44800 cis. C ,, = the concentration of radionuclide i,in uCuml,in the intake of Fairfield Pumped Storage Station (that is,in the Monticello Reservoir). Inclusion of this term will correct for possible long term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities. For expected discharges of liquid wastes, the summation will be much less than 10 and can be ignored (Reference 6). ODCM, V. C. Summer, $CE&G: Revision 13 (June,1990) 2.0 10
I l l 1 f, = the flow rate of the tank discharge, either af ., or f, ' o f, = flow rate of Nuclear Blowdown Monitor Tank discharge. (Conservatively this value will be either zero,if no release is to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted ) f, = flow rate of Waste Monitor Tank discharge. (Consdrva-tively this value will either be zero, if no telease is to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted,) DF = the Dilution Factor from Step 2. If f, : f,,, the release may be made as planned and the flow rate monitor setpnints should be established as in Step 4 (below). Because F,is normally very large compared to the maximum d:scharge pump capacities for the Waste Monitor Tank and the Nuclear Blowdown Monitor Tank, it is extremely unlikely that f, < fa ,. However, if a situation should arise such that f, < f,,, steps must be taken to assure that equation (1)is satisfied prior to making the release These steps niay include decreasinga f , by decreasing the flow rate of f, or f,, and/or increasing F, When new candidate flow rates are chosen, the calculations above should be repeated to verify that they combine to form an acceptable release, if they do, the estabhshment of flow rate monitor setpoints may proceed as follows in Step 4 If they do not, the choice of candidate flow rates must be repeated until an acceptable set is identified. Note that if DF '- 1, the waste tank concentration for which the calculation is being performed incluaes safety factors in Step 2 and meets the limits of 10CFR 20 without further dilution. Even though ODCM, V. C. Summer, SCE&G: Revision 13 (June,1990) 2.0 11
no dilution would be required, there will be no dinharge if minimum dilution flow is not available, since the penstock minimum flow interlock will prevent discharge.
- 4) The dilution flow rate setpoint', F,is estabbshed at 90 percent of the expected available dilution flow rate:
F = (0 9) F, (8) The flow rate monitor setpoint' for the effluent stream shall be set at the selected discharge pump rate (normally the maximum discharge pump rate or zero) f, or f, chosen in Step 3 above.
- 5) The radiation monitor setpoints may now be determined based on the values of ECi , F , and f which were specified to provide comph-ance with the limits of 10CFR 20, Appendix B, Table 11 Column 2. The.
monitor response is primarily to gamma radiation, therefore, the actual setpoint is based on EC,. The setpoint concentration, c,is determined as followv ccN]cya (9) e A= Adjustment f actor which will allow the setpoint to be established in a practical manner for convenience and to prevent spurious alarms. A = f,/r , 10 if A d 1, Calculate c and determine the maximum value for the actual monitor setpoint (cpm) from the monitor cahbra-tion graph.
- Set points for flow rates are administrative hmits.
ODCM, V, C. Summer, SCE&G Revision 13 (June,1990) 2.0 12
t if A < 1,No release may be made. Reevaluate the alternatives presented in Step 3. NOTE: If calculated setpoint values are near actual concentrations planned for release, it may be impractical to set the monitor alarm at this value. In this case a new setpoint may be calculated following the remedial methodology presented in Step 3 for the case of f, < f,. Within the limits of the conditions stated above, the specific monitor setpoint concentrations for the three liquid radiation monitors RM-L5, RM L7, and RM-L9 are determined as follows: 2.1.2,1 RM L5, Waste Monitor Tank Discharge Line Monitor: Cu5 1C s u 'O (11). 8 Cu is in uCi/mi
*See GENERAL NOTE under 2.1.
2.1.2.2 RM L7, Nuclear Blowdown Monitor Tank Discharge Line Monitor: Un$ 1 C, a('u (12) e Cs is in uCi/ml : NOTE: io no case should discharge be made directly from the Nuclear Blowdown Holdt ' Tank to the penstocks.
'See GENERAL NOTE under 2.1.
l l I l ODCM, V. C. Summer, SC E &G: Revision 15 (February 1991) 2.0 13
2.1.2 3 R_M L9, Combined liquid Waste Processina System and Nuclear Blowdown Waste Effluent Discharge Line Monity The monitor setpoint concentration on the common line, cc , should be the same as the setpoint concentration for the monitor on the active individual discharge line (i.e , c,,, or c, as determined above); eg s un < cu ,cy i (13)
'See GENERAL NOTE under 21 t in all cases, c,,, c,, and c c are the setpoint concentration values in uCi/ml. The actual monitor setpoints (cpm) for RM LS, RM L7, and.
RM L9 are determined from the calibration graph for the particular monitor. Initially, the calibra-tion curves were determined conservatively from f amihes of response curves supplied by the monitor manufacturers. A sample is shown in Figure 2.1-1. As releases occur, a historical correlation will be prepared and placed in service when sufficient data are accumulated. 2.1.3 Liquid Radwaste Discharae Via industrial and Sanitary Waste System (RM LS) in the Virgil C. Summer Nuclear Station liquid waste effluent system design, there exists a mechanism for dischargsag liquid wastes via the Industrial Sanitary Waste System. The sample point prior to discharge is one of the Waste Monitor Tanks. The analysis requicements are the requirements listed in Table 1.14. O DCM, V. C. Su mmer, SC E &G : Revision 13 (June,1990) 2.0 14
mumm e im l This effluent pathway shall only be used when the following conditio,. a met for all radionuclides,i: 5 C. u C. us; (14) c' = the concentration of radionuclide i in the waste con-V tained within the Waste Monitor Tank serving as the holding f acility for sampling and analysis prior to discharge,in uCuml-C. uj, = the Lower Limit of Detection,(LLD) for radionuclide i in the liquid waste in the Waste Monitor Tank as deter-mined by the analysis required in Table 1.14,in uCi/ml. When the conditions of equation (14) are met, liquid waste may be. released via the Industrial and Sanitary Waste System pathway. The RM L5 setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm should an inadvertent high concentration release occur. 2.1.4 Steam Generator L' lowdown, Turbine Building Sump, and Conden-sate Demineralizer B6c.kwash Ef fluent lines (RM L3, RM L10, RM L8, RM -L11) Concentrations of radionuclides in the liquid effluent discharges made via the Turbine Building Sump, Steam Generator Blowdown, and Condensate Demineralizer Backwash are expected to be very low or nondetectable. The first two releases are expected to be continuous in nature and the last a batch release. All will be sampled in an appropriate manner as specified in Table 1.14 of the ODCM. The Steam Generator Blow-down Monitors, the Turbine Building Sump Monitor, and the Condensate Demineralizer Backwash Monitor provide alarm and automatic termination of release prior to exceeding the concentration limits specified in 10CFR 20 Appendix B, Table 11, Column 2 at the release point to the unrestricted area. ODCM, V C. Summer, SCE&G: Revision 13 (June,1990) 2.0 15
In reality, all of these effluent pathways utilize the circulating water as dilution to the effluent stream, with the circulating water discharge canal being the point of release into an unrestricted area. However, to compensate for uncertainties in the transit times of activity discharge to the Industrial and Sanitary Waste System, discharges to that system ' not be credited with dilution for the purpose of monitor setpoint calculations. The Turoine Building Sur and Condensate Demineralizer Backwash Effluents enter Circulating Water via the (. umps and ponds of the Industrial and Sanitary Waste System. Steam Generator Blowdown dffh.ent may be released to the Circulating Water either directly in the Condenser outflow (the normal flow path) or in the first hours following startup via the Industrial and Sanitary Waste System for chemical reasons. For the sake of clarity, two mutually exclusive setpoint calculation processes are outlined below. Section 2.1.4.1 is to be used whenever Steam Genetator Blowdown is being released directly to the Circulating Water in the Condenser outflow, which is the normal mode. Section 2.1.4.2 is to be used whenever Steam Generator Blowdown is being released to the' Industrial and Sanitary Waste System, or diverted to the Nuclear Blowdown Processing System, both of which are alternate modes. Each section covers all four monitors (RM-L3, RM L8, RM-L10 and RM-L11',. Normally, water collected by the Nuclear Blowdown Processing Systern has very low specific activity. This water may be processed to the Turbine Building sump. NOTE: When Circulating Water is unavailable for effluent dilution, releases containing activity above LLD should be discouraged via pathways which lead to it. Steam Generator Blowdown should be diverted to the Nuclear Blowdrwn Processing System. Condensate Deminera-
~
lizer Backwash may be diverted to the Turbine Building Sump or not released. Turbine Building Sump effluent should be diverted to the Excess Liquid Waste Processing System. (These steps are to keep the calculated dose to individuals as low as raasonably achievable.) Furthermore, sampling and analysis of the Industrial and Sanitary Waste System is to be initiated and the measured concentrations used in the dose calcul6tions of Section 2.2. ODCM, V. C. Summer, SCE&G: Revision 15 (February 1991) 2.0 16
2.1.4.1 Steam Generator Blowdown Effluent Direct to Circulating Water (Normal Mode) Equation (1) is again used to assure that effluents are in compliance with the aforementioned specification: es ? W + /) The available dilution water flow (F,) is dependent ypon the mode of operation of the Circulating Water System Any change in this value will be accounted for in a recalculation of equation (1). The Steam Generator Blowdown flow rate (f,) and the Stearr Generator Blowdown monitor setpoints (c u and c ) are set to meet the condition of equation (1). The Turbine Building Sump and Condensate Dimeneralizer effluents will be limited to concentra-tions less than MPC without claiming dilution (see below). There-fore,it is not necessary to consider their flow rates or concer trations in determining the required dilu'. ion and monitor setpoints for Steam Generator Blowdown. For conseivatism, the Turbine Building Sump and Conden-sate Demineralizer Backwash monitor setpoints (c, and cg ) will claim no dilution from the Circulating Water, and will be set at the appli-cable concentration limit; That is: c~C (15) The Turbine Building Sump monitor, RM L8, alarms and terminates release upon exceeding the monitor setpoint (c.). The
. discharge can then be manually diverted to the Excess Waste Processing System. RM l.11, the Condensate Demineralizer Backwash monitor, alarms and terminates release upon exceeding the monitor setpoint (co). The discharge may then be manually oiverted to the Turbine Building Sump or simply delayed.
CDCM, V. C. Summer, SCE&G. Revision 13 (June,1990) 2.0 17
RM L3, the first monitor in the Steam Generator Blowdown discharge pathway, alarms and terminatet *elease of the stream. The discharge is then automatically diverted to the Nuclear Blowdown
)
Processing System. RM-L10, the last monitor in the Steam Generator Blowdown discharge pathway, alarms and terminates the release, i l Thus, RM L10 is redundant to RM L3 and the setpoint (c w ) will be ! determined in tne same manner as RM-L3 (c.,,). The method by which the monitor setpoints'are determined is as follows:
- 1) The isotopic concentrations for any release source to be or being released are obtained from the sum of the measured concentrations as determined in Table 1.14. Equation (2)is again employed for this calculation:
1 c, = 1 C, t C, t c, t C, 4 C7 i s where:
,E C, = the sum of the measured concentrations as determined by the analysis of the waste sample,in uCi/ml.
{C, = the sum of the concentrations C,of each measured gamma em:tting nuclide observed by gamma-ray spectroscopy of the waste sample,in utilml. C, = the measured concentration C, of alpha emitting composite sample,in uCi/ml. C, = the measured concentrations of Sr-89 and Sr 90 in liquid waste as determined by analysis of the most recent available quarterly composite sample, in uCi/ml. ODCM, V. C. Summer, SCE &G : Reviston 13 (June,1990) 2.0 18
C, = the measured concentration of H 3 in hquid waste determined by analysis of the monthly composite sample,in uCi/ml C, = the measured concentration of Fe-55 in liquid waste as determined by analysis of the most recent available quarterly composite sample, in uCi/ml. Isotopic concentrations for the Steam Generator Blowdown System effluent, the Turbine Building Sump Effluent, and the Condensate Demineralizer Backwash effluent may be calculated using equation (1). l l l
- 2) Once isotopic concentrations for the Steam Generator Blowdown have been determined, these values are used to calculate a Dilution Factor, DF, which is the ratio of the, total dilution flow rate to ef fluent stream flow rate required to assure that the limiting concentrations of 10CFR, Part 20, Appendix B, Table ll, Column 2 are met at the point of discharge.
c Dr= , , ' ,3 + SF (16) c c c c. c
* * ~
~
bil' : , All'ca ' All'C. AIPc. 7 Ifl i S where: C, = Co . C,, C,, C,, and C,, measured concentrations as defined in Step 1. Terms C,, C,, C,, and C, will be included in the calculation as appropriate. C V '
= the sum of the ratios of the measured concen-
^
7 Afl'C, tration of nuclide i to its limiting value MPC, for the Steam Generator Blowdown effluent. O DCM, V. C. Su mmer, SCE &G: Revision 13 (June,1990) 2.0 19
e . t MPC, = MPCq , MF C , MPC,, MPC,, and MPC, are limiting concentrations of the appropriate radionuclide f rom 10C R, Part 20, Appendix B. Table 11 Column 2 limits For gamma-emitting nv5le gas radionuct des, MPC is to be set equal to 2 x 104 uCiln SF = the same generic term as used in Section 2.1.2. Step 2.
= 0.5
- 3) The maximum perrnasible ef!!uent discharge flow rate, f a, may now be calculated for a release from the Steam Generator Blowdown.
"' (18) f, = -
the V, ., , y , where: F, = Dilution flow rate for use in effluent monitor setpoint calculations, based on 90 percent of the expected flow rate of the Circulating Water System during the time of release and corrected for any recirculated activity; c F ,, = m to Fa l1 - [ ','g l (19) where. F a
= the flow rate of the Circulating Water System during the time of the release.u F MW nomaHy fall between 1.78 X 105 and 5.34 X 105 gpm when the plant is operating and should be 5000 gpm when the plant is shi *down and the Circulating Water Jockey pump is operating ODCM, V. C. Summer, SCE 6G: Revision 13 (June,1990) 2.0 20
C,, = the concentration of radionuclide i,in uCi/ml,in the Circulating Water System intake, (that is, in the Monticello Reservoir). Inclusion of this term will correct for possible long-term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities. For expected discharges of liquid wastes, the summa- j tion will be much less than 1.0 and can be ignored ) (Reference 6). f, = Flow rate of Steam Generator Blowdown discharge. (This value normally will be either zero,if no release is to be conducted, or the maximum rated capacity of the discharge pump (250 gpm), release is to be conducted.) Note that the equation is valid only for DF > 1; for DF G 1, the effluent concentration meets the limits of 10CFR 20 without dilution as well as being in compliance with the conservatism imposed by the Safety Factor in Ctep 2. If f - a f,,, releases may be made as planned, Because F,is normally very large compared to the maximum discharge pump capacity of the Steam Generator Blowdown System, it is extremely unlikely that f, < f a,. However, if a situation should arise such that fa < f ,,a steps must be taken to assure that equation (1) is satisfied prior to making the release. These steps may include diverting Steam Generator Blowdown to the Nuclear Blowdown Processing System or-decreasing the effluent flow rate. When new candidate flow rates are chosen, the calculations above should be repeated to verify that they combine to form an acceptable release. If they do, the ODCM, V. C. Summer, SCE &G : ftevision 13 (June,1990) 2.0 21
establishment of flow rate monitor setpoints should proceed as follows in Step 4. If they do not provide an acceptable release, the choice of candidate flow rates must i be repeated until an acceptable set is identified.
- 4) The dilution flow rate setpoint for minimum flow rate, F, is established at 90 percent of the expected available dilution flow rate:
F = (0.9)(Fa) (20) Flow rate monitor setpoints for the Steam Generator Blowdown effluent stream s'1all be set at the selected discharge pump rate (normally the maximum discharge pump rate) f a, chosen in Step 3 above.
- 5) The Steam Generator Monitor setpoints may be specified based on the values of E C , F,and f which were specified to provide compliance with the limits of 10CFR 20, Appendix B, Table 11, Column 2. The monitor response is primarily to gamma radiation, therefore, the actual setpoint is based on E Cg. The monitor setpoint in cpm which corresponds to the calculated value c is taken from the monitor calibration graph. (See NOTE, page 2.0-14.) The setpoint concentra-tion, c,is determined as follows: g c s 1 c, X 11 (21) a B = fjf a, (22)
If B J 1 Calculate c and determine the max; mum value 3 for the actual monitor setpoint (cpm) from the monitor calibration graph. i 03CM, V. C. Summer, SCE &G: Revision 13 (June,1990) 2.0 22
l If B < 1, No release may be made Reevaluate the alternatives presented in step 3. NOTE: if the calculated setpoint value is near actual concentrations being released or planned for release, it may be impractical to set the monitor alarm at this value. In this case a new setpoint rnay be calculated following the remedial methodology presented in steps 3 and 4 for the case f, < fo ,.
- 1 l
- 6) The Turbine Building Sump and Condensate Demineralizer Ostkwash monitor setpoints are to be established independently of each other and without crediting dilution. They are to be based on the measured radio-nuclide concentrations of the effluent stream and are to ensure compliance with the limits of 10CFR 20, Appendix B, Table !!, Column 2 prior to discharge, For each effluent stream, a concentration factor CF must be calculated, measuring the nearness of approach of the undiluted waste stream to the specified limiting condition of the Maximum Permissible Concentration. That is, C
#"
- SF (23)
,yp C
Fr" r + SF (24)
.ype C'
CF g = 1'
- n + SF (25) where:
ODCM, V. C. Summer, SCE&G; Revisior,13 (June,1990) 2.0 23
e
\
,,, 9f,'7, T
= the sum of the ratios of the measured concentration of nuclide i to its limiting value MPC, for the Turbine
, 7 Building Sump effluent.
c' \ = the sum of the measured concentration of nuclide i(in D 7 UI'C, liquid only) to its limitmg value MPC, for the Condensate Demineralizer Backwash effluent. CF 7 = the concentration f actor for the Turbine Building Sump Effluent. CF o = the concentration factor for the Condensate Demin-eralizer Beckwash Effluent SF = the generic engineering safety factor used in Section 2.1.2, Step 2.
= 0.5 IfCF 1, calculate c and determine the actual monitor setpoint (cpm) from the calibration curve.
If CF > 1, no release may be made via this path, The release must either be delayed or diverted for additional processing. Because of spurious alarms, these remedial steps may be required if the monitor setpoints are only near the actual concentrations being released. Within the limits of the conditions stated above, the specific monitor setpoint concentrations for the tvec Steam Generator Blowdown monitors RM-L3 and RM L10 rid the setpoint concentrations for RM L8 and RM L11 may now be calculated. Because they are primarily sensitive to gamma O DCM, V. C. Surnmer, SCE & G : Revision 13 (June,1990) 2.0 24
radiation, their setpoints wik be based on the concen-trations of gamma emitting iadionuclides as follows: 2.1.4.1.1 For RM L3, Steam Generator Blowdown Dis-charge initial nionitor, and for RM L10, Steam Generator Blowdown Disch arge fir.al monitor-cs ycy s 1 C,, 3 iID (26) a .
= the isotopic concentration of the Steam Generator 1, C, s Blowdown effluent as obtained from the sum of the a measured concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/mt.
*See GENERAL NOTE under 2.1.
2.1.4,1.2 For RM-L8, Turbine Buildinq Sump Discharge Monitor: Where: e S r 1C e r ' CF r (27) The gamma isotopic concentration of the Turbine Building Y c,
- r= Sump effluent as obtained from the sum of the measured 8
concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/ml. CFT = The Turbine Building Sump Effluent Concentration Factor from equation (24).
'See GENERAL NOTE under 2.1.
2.1.4.1.3 For RM-L11, Condensate Derilineralizer Backwash Discharge Monitor: eg s 1 C,, ,, - CF i, (28) a O DCM, V. C. Su mmer, SCE &G Revision 13 (June,1990) 2.0-25
= The gamma isotopic concentration of the Condensate 1, C, o Demineralizer Backwash effluent (including solids) as ob-e tained from the sum of the measured concentrations determined by the analysis required ODCM Table 11-4,in uCi/ml.
CFo = The Condensate Demineralizer Backwash Ef fluent Concen-tration Factor from equation (25).
*See GENERAL. NOTE under 2.1. .
2 1.4.2 Steam Generator Blowdown Effluent Not Directly to Circu-lating Water ( Alternate Mode) Equation (15) is again used to assure that effluents are in compliance with the aforementioned specification before dilution in the receiving water: C Because dilution is not considered in the setpoint calculation, it is not necessary to calculate maximum permissible discharge flow rates or anticipated available dilution flow rate. The functions of the four monitors whose setpoints are to be established are described in Section 2.1.4.1 above. The method for the determination is as follows: ,
- 1) If a release is found to be permissible, flow rate mon. tors for the active effluent streams (Steam Generator Blowdown - f g, Turbine Building Sump - f,, and Condensate Demineralizer - f,) may have their setpoints established at any operationally convenient value. Since 10CFR 20 is to be complied with before dilution, the flow rate of discharges is irrelevant.
ODCM, V. C. Summer, SCE &G: Revision 13 (June,1990) 2.0-26
- 2) The Concentration Factor of equations (23) (25) is again used to ensure the permissibihty of the release:
C VY " SY Sq; y C CF r = - SF , 1, m,'c. i C' CF = k.. o
~ SF u Sgpc, c
CF.= 3 gl. , - SF (29) in which all terms are defined in subsection 1.1.3.1 and subscripts T, D, and 5 refer respectively to the Turbine Building Sump Effluent, the Condensate Demineralizer Backwash Effluent, and the Steam Generator Blowdown E f fluent. If CF i 1, calculate c and determine the actual monitor setpoint (cpm) from the calibration curve. If CF > 1, no release may be made via this path. The release must either be delayed or diverted for additional processing. Because of spuricus alarms, these remedici steps may be required if the monitor setpoints are only near the actual concentrations being released. Within the above limitation, setpoint concentrations may now be calculated for the four effluent monitors. Because they are primarily sensitive to gamma radiation, their setpoir.ts will be based on the concentrations of gamma emitting radionuclides as follows: ODCM, V. C. Summer, SCE &G: Revision 13 (June,1990) l l 2.0-27
2.1.4.2,1 For RM L8, Turbine Buildino Sump Discharge Monitor (usina equation (27) above): C7s [ C, 7 - cF,. where:
= The gamma isotopic concentration of the Turbine Building N, C Sump effluent as obtained from the sum of the measured 7 ' concentrations determined by the analysis required in ODCM Table 1.1-4,in uCi/ml.
C F, = The Turbine Building Sump Effluent Concentration Factor from equation (24). 1
*See GENER AL NOTE under 2.1. l l
I 2.1.4.2.2 For RM-L11, Condensate Demineralizer Backwash Discharoe Monitor (usina equation (28) above):
\ - CF g eg s _ C, y a
where: y = the gamma isotopic concentration of the Condensate Demin- _ C,. n eralizer Backwash effluent (including solids) as obtained from c the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4, in uCi/ml. CF o = The Condensate Oemineralizer Backwash Effluent Concen-tration Factor from equation (25).
*See GENERAL NOTE under 2.1.
ODCM,V.C. Summer SCE&G: Revision 13 (June,1990) 2.0 28
2.1.4.2.3 For RM L3, Steam Generator Blowdown Dis-charge initial monitor, and RML 10, Steam Generator Blowdown Discharae final monitor:
% ""'s8 5 C
- CFs e 3 (30) e where:
N' c = The isotopic concentration of the Steam Generator Blow-7"# down effluent as obtained from the sum of the measured concentrations determined by the analysis required in CDCM Table 1.14,in uCi/ml. CF 3 = The Steam Generator Blowdown Effit. nt Concentration ' Factor from equation (29).
'See GENERAL NOTE under 2.1.
I
. ?
ODCM, V. C. Summer, SCE &G: Revision 13 (June,1990) 2.0-29 g _ ._____w. - _ - - _ - _ - _ _ _ - - - - - -
Figure 2.1 1 Example Liquid Effluent Monitor Calibration Curve 10 -
.- - & +.,=.L= g .5 y_: .g3 5 : - .= g &s . .1_.L_'.___f3 _; u . 5 = h'**
':-.-- ....._~....,.....L,._... .-'_y2_ ........r'.L... ..L. ' .: ^
. LL. -
L a .: .' -~---L-
. IL.
6
-- _ = p _, =_ ,
=,e . = ge g. . . _ -
10 l., _e = .' _.. . _...._ . . . _ _ .. . . _ _ _ . mF
.. 6 s u: .
C m e==== ==J = = = 4AkN . i= Mag-o 'w.-.L- - ---=== = g = x c -_= . . . .
.. . . . . =
a
. g _ = q--. g=:=.=g u;= a.= -
_== n -- - --
, *l*
*0
{n..... .
....._4 ..
' E .;'-'
=a1==-~=," m === = ==~'~~~
c -- l , ' e '- - '-- -~~ d--rc--' i----
,~ : . . . 2 ' .. 2c . ' ._. r . .. u u . _4 . _ _ _ . :_._ _1 __ : _ _ a__. _.. s. i m - .mi ' * '- - - -- - .- ' N n , . . . .
, . .. c . . e., - . . .
+s==
t x.g_==m1w c==2g
-.2-____..__,._ g % = ===,e=2; s = = = ="___"
r o .
'" 1 _ -. ;.. .
-z
=
_ ~_ =
._i.._._....g _ ;g g : .7 g .5 g 4 e 1 _ J _ --
2--
- s a s .
., ,..i.. . . . .. c ,
t ."- - - - . 7 - : . - ~ A ...m =am -m -
.3 ..
-:',-.u_.....--.:.._:_.
~
..z..;..
a .:
.,t ---~
? . ;::'^-_
- - - - + - -
7 --.w:.:_. o 3 to ..
=gr 2 n 1(e 1
.---1 -. a . . . . _ _ .1____ = -
= = - - - - -
. . i.?
,C..< -
e .
. _ . . . . . ..y__ ....__---
< l
~- ~
(uCi/ml) - . _ . . . .
- . ...- - . A = * =5
- ,e --- . _ - - - - - - - _._...
10 4 ; . .-
. on .,. ,
-_. : g --- ~.= - == = _ _ =.
..y.. : .. ; 2. . _ . . . .- . : . . .x }
l
.g 2.
. .. .. .. ...a_.-.:.... : . .. .. . ..
._n___.__:--.
i.. 1-M _ a a.g i = xE P -~1=.= a _:_ M xxc_Es ==
....3 . .
j LO 10 10' 10 10
.... 3 10 1;,
i j Count Rate (cpm) i ODCM, V. C. Summer SCE&G, Revision 13 (June,1990) 2.0 30 i i
1 I 2.2_ Dose Calculation For Liauid Effluents ~ .. _ The method of.this section is to be used in all cases for calculating 1 doses to individuals from routine' liquid effluents. Four notes at the end of the section confirm the values which certain parameters are to be assigned in some special cases. 2.2.1 Liquid Effluent Dose Calculation Parameters S " Term : Definition Ini Use Ah = the site related ingestion dose commitment 2.2.2 factor to the total body or any organ t, for *
)
each identified principal gamma and beta l emitter listed in Table 2.2 3 in mrem mi per i
-hr pCi.
l B F, = Bioar.cumulation Factor for nuclide i, in fish, 2.2.2 pCi/Kg per pCi/l, from Table 2.21. C, = the average concentration of radionuclide, 2.2.2 i,in undiluted liquid effluent during time penod At, from any liquid released,in uCi/ml. D F" n a dose conversion factor for nuclide, i, for 2.2.2 adults in preselected organ, t,in mrem /pCi found in Table 2.2-2. D' = ~ the cumulative dose commitment to the 2.2.2 total body or any organ, t , from the liquid effluents for the total time period, Eatg in mrem (Ref.1). D* -
= Dilution Factor from the near field area - 2.2.2 ,
within one-quarter mile of the release points to the potable water intake for adult
. water consumption; for V. C. Summer, D, = 1.
Fc =. the near field average dilution factor for C, 2.2.2 during any liqui _d effluent release.
.K = 1.14 x 105, units conversic,n factor = 2.2.2 (106 pCi/uCi)(103 ml/l) + 8760 hr/yr
. ODCM, V. C. Summer, SCE &G : Revision 13 (June,1990) 2.0 31 i; ,, - - . . _ . , __..s. _ ., .., _, _ .# , m
I Liquid Effluent Dose Calculation Parameters (continuedj Term Definition S " ' ni Use atk = the length (in hours) of a time period over 2.2.2 which concentrations and flow rates are averaged for dose calculations. U, = 21 kg/yr, fish consumption (adult) 2 2.2 (Reference 3). U, = 730 kglyr, water consumption (adult) 2.2.2 (Reference 3). l l Z = applicable near field dilution factor when 2.2.2 no additional dilution is to be considered; Z = 1. 2.2.2 Methodology The dose contribution from all radionuclides identified in liquid effluents released to unrestricted areas is calculated using the following expression: D=Y .1 Y .it, C, F. (31)
. k-i A = K,( (U,/D,) + U,BF,) DF, (32)
FL= (averaae undiluted liquid waste flow) (33) (average flow from the discharge structure) (Z) NOTE 1: If radioactivity in the Monticello Reservoir (C,,) becomes > the LLD specified in ODCM, Table 1.1-4, that concentration must be included in the Dose determination. Fc,r this part of the dose calculation, F, = 1 and at, = the entire time period for which the dose is being calculated. NOTE 2: During periods when the Circulating Water Pumps are not in operation, the possibility of leakage of activity from the industrial Water System will be accounted for as follows. Sampling of the liquid in the Sanitary and Industrial Waste ODCM, V. C. Summer, SCE&G: Revision 13 (June,1990) 2.0 32
n . System will be initiated, and the measured concentrations of radionuclides will be used in the dose calculations with F, = 1 and a t, = the entire time period for which the doe being calculated. NOTE 3: During periods when the Circulating Water Pumps are in operation, any releases to the Sanitary and Industrial Waste System a_g to be credited with dilution in Circulating Water for dose calculation purposes, even though such dilution was not claimed in the setpoint calculation, When taken in union with the note above, this procedure results in some overestimation of dose to the population because discharges made to the Sanitary and industrial Waste System just before loss of Circulating Water will Le counted twice in the dose calculation process. NOTE 4: If radioactivity in the Service Water becomes > LLD as determined by the analysis required by ODCM, Table 1.1-4, that concentration must be included in the Dose determination. For this part of the dose calculation, F, = 1 and A t, = the entire time since the last Service Water sample was taken. ODCM, V. C. Summer, SCE &G: Revision 13 (June,1990) i 2.0-33
i TABLE 2.21 BIOACCUMULATION FACTOR 5' (pCi/kg per pCi/ liter) ELEME NT ' FRESHWATER FISH ,_ H 9.0E-01 < C 4.6E 03 l F 1.0F 01 Na 1.0E 02 P 1.0E 05 Cr 2.0F 02 Mn 4.0E 02 Fe- 1.0E 02 Co 5.0E 01 i I Ni 1.0E 02 i Cu 5.0E 01 Zn 2.0E 03 Br 4.2E 02 Rb 2.0E 03 Sr 3.0E 01 Y 2.5E 01 Zr 3.3E 00 Nb 3.0E 04 Mo 1.0E 01 Tc 1.5E 01 Ru 1.0E 01 Rh 1.0E 01 Sb 1.0E 00 Te 4.0E 02 1 1.5E 01 Cs 2.0E 03 Ba 4.0E 00 La 2.5E 01 Ce- 1.0E 00 Pr 2.5E 01 Nd 2.5E 01 W 1-2E 03 Np 1.0E 01
- Values in Table 2.2-1 are taken from Reference 3, Table A-1.
ODCM, V. C. Su mmer, SCE &G : Revision 13 (June,1990) 2.0-34
TABLE 2.2 2 l Page l of 2 ADULT INGESTION DOSE FACTOR 5* (mrem /pCiingested) NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 NO DATA 1.05 E-07 1.05E 07 1.05E-07 1.05E-07 1.05E 07 1.05E 07 C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 t F-18 6.24E-07 NO DATA 6.92 E -08 NO DATA NO DATA NO DATA 1.85E-08 l N A-24 1.70E 06 1.70E-06 1.70E-06 1.70 E-06 1.70E-06 1.07 E-06 1.70E-06 P-3 2 1.93 E-04 1.20E 05 7.46E-06 NO DATA NO DATA NO DATA 2.17 E-05 CR-51 NO DATA NO DATA 2.66E-09 1.59E 09 5.86E-10 3.53E-09 6.69E 07 MN-54 NO DATA 4.57E-06 8.72 E-07 NO DATA 1.36E-06 NO DATA 1.40 E-05 MN-56 NO DATA 1.15 E-07 2.04E-08 NO DATA 1.46E 07 NO DATA 3.67E-06 FE-55 2.75 E-06 1.90 E-06 4.43E-07 NO DATA NO DATA 1.06E-06 1.09 E-06 FE 59 4.34E-06 1.02 E-05 3.91 E-06 NO DATA NO DATA 2.85E-06 3.40E-05 tCO 57 NO DATA 1.75E-07 2.91 E-07 NO D ATA NO DATA NO DATA 4.44E-06 l CO-58 NO DATA 7.45E-07 1.67E-06 NO DATA NO D ATA NO DAT A 1.51 E-05 CC-60 NO DATA 2.14 E-06 4.72 E -06 NO DATA NO DATA NO DATA 4.02 E-05 NI63 1.30E 04 9.01 E-06 4.36E-06 NO DATA NO DATA NO DATA 1.88E-06 Ni65 5.28E-07 6.86E 08 3.13 E-08 NO DATA NO DATA NO DATA 1.74E-06 CU 64 NO DATA 8.3 3 E-08 3.91 E-08 NO DATA 2.10 E -0 NO DATA 7.10 E-06 ZN-65 4.84E-06 1.54E-05 6.76E-06 NO DATA 1.03 E-05 NO DATA 9.70E 06 ZN-69 1.03 E-08 1.97E-08 1.37 E-09 NO DATA 1.28E-08 NO DATA 2.96E-09 tBP-82 NO DATA NO DATA 2.26E-06 NO DATA NO DATA NO DATA 2.59E 06 l BR-83 NO DATA NO DATA 4.02 E-08 NO DATA NO DATA NO DATA 5.79 E-08 BR-84 NO DATA NO DATA 5.21 E-08 NO DATA NO DATA NO D ATA 4.09E-13 3R-85 NO DATA NO DATA 2.14E-09 NO DATA NO DATA NO DATA LT E 24 *
- RB-86 NO DATA 2.11 E-05 9.83E-06 NO DATA NO DATA NO DATA 4.16E-06 RB-88 NO DATA 6.05E-08 3 21E-08 NO DATA NO DATA NO DATA 8.36E-19 RB-89 NO DATA 4.01 E-08 2.82 E-08 NO DATA NO DATA NO DATA 2.33E-21 SR-89 3.08E 04 NO DATA 8.84E-06 NO DATA NO DATA NO DATA 4.94E 05 SR-90 7.58E-03 NO DATA 1.86E-03 NO DATA NO DATA' NO DATA 2.19E 04 SR 91 5.67E-06 NO DATA 2.29E-07 NO DATA NO DATA NO DATA 2.70E-05 SR-92 2.15E-06 NO DATA 9.30E-08 NO D ATA NO DATA NO DATA 4.26E-05 Y-90 9.62E-09 NO DATA 2.58E-10 NO DATA NO DATA NO D ATA 1.02 E-04 Y-91 M 9.09E-11 NO DATA 3.52E-12 NO DATA NO DATA NO DATA 2.67E-10 Y-91 1.41 E-07 NO DATA 3.77E 09 NO DATA NO DATA NO DATA 7.76E-05 Y-92 8.45E-10 NO DATA 2.47 E- 11 NO DATA NO DATA NO DATA 1.48E-05 Y 93 2.68E-09 NO DATA 7.40E-11 NO DATA NO DATA NO DATA 8.50E-05 ZR 95 3.04E-08 9.75E-09 6.60E 09 NO DATA 1.53E-08 NO DATA 3.09E-05 ZR-97 1.68E-09 3.39E 10 1.55E-10 NO DATA 5.12 E-10 NO DATA 1.05E-04 NB-95 6.22 E-09 3.46E-09 1.86 E-09 NO DATA 3.42E-09 NO DAT A 2.10 E-05 MO-99 , NO DATA 4.31 E-06 8.20E-07 NO DATA 9.76E-06 NO DATA 9.99E-06 tValues'aken from Reference 13, Table 4. l
- Values in Table 2.2-2 are taket from Reference 3, Table E-11
*
- Less than E-24.
ODCM, V.C. Summer, SCE& G. Revision 15 (February 1991) l 2.0-35
j TABLE 2 2-2 (contmued) Page 2 of 2 NUCLIDE BONE LIVER T.B O DY THYROID KIDNEY LUNG GI LLI TC-99M 2.47 E - 10 6.98 E- 10 8.89E 09 NO DATA 1.06E 08 3.42 E- 10 4.13 E -07 TC 101 2.54 E- 10 3.66E-10 3.59E 09 NO DATA 6.59E-09 1.87 E-10 1.10 E-21 R U- 103 1.85E-07 NO DATA 7.97E-08 NO DATA 7.06E 07 NO DATA 2.16E 05 RU-105 1.54 E-08 NO DATA 6.08E-07 NO DATA 1.99E 07 NO DATA 9.42E 06 RU-106 2.75E 06 NO DATA 3.48E-07 NO DATA 5.31 E -06 NO DATA 1.78E-04 AG-110M 1.60E-07 1.48E-07 8.79E 08 NO DATA 2.91 E-07 NO DATA 6.04 E-0 5 tSB-124 2.80E-06 5.29E-08 1.11 E-06 6.79E-09 NO DATA 2.18E-06 7.95E-05 158 125 1.79E-06 2.00E-08 4.26E 07 1.82 E-09 NO DATA 1.38E-06 197E-05 tW 46 1.15 E-06 2.34E-08 4.15 E -07 7.04E 09 NO DATA 7.05E-07 9.40 E -05 T c-12 5M 2.68E-06 9.71 E-07 3.59E-07 8.06E 07 1,09E 05 NO DATA 1.07 E-05 TE-127M 6.77E-06 2.42 E-06 8.2 5 E-07 1.73 E-06 2.75E-05 NO DATA 2.27E-05 TE 127 1.10 E-07 3.95E-08 2.38E-08 8.15E 08 4.48E 07 NO DATA 8.68E-06 TE-129M 1.15E 05 4.29E-06 1.82 E -06 3.95E 06 4.80E-05 NO DATA 5.79E-05 TE-129 3.14 E -08 1.18E-08 7.65E-09 2.41 E-OS 1.3 2 E -07 NO DATA 2.37E-08 TE-131M 1.73 E-06 8.46E-07 7.05 E -07 1.34 E -06 8.57E 06 NO DATA 8.40 E -0 5 TE-131 1.97E 08 8.2 3 E-09 6.2 2 E -09 1.62 E-08 8.63 E-08 NO DATA 2.79E-09 TE-132 2.52 E-06 1.63 E-06 1.53 E-06 1.80E-06 1.57E-05 NO DATA 7.71 E-05 l l-130 7.56E-06 2.2 3 E-06 8.80E-07 1.89 E-04 3 48E-06 NO DATA. 1.92 E -06 l-131 4.16E-06 5.95 E-06 3.41 E-06 1.95 E-03 1.02 E-05 NO DATA 1.57 E-06 l-132 2.03 E-07 5.43E 07 1.90E 07 1.90E-05 8.65E-07 NO DATA 1.02E 07 l-133 1.42 E-06 2.47E-06 7.53E-07 3.63E-04 4.31 E-06 NO DATA 2.22 E-06 l-134 1.06E-07 2.88E-07 1.03E 07 4.99E-06 4 58E-07 NO DATA 2.51 E - 10 1-135 4.43 E-07 1.16E 06 4.28E-07 7.65E 05 1.86E-06 NO DATA 1.31 E -06 CS-134 6.22 E-05 1.48E-04 1.21 E -04 NO DATA 4.79E-05 1.59E-05 2.59E-06 CS-136 6. 5 '. E -06 2.57E-05 1.85E-05 NO DATA 1.43 E-05 1.96E-06 2.92E-06 C5-137 7.97E-05 1.09E-04 7.14 E -05 NO DATA 3.70E-05 1.2 3 E -05 2.11 E -06 CS-138 5.52E-08 1.09E-07 5.40E-08 NO DATA 8.01 E-08 7.91 E-09 4.65E-13 B A-139 9.70E-08 6.91 E-11 2.84E-09 NO DATA 6.46E-11 3,92 E- 11 1.72 E -07 B A-140 2.03E-05 2.55E-08 1.33 E-06 NO DATA 8.67E-09 1.46E-08 4.18 E-0 5 B A-141 4.71 E-08 3.56E-11 1.59E-09 NO DATA 3.31 E- 11 2.02 E-11 2.22 E -17 B A-142 2.13 E-08 2.19 E- 11 1.34E 09 NO DATA 1.85E-11 1.24 E- 11 3.00E-26 LA-140 2.50E-09 1.26E-09 3.33 E-10 NO DATA NO DATA NO DATA 9.2 5 E-05 LA-142 1.28E-10 5.82E-11 1.45 E- 11 NO DATA NO DATA NO DATA 4.25E-07 C E-141 9.36E-09 6.33 E-09 7.18E-10 NO DATA 2.94E-09 NO DATA 2.42E 05 CE-143 1.65E-09 1.22 E-06 1.35 E-10 NO DATA 5.37 E- 10 NO DATA 4.56E-05 C E-144 4.88E-07 2.04E-07 2.62E-08 NO DATA 1.21 E-07 NO DATA 1.65E-04 PR-143 9.20E-09 3.69E-09 4.56E-10 NO DATA 2.13E-09 NO DATA 4.03 E-05 PR-144 3.01 E- 11 1.2 5 E- 11 1.53 E-12 NO DAYA 7.05E-12 NO DATA 4.33 E- 18 ND-147 6.29E 09 7.27E-09 4.3 5 E- 10 NO D ATA 4.2 5 E-09 NO DATA 3.49E-05 W-187 1.03 E-07 8.61 E-08 3.01 E-08 NO DAT A NO DATA NO DATA 2.82 E-0 5 j NP-239 1.19E-09 1.17 E- 10 6.45E-11 NO DATA 3.65E-10 NO DATA 2.40E-05 ODCM, V.C. Su mmer, SCE &G Rewsion 15 (February 1991) 2.0 36
TABLE 2.2-3 SITE RELATED INGESTION DOSE COMMITMENT FACTOR, Au* (mrem /hr perpCi/ml) Page 1 of 2 NUCLIDE BONE LIVER T.B O DY THYROID KIDNEY LUNG GI-LLI H-3 NO DATA 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 C 14 3.15E + 04 6.30E + 03 6.30E + 03 6.30E + 03 6.30E + 03 6.30E + 03 6.30E + 03 F-18 6.69E + 01 NO DATA 7.42E + 00 NO DATA NO DATA NO DATA 1.96E + 00 NA-24 5.48E + O2 5.48E + 02 5.48E + 02 5.48E + 02 5.48E + O2 5.48E + 02 5.48E + 02 P-32 4.62E + 07 2.87E + 06 1.79E + 06 NO DATA NO DATA NO DATA 5.20E + 06 CR-51 NO DATA NO DATA 1.49E + 00 8.94E-01 3.29E-01 1.98E + 00 3 76E + O2 l MN 54 NO DATA 4.76E7 03 9.08E + 02 NO DATA 1.42E + 03 NO DATA 1.46E + 04 MN 56 NO DATA 1.20E + 02 2.12L + 01 NO DATA 1.52E + 02 NO DATA 3.82E + 03 FE-55 8.87E + 02 6.13E + 02 1.43r + 02
~
NO DATA NO DATA 3.42E + 02 3.52E + 02 i FE 59 1.40E + 03 3 }9E + 03 1.26E + 03 NO DATA NO DATA 9.19E + 02 1.10E + 04 CO-57 NO DATA 3.55E + 01 5.910 + 9 NO DATA NO DATA NO DATA 9.01 E + 02 l CO-58 NO DATA 1.51 E + 02 3.39E + O2 NO DATA NO DATA NO DATA 3.06E + 03 i CO-60 NO DATA 4.34E + 02 9.58E v 02 NO DATA NO DATA NO DATA 8.16E + 03 NI63 4.19E + 04 2.91E + 03 1.41E + 03 NO DATA NO DATA NO DATA 6.07E + O2 l NI-65 1.70E + 02 2.21E + 01 1.01E + 01 NO DATA NO DATA NO DATA 5.61 E + 02 CU;64 NO DA A 169E + 01 7.93E + 00 NO DATA 4.26E + 01 NO DATA 1.44E + 03 ZN 65 2.36E + 04 7.50E + 04 3.39E + 04 NO DATA 5.02E + 04 NO DATA 4.73E + 04 ZN-69 5.02E + 01 9.60E + 01 6.67E + 00 NO DATA 6.24E + 01 NO DATA 1.44E + 01 BR-82 NO DATA NO DATA 2.46E + 03 NO DATA NO DATA NO DATA 2.82E + 03 l BR-83 NO DATA NO DATA 4.38E + 01 NO DATA NO DATA NO DATA 6.30E + 01 B R-84 NO DATA NO DATA 5.67E + 01 NO DATA NO DATA NO DATA 4.45E 04 BR-85 NO DATA NO DATA 2.33E + 00 NO DATA NO DATA NO DATA 1.09E - 15 RB-86 NO DATA 1.03E + 05 4.79E + 04 NO DATA NO DATA NO DATA 2.03E + 04 RB-88 NO DATA 2.95E + 02 1.56E + 02 NO DATA NO DATA NO DATA 4.07 E - 09 RB-89 NO DATA 1.95E + 02 1.37E + O2 NO DATA NO DATA NO DATA 1.13E - 11 SR-89 4.78E + 04 NO DATA 1.37E + 03 NO DATA NO DATA NO DATA 7.66E + 03 SR-90 1.18E + 06 NO DATA 2.88E + 05 NO DATA NO DATA NO DATA 3.48E + 04 SR-91 8.79E + O2 NO DATA 3.55E + 01 NO DATA NO DATA NO DATA 4.19E + 03 SR-92 3.33E + 02 NO DATA 1.44E + 01 NO DATA NO DATA NO DATA 6.60E + 03 Y-90 1.38E + 00 NO DATA 3.69E - 02 NO DATA NO DATA NO DATA 1.46E + 04 Y-91 M 1.30E - 02 NO DATA 5.04E - 04 NO DATA NO DATA NO DATA 3.82 E - 02 Y-91 2.02E + 01 NO DATA 5.39E - 01 NO DATA NO DATA NO DATA 1.11 E + 04 Y-92 1.21 E - 01 NO DATA 3.53 E - 03 NO DATA NO DATA NO DATA 2.12E + 03 Y-93 3.83 E - 01 NO DATA 1.06E - 02 NO DATA NO DATA NO DATA 1.22E + 04 ZR-95 2.77E + 00 8.88E - 01 6.01 E - 01 NO DATA 1.39E + 00 NO DATA 2.82E + 03 ZR 97 1.53E - 01 3.09E - 02 1.41 E - 02 NC DATA 4.67E - 02 NO DATA 9.57E + 13 N B-95 4.47E + 02 2.49E + 02 1.34E + 02 NO DATA 2.46E + 02 NO DATA 1.51 E + 06
- Calculated using equation (32) and Tables 2.2-1 and 2.2-2.
ODCM, V.C. Summer, SCE&G: Revision 15 (February 1991) 2.0-37
TABLE 2.2-3 , SITE RELATED INGESTION DOSE COMMITMENT FACTOR, Au c (mrem /hr perpCi/ml) Page 2 of 2 NUCLlDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI LLI MO 99 NO DATA 4.62E + O2 8.79E + 01 NO DATA 1.05E + 03 NO DAT A 1.07E + 03 TC-99M 2.94E - 02 8.32E - 02 1.06E + 00 NO DATA 1.26E + 00 4.07E - 02 4.92E + 01 TC-101 3.03E - 02 4.36E - 02 4.28E - 01 NO DATA 7.85E - 01 2.23E-02 1.31E - 13 RU 103 1.98E + 01 NO DATA 8.54E - 01 NO DATA 7.57E + 01 NO DATA 2.31 E + 03 RU-105 1.65E + 00 NO DATA 6.52 E - 01 NO DATA 2.13E + 01 NO DATA 1.01 E + 03 RU-106 2.95E + 02 NO DATA 3.73E + 01 NO DAT A 5.69E + 02 NO DATA 1.91E + 04 AG 110M 1.42 E + 01 1.31 E (01 7 TOT 4 00 NO DATA 2.58E + 01 NO DATA 576E + 03 58-124 2.40E + O2 4.53E + 00 9.50E + 01 5.81 E-01 NO DATA 1.87E + 02 6.81E + 03 58-125 1.53E + 02 1.71E + 00 3.Mi + 01 1.56E 01 NO DATA 1.18E + O2 1.69E + 03 58 126 9.85E + 01 2.00E + 00 3.55E + 01 6.03E 01 NO DATA 6.04E + 01 8.05E + 03 l TE-125M 2.79E + 03 1.01 E + 03 374E + 02 8.39E + 02 1.13E + 04 NO DATA 1.11 E + 04 TE 127M 7.05E + 03 2.52E + 03 8.59E + 02 1.80E + 03 2.86E + 04 NO DATA 2.36E + 04 TE-127 1.14E + 02 4.11 E + 01 2.48E + 01 8.48E + 01 4.66E + O2 NO D ATA 9.03E + 03 TL-129M 1.20E + 04 4.47E + 03 1866 + 03 4.11E + 03 5.00E + 04 NO DATA 6.03E + 04 TE-129 3.27E + 01 1.2 3 E + 01 7.96E + 00 2.51E + 01 1.37E + 02 NO DATA 2.47E + 01 TE-131M 1.88E + 03 8.81 E + 02 7 34E + 02 1.39E + 01 8.92E + 03 NO DATA 8.74E + 04
'TE 131 2.05E + 01 8.57E + 00 6.476~ + 00 1.69E + 01 8.98E + 01 NO DATA 2.90E 4 00 TE-132 2.62E + 03 1.70E + 03 1.59E + 03 1.87E + 03 1.63E + 04 NO DATA 8.02E + 04 l-130 9.01E + 01 2.66E + 02 1.05E + 02 2.25E + 04 4.15E + 02 NO DATA 2 29E + 02 1-131 4.96E + 02 7.09E + 02 4.06E + 02 2.32E + 05 1.22E + 03 NO DATA 1TfE+O2 1-132 2.42E + 01 6.47E + 01 2.26E + 01 2.26E + 03 1.03E + O2 NO DATA 1.22E + 01 1-133 1.69E + O2 2.94E + 02 8 97E + 01 4 32E + 04 5.13E + 02 NO DATA 2 i4E + 02 1-134 1.26E + 01 3.43E + 01 1.2 3 E + 01 5.94E + 02 5.46E + 01 NO DATA N 9E-02 1-135 5.28E + 01 1.38E + 02 5.10E + 01 9 *1E 4 03 2.22E + O2 NO DATA 1.56E + 02 C5-134 3.03E + 05 7.21E + 05 5.80E + 05 NO DATA 2.33E + 05 7.75E + 04 1.26E + 04 C5-136 3.17E + 04 1.25E + 05 9.0 i s + 04 NO DATA 6.97E + 04 9.55E + 03 1.42E + 04 C5-137 3.88E + 05 5.31E + 05 3.48E + 05 NO DATA 1.88E + 05 5.99E + 04 1.03E t 04 C5-138 2.69E + 02 5.31E + 02 2.63E + 02 NO DATA 3.90E + O2 3.85E + 01 2.27E - 03 8 A-139 9.00E + 00 6.41 E - 03 2.64E - 01 NO DATA 5.99E-03 3.64E - 03 1,60E + 01 B A-140 1.88E + 03 2.37E + 00 1.23E + 02 NO DATA 8.05E - 01 1.35E + 00 3.88E + 03 B A-141 4.27E + 00 3.30E - 03 1.48E - 01 NO DATA 3.07E - 03 1.87E - 03 2.06E - 09 B A-142 1.98E + 00 2.03 E - 03 1.24E - 01 NO DATA 1.72E - 03 1.15E - 03 2.78E - 18 LA-140 3.58E - 01 1.80E - 01 4.76E - 02 NO D ATA NO DATA NO DATA 1.32E + 04 LA-142 1.83 E - 02 8.3 3 E - 03 2.07E - 03 NO DATA NO DATA NO DATA 6.08E + 01 CE-141 8.01 E - 01 5.42E 01 6.15E - 02 NO DATA 2.52 E - 01 NO DATA 2.07E + 03 CE 143 1.41 E - 01 1.04E + 02 1.16E - 02 NO DATA 4.60E - 02 NO DATA 3.90E + 03 CE-144 4.18E + 01 1.77E + 01 2.24E + 00 NO DATA 1.04E , 01 NO DATA 1.41E + 04 PR-143 1.32E + 00 5.28E - 01 6.52E - 02 NO DATA 3.05E - 01 NO DATA 5.77E + 03 PR-144 4.31E 03 1.79E - 03 2.19E - 04 NO DATA 1.01 E - 03 NO DATA 6.19E - 10 ND-147 9.00E - 01 1.04E + 00 6.22E - 02 NO DATA 6.08E - 01 NO DATA 4.99E + 03 W-187 3.04E + 02 2.55E + 02 8.90E + 01 NO DATA NO DATA NO DATA 8.34E + 04 NP-239 1.28E - 01 1.2 5E - 02 6.91 E - 03 NO DATA 3.91E 02 NO DATA 2.57E + 03 ODCM, V.C. Summer, SCE & G: Revision 15 (February 1991) 2.0-38
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30 G ASEOUS EFFLUENT 31 Gaseous Effluent Monitor Setpoints l The calculated setpoint values will be regarded as upper bounds for the actual setpoint adjustments. That is, setpoint adjustments are not required to be performed if the existing setpoint level corresponds to a lower count rate than the calculated value. Setpoints may be established at values lower than the calculated values,if desired. Calculated monitor setpoints may be added to the ambient background count rate. 3.1 1 Gaseous Effluent Monitor Setpoint Calculation Parameters Section of Term Definition Initial Use C '
= count rate of a station vent monitor (312) conesponding to grab sample radio-nwlide concentrations, X v,i as determined from the monitor's calibration curve, in cpm.
C'=' the count rate of the monitor on vent v (31.4) corresponding to X,' uCi/cc of Xe-133,in cpm. c = count rate of the gas decay system (3.1.3) monitor for measured radionuclide concentrations con ected to discharge pressure, in cpm. c' = the count rate of the waste gas decay (3.1.4) system monitor corresponding to the total noble gas concentration in cprn. D s3 = limiting dose rate to the skin (3000 (3.1.2) mrem / year). Du = limiting dose rate to the total body (3.1.2) (500 mrem / year). F.' = the flow rate in vent v (cusec) (3.1.2) (1 cc/sec = 0_002119 cfm). f, = the maximum permissible waste gas (3.1.3) discharge rate, based on the actual radionuclide mix and skin dose rate (cusec). ODCM, V. C. Summer, SCE &G Revision 13 (June 1990) 3.0 1
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1 9 4 Section of
' Term Definition initial Use fi = the maximum permissible waste gas discharge rate, based on the actual (3.13) radionuclide mix and total body dose rate (cusec).
f^ = the maximum permissible waste gas (3.1.3) discharge rate, the lesser of f,and f,(cusec). f,' = the conservative maximum per- , (3.1.4) missible waste gas discharge rate based on Kr-89 skin dose rate (cusec). , t' i
= the conservative maximum permissible (3.1.4) waste gas discharge rate based on Kr 89 total body dose rate (cusec).
K, = total body dose f actor due to gamma (3.1.2) emissions from isotope (mrem / year per uCi/m3) from Table 3.1 1. K,,.g = total body dose factor for Kr-89, the most (3.1.3) restrictive isotope from Table 3.1 1 (mrem /yr per uCi/m ). 3 L' .= Skin dose factor due to beta emissions (3.1.2) from isotope i(mrem /yr per uCi/m 3) from Table 3.1-1. L, , , = Ain dose factor for Kr-89, the most restrictive (3.1.3) ise: ope, from ?able 3.1-1 (mrem /yr per uCi/m ), 3
.M ' = air dose factor due to gamma emissions- (312) from isotope i(mradlyr per uCi/m3 ) from Table 3.1 1.
M ,,.g = air dose factor for Kr-89, the most restrictive (3.1.3) isotope, from Table 3.1 1 (mradlyr per uCi/m ). 3 R, = - count rate per mrem /yr tn the skin. (3.1.2) R, = count rate per mrem /yr to the total (3.1.2) body. R ,' = conservative count rate per mrem to (3.1.4) the skin (Xe 133 detection, Kr 89 dose). R ,' = conservative count rate per mrem to (3.1.4) the total body (Xe 133 detection, Kr-89 dose). ODCM, V. C. Su mmer, SC E & G : Revision 13 (June 1990) A.0 2
Section of Term Definition initial Use 5, = count rate of the waste gas decay (3.1.3) system noble gas monitor at the alarm setpoint, ir, cpm. 5, = count rate of a station vent noble gas (3.12) monitor at the alarm setpoint,in cpm. 5, , count rate of the containment purge (3.12) noble gas monitor at the alarm setpoint,in cpm. 5,, = count rate of the plant vent noble gas (31I) monitor at the alarm setpoint, in cpm. X,, = the concentration of noble gas radio- (3.13) nuclide iin a waste gas decay tank, as corrected to the pressure of the dis-charge stream at the point of its flow measurement in uCi/cc. X~ = the measured concentration of noble (3.1.2) gas radionuchde i in the last grab sample analyzed for vent v in uCi/cc. X ,' = the total noble gas concentration in a waste (3.1.4) gas decay tank, as corrected to the pressure of the discharge stream at the point of its flow measurement in uCi/cc. X' *
= a concentration of Xe-133 chosen to be in the (3,14) operating range of the monitor on vent v in uCi/cc.
X/O = the highest annual average relative concentra- (3.1.2) tion in any sector, at the site bounoary in sec/m3 1.1 = mrem skin dose per marad aia dose (3.1.2) 0.25 = the safety factor applied to each of the two (3.1.2) vent noble gas monitors (plant vent and contain-ment purge) to assure that the sum of the releases has a combined safety factor of 0Jwhich allows a 100 percent margin for cumulative uncertainties of measurements. ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0 3
1 l i i TABLE 3.1-1 1 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES
- Nuchde y-Body * * * (K i } B Skin * * *(L,1 y- Air * * (M,} ll Air * * (d,}
Kr-85m 1.17 E + 03 * * *
- 1.46E + 03 1.23E + 03 1.97E + 03 Kr-85 1.61 E + 01 1.34E + 03 1.72E + 01 1.95E + 03 Kr-87 5.92E + 03 9.73E + 03 617E + 03 103E + 04 Kr-88 1.47E + 04 2.37E + 03 1.52E + 04 2 93E + 03 Kr 89 1.66E + 04 1.01E + 04 1.73E + 04 106E + 04 Kr-90 1.56E + 04 7 29E + 03 1.63E + 04 7 83E + 03 Xe 131m 9.15E + 01 4.76E + 02 1.56E + 02 1.11 E + 03 Xe-133m 2.51E + 02 9:94E + 02 3.27E + 02 148E + 03 Xe-133 2.94E + 02 3.06E + O2 3.53E + 02 105E + 03 Xe 135m 3.12E + 03 7.11 E + 02 3.36E + 03 7.39E + 02 Xe 135 1.81 E + 03 1.86E + 03 1.92E + 03 2.46E + 03 Xe 137 1.42E + 03 1.22E + 04 151E + 03 1^27E + 04 Xe-138 8.8?5 + 03 4.13E 4 03 9.21E + 03 4.75E + 03 Ar-41 8.84E + 03 2.69E + 03 9.30E + 03 3.28E + 03
' Values taken from Reference 3, Table B 1
*
- mrad-m3 pCi yr
* *
- mrem-m3 pCi-yr
* * *
- 1.17E + 03 = 1.17 x 103 l ODCM, V. C Summer, SCE&G: Revision 13 (June 1990) l l 3.0-4 l
- , -~ _ . - . _ . .-- ..
l col + 3.12 Station Vent Noble Gas Monitors (RM A3 and RM A4) For the purpose of implementation of section 1.2.1 of the ODCM, the alarm setpoint level for the station vent noble gas monitors will be calculated as follows: 5, = count rate c,f the plant vent noble gas monitor ( = 5,y for RM-A3) or the containment purge noble gas monitor ( = S y for RM-A4) at the alarm setpoint level. 0.2 5 x R, x D1 , (34)* 6 the lesser of or 0.25 x R, x D u (35) t
- 0.25 = the safety factor applied to each of the two vent noble l
gas monitors (plant vent and containment purge) to assure that the sum of the releases has a combined safety factor of 0_5 which allows a 100 percent margin for cumulative uncertainties of measurements. D rs = Dose rate kmit to the total body of an individual
- = 500 mremlyr l
j R, = count rate per mrem /yr to the total body t
= C, /((X/Q) x F, x K K,X,,) (36) l Dss = Dose rate limit to the skin of the body of an individual in an unrestricted area.
! = 3000 mrem / year. Rs = count rate per mrem /yr to the skin.
= C, + R x F, x y (L, + 1.1 M.) X.,1 (37)
I X, =- the measured concentration of noble gas radionuclide i in the last grab sample analyzed for vent v, pCi/ml. (For the plant vent, grab samples are taken at least O DCM, V. C. Su mmer, SCE & G : Revision 13 (June 1990) 3.0 5 i I
_ _ _ _ _ ___. _ _. _ _ _ . _ . = _ . . _ _ . _ _ . . _ _ __ _ _ _ _ _ . -_ _ . . _ _~ _ . monthly. For the 6" and 36" containment purge iines, the sample is taken Just prior to the release and also monthly,if the release is continuous ) F, = the flow rate in vent v, cc/sec. (1 ccssec = 0.002119 cf m) C, = count rate, in cpm of the monitor on station vent v corresponding to grab sample noble gas concentra-tions, X ,, as determined from the monitor's calfbration curve. (Initial calibration curves of the type shown in Figure 2.1-1 have been determined conservatively from families of response curves supplied by the monitor manufacturers. As releases occur, a historical correla-tion will be prepared and placed in service when sufficient data are accumulated.) X/Q = the highest annual average relative concentration in any sector, at the site boundary. ! 1
= 5.3 x 10 6 sec/m3 in the SE sector
- K, = total body dose factor due to gamma emissions from isotope i(mrem /yr per pCum ) from Table 31-1. 3 L, = skin dose factor due to beta emissions from isotope i (mrem /yr per pCi/m3 ) from Table 3.1-1.
1.1 = mrem skin dose per mrad air dose M, = air dose factor due to gamma emissions from isotope (mradlyr per pCi/m ) from Table 3.1-1. 3 Reference 4, Section 11.3.8 states that this is the annual average relative dispersion at the point on the exclusion boundary where highest concentra-tions may be expected. ODCM, V. C. Summer, SCE &G : Revision 13 (June 1990) 3.0 6
- . - - . - . - _ _ _ . . . . . . . - - - - - - ~ ~ . . . . _ - - -
1: NOTE At plant startups when no grab sample analysis is available for the continuous releases, the Alternate Methodology of Section 31.4 must be used. 313 Waste Ga> Decay Ssstem Monitor (RM A10) The permissible conditions for discharge through the waste gas decay system monitor (RM- A10) will be calculated in a manner similar to that for the plant vent noble gas monitor in the case of the waste gas system, howe'ver, the discharge flow rate is continuously controllable by valve HCV-014 and permissible release conditions are therefore defined in terms of both flow rate and concentration. Therefore, RM A10 is used only to insure that a repre-sentative sample was obtained For operational convenience, (to prevent spurious alarms due to fluctuations in backotound) the setpoint level wil1 be estab;ished at 15 ti:nes the measured waste concentration. The maximum permissible flow rate will be set on the seme basis but include the engineering safety factor of 0.5. The RM A10 setpoint level 5, is ! defined as: 5, < 1.5c (38) l l- where: l c = count rate in CPM of the waste gas decay system monitor 1 i corresponding to the measured concentration (taken from. the monitor calibration curves). l l
-The maximum permissible waste gas flow rate f, (cc/sec) is calculated from the maximum permissible dose rates at the site boundary according to:
f, L the lesser of f, or f, (39) ODCM, V. C. Summer, SCE &G: Revision 13 (June 1990) 3.0 7
f, = the maximum permissible discharge rate based on total body dose rate.
= 0 25 x 0 3/ [XTO x 1 5 E X , K ) (40) f, = the maximum permissible discharge rate based on skin dose rate.
= 0.25 x Dss / [f0f) x 1.5 E X , (L + 1.1 M,)] (41)
X,d = the concentration of noble gas radionuchde in the waste gas decay tank whose contents are to be discharged, as co'rrected to the pressure of the discharge stream at the point of the flow rate measurement The maximum discharge pressure as governed by the diaphragm valve,7896,is 30 psia. NOTE: The factor of 15 in the denominators of equations (40) and (41) places f, on the same basis as 5, When a discharge is to be conducted, valve HCV 014 is to be opened until (a) the waste gas discharge flow rate reaches 0.9 x f, or (b) the count rate of the plant vent noble gas monitor RM-A3 approaches its setpoint, whichever c of the above conditions is reached first. When no discharges are being made from the Waste Gas Decay System, the RM-A10 setpoint should be established as near background as practical to prevent spurious alarms and yet alarm in the event of an inadvertent release. 3.1.4 Alternative Methodoloav for Establishino Conservative Setpoints A more conservative setpoint may be calculated to minimize requirements for adjustment of the monitor as follows: For a plant vent: R ,' = conservative count rate per mrem /yr to the total body (Xe-133 detection, Kr-89 dose). ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-8
= C,' + [X7Q x K , g xX,'xF,), (42) where:
X ,' = a concentration of Xe 133 chosen to be in the operating range of the monitor on vent v, pCi/cc. C,' = the count rate in CPM of the monitor on vent v corresponding to X,' pCi/cc of Xe-133. K,, y = total body dose factor for Kr-89, the most restrictive isotope from Table 3.1-1. R,' = count rate per mrem /yr to the skin.
= C,' + [EQ x (L,, g + 1.1 M,, g) x X,'x F,) (43) -
where: _ L,,, = skin dose factor for Kr-89, the most restrictive isotope from Table 3.1-1. M , , . ,, = air dose factor for Kr-89, the most restrictive isotope, from Table 3.1-1. For the waste gas decay system: f' = the conservative maximum permissible discharge rate based on Kr 89 total body dose rate.
= 0.2 5 x Da + [X/Q x 1.5 x X,'x K,,.gl (44) f,' = the conservative maximum permissible discharge rate based on Kr 89 skin dose rate.
= 0.25 x D 33 + @ x 1,5 x X o'x R,, ,, + 1 A M,,.,,H W ODCM, V. C. Summer, SCE &G: Revision 13 (June 1990) 3.0 9
~
X,' = the total concentration of noble gas radionuclides in the waste gas decay tank whose contents are to be discharged, as corrected to the pressure of the discharge stream at the point of the flow measurement. c' = count rate in cpm of the waste gas decay system monitor corresponding to X,' pCi/cc of Kr 85. i l 7 l t l ODCM, V. C. Summer, SCE &G: Revision 13 (June 1990) l 1 3.0 10
1 l l i Figure 3.1-1 Example Noble Gas Monitor Cahbration Curve : 1 1 E.J .. -. - .O:? g
..? .
, 5554diL5c .x2 E . _. " , p .b;( ,_ 3._.. _.,..c.,..... 2 . ..; :._ _ .a. g ; " .. . g ici-$
' , . . -O ..
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..r..-&#= ;=
~-
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- z. ; - . . . .
- .... q t:i.
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, or .
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. , . *,-> q -,- y .t_ ,3;.a ._ , . _ , . . . . . , , , .. j... _ . g _ ; ._g - . _2;_ m - 2 _ ;.-1_=
n e. . ,.
* &pa: - - ~ ~ - - -
....w= ..,:.---.... . . . . - . .
..-2:},__ - - - - _ y[h c - -, = = .
7,-.=,_.p . ,-n,=---- - - - . ' 4 - -
.eseaumende.- en es e .
h O.-.t - 3 _;;5..3.:_; =. u - " -Et.z:L-
. w'* - ~ -- = . .a=. .-h;;. u= := 4 :
w _--. _._ __; _ _ _;- - - - -
- - - - - - r.,_ . , m
, y n . , e.
l ; - j i
,r > . . . . . . . . . . . . . . . . . . . . i- . .
g t , i_ - u:. . ..*r 4 3. . . . . ;._.._ ,_. _ . ; ..m..,;, . m- .. . m .c.;p; y . wa_ J _t:",=. : r .A - - - a .
.y : 7.-- .! .. . . . . ,' :
;;^ _ - ,p.. . . . _ .; - . . . .s.. . a t-r - - - y j..g g. .
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=.=
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l (uComi) ' , 4 1 .
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=.g i
2 . . .
. _ . . --;:--- ~ ~ ,
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, x =.
. . . . . . . . . . . . .. . . - , - . . . . ; ; _n;;,;, _ ; _ -
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.O, go'- 3 LO 10 to' go 5 Count Rate (cpm)
ODCM, V. C. Summer, SCE &G Revision 13 (June 1990) 3.0 11
3.2 Dose Calculation for Gaseous Effluent 3.2.1 Gaseous Effluent Dose Calculation Parameters Section of Term, Definition Initial Use D, = average organ dose rate in the current (3.2.2.2) year (mrem /yr). D, = dose to an individual from radioiodine (3.2.3.2) and radionuclides in particulate form and radionuclides (other than noble gases), with half-lives greater than eight days (mrem). D, = average skin dose rate in current year (3.2.2.1) (mrem / year). Dt = current total body dose rate (mrem /yr) (3.2.21) Dp = air dose due to beta emissions from (3.2.3.1) noble gas radionuclides (mrad). Dy = air dose due to gamma emissions from (3.2.3.1) noble gas radionuclides (mrad). Ki = total body dose factor due to gamma emissions (3.2.2.1) from isotope i(mrem / year per uCi/m3 ) from Table 3.1-1. Li = skin dose factor due to beta emissions from (3.2.2.1) noble gas radionuclide i(mradlyr per pCi/m 3) from Table 3.1-1. M' = air dose factor due to gamma emissions from (3.2.2.1) noble gas radionuclide i(mradlyr per pCi/m 3) from Table 3.1-1. N' = air dose factor due to beta emissions (3.23.1) from noble gas radionuclide i(mrad per uCi/m3) from Table 3.1 1. P, = dose parameter for radionuclide i, (3.2.2.2) (mrem /yr per uCi/m3 ) for inhalation, from Table 3.2-1. D, = the release rate of nobla gas radionuclide (3.2.2.1) i as determined from the concentrations measured in the analysisof the appropriate sample required by Table 1.2-3 (pCi/sec). ODCM, V. C. Summer, SCE&G: Revision 15 (February 1991) 3.0 12
- - .. - , . - _ . . - - _ - - . ~ . . - _ . - - -. - _.. - . .- . - - Section of Term Definition initial Use k'= the release rate of non-noble gas radionuclide i as determined from the concentrations measured (3.2.2.2) in tha analysis of the appropriate sample required by Table 1.2 3 (pCi/sec). )
~ )
O, = cumulative release of noble gas radionuclide i (3.2.3.1) over the period of interest (pCi), i
~
Q ,' = cumulative release of non noble gas radionuclide i (3.2.3.2) (required by ODCM Specification 1.2.4.1) over the period of interest (pCi). R ,' = dose factor for radionuclide i and pathway j, (3.2.3.2) (mrem /yr per uCi/m ) or (m 2mrem /yr per pCi/sec) 3 from Tables 3.2 2 through 3.2-6. W, = relative dispersion parameter for the maximum (3.2.3.2) exposed individual, as appropriate for his exposure pathway j and radionuclide i.
= SUQ' for inhalation and all tritium pathways
= B7Q' for other pathways and non-tritium radionuclides 5UQ = the highest annual average relative concentration (3.2.2.1) in any sector, at the site boundary in sec/m3 3.17 x 10 8 = the fraction of one year per one second (3.2.3.1)
X/Q' = Annual average relative concentration for the (3,2.3.2)
. location of the maximum exposed individual for the site (sec/m3).
D/Q' = Annual average relative deposition for the location (3.2.3.2) of the maximum exposed individual for the site (m 2), ODCM, V. C. Summer, SCE&G: Revision 14 (December 1990) 3.0-13
3.2.2 Unrestricted Area Boundary Do;e 3.2.2.1 For the purpose of implementation of se: tion 1.2.2.la, (> 500 mrem / year - total body, - 3000 mrem / year - skin) the dose at the unrestricted area boundary due to noble gases shall be calculated as follows: D, = current total body dose rate (mrem /yr)
=
X/O E K k (46) l D, = current skin dose rate (mrem /yr)
= X/O (L + 1.1M,)k (47) where:
k = the release rate of noble gas radionuclide i as l determined from the concentration measured in the analysis of the appropriate sample required by Table 1.2 3 (pci/sec.). l X/O = the highest annual average relative concen. l tration in any sector, at the site boundary (for i value, see Section 3.1.2). l
- i. K., L., and M, will be selected for the appropriate
- eadionuclide from Table 3.1 1. s i !
t l' 3 2.2.2 For the purpose of implementation of section 1.2.2.1.b l (L 1500 mremlyr any organ) organ doses due to radiciodines and all radioactive materials in particulate form and radionuclides (other i than noble gases) with half-hves greater than eight days, will be I ! . calculated as follows: Do = current organ dose rate (mremlyr) l = E X/Q P, l (48) where: 1 ODCM, V.C. Summer, SCE &G: Revision 13 (June 1990) i 3.0 14 i
- - - , - . . - . - . - - . - - - - - - - , - - - . . , . ~ . -
X/O = the highest annual average relative con;entration in any sector, at the site boundary (for value, see Section 3.1.2) P, = dose parameter for radionuclide i, (mremryr per pCi/m i ) for inhalation, from Table 3.1-1.
,' = the release rate of non noble gas radionuclide i as determined ' rom the concentrations measured in the analysis of the :ppropriate sample required by Table 1.2 3 (pCi/sec).
3 2.3 Unrestricted Area Dose to individual 3.2.3.1 For the purpose of implementation of section 1.2.3.1 (Calendar quarter: 5 mrad - y and 10 mrad - p, Calendar year:
- = 10 mrad y and a 20 mrad - p) and section 1.2.5.1 (air dose averaged over 31 days: L O.2 mrad - y and 10.4 mrad p),the air dose in unrestricted areas shall be determined as follows:
Dy = air dose due to gamma emissions from noble gas radionuclide i (mrad)
= 3.17 x 10 " ! M, X/O b, (49)
I where: 3.17 x 10-e r- the fraction of one year per one second Q, = cumulative release of noble gas radionuclide i over the period of interest (pCi), ODCM, V.C. Summer, SCE &G: Revision 13 (June 1990) 3.0-15
ll i Dp = air dose due to beta emissions from noble gas radio. l nuclide i (mrad). i
= 3.17 x 10 8 E N, X7Q , (50) where, Ni = air dose factor due to beta emission from noble gas radionuclide i (mradlyr per uCi/m3) from Table 3.1-1.
3.2.3.2 Dose to an individual from radioiodines and radioactive materials in particulate form and radionuclides (other than noble gases), with half lives greater than eight (8) days (Calendar quarter:
% 7.5 mrem any organ, Calendar year: 515 mrem any organ) will be calculated for the purpose of implementation of section 1.2.4.1 as follows:
D, = dose to an individual from radioiodines and radio. nuclides in particulate form, with half lives greater than eight days (mrem)
= 3.17 x 10 s 3 g , w,,,6
, (51) in where:
W' ,,
= relative concentration or relative deposition for the maximum exposed individual, as appropriate for exposure pathway J and radionuclide i. -
WQ' for inhalation and all tritium pathways
= 2.2 x 10 6 sec/m3
=<
UTQ' for other pathways and non-tritium radionuclides
= 1.2 x 10-8 m 4 (See the notes to Table 3.2 7 and 3.2 8 for the origin of >
these factors.) ODCM, V. C. Summer, SCE&G: Revision 14 (December 1990) 3.0 16
R = dose f actor for radionuclide i and pathway j, (mrem /yr per pCi/m ) or (m 3 2 mrem /yr per pCesec) from Table 3 2 2. Q',
= Cumulative release of non-noble gas radionuclide i (required by ODCM Specification 1.2.41) over the period of interest (pCi)-
3.2.3.3 For the purpose of initial assessments of the impact of unplanned gaseous releases, dose calculations for the critical receptor in each affected sector may be performed using section 3.2,31 and section 3.2.3 2 equations as follows: (1) For each location, X/Q' and D/Q' will be calculat-d according to the methods of Section 3 3. using the measured meteoro-logical parameters for the period of the unplanned release. (2) The location of the critical receptors and the pathways j which should be analyzed are shown in Table 3.2-7. (For very rough calculations, the annual average X/Q and D Q for each receptor are shown in Table 3.2-8.) (3) The R,, for the appropriate exposure pathways and age groups will be selected from Tables 3 2 3 through 3.2-6. ODCM, V.C. Summer, SCE&G. Revision 13 (June 1990) 3.0-17
- - . . - . .. - . -. = _ . . . - . TABLE 3.21 i PATHWAY DOSE FACTORS FOR SECTION 3.2 2.2 (P,)* Page 1 of 3 AGE GROUP l (CHILD) ISOTOPE j INHALATION H3 1.125E + 03 C-14 3.589E + 04 NA 24 1.610E + 04 P-3 2 2.605E + 06 CR 51 1.698E + 04 MN-54 1.576E + 06 MN 56 1232E + 05 FE-55 1.110E + 05 FE 59 1.269E + 06 CO 58 1.106E + 06 CO 60 7.067E + 06 NI-63 8.214E + 05 NI65 8.399E + 04 CU 64 3.670E + 04 2N-65 9.953E + 05 ZN 69 1.018E + 04 BR 83 4.736E + 02 BR-84 5.476E + 02 BR-85 2.531E + 01 RB 86 1.983E + 05 RB 88 5.624E + 02 RB 89 3.452E + 02 SR-89 2.157E + 06 SR 90 1010E + 08 SR 91 1,739E + 05
- See note, page 3.0-20 Units - mrem /yr per pCi/m3 ODCM, V.C. Summer, SCE & G: Revision 13 (June 1990) 3.0-18
TABLE 3.21 PATHWAY DOSE FACTORS FOR SECTION 3.2.2.2 (P.) Page 2 of 3 AGE GROUP (CHILD) ISOTOPE INHALATlON
~
SR 92 2.424E + 05 Y-90 2.679E + 05 Y-91 M 2.812E + 03 Y 91 2.627E + 06 . Y 92 2.390E + 05 Y 93 3.885E + 05 ZR 95 2.231E + 06 2R 97 3.511E + 05 NB 95 6.142E + 05 MO 99 l 1.354E + 05 TC 99M 4.810E + 03
~
TC-101 5.846E + 02 RU 103 6.623E + 05 RU-105 ' 9.953E + 04 RU-106 1.476E + 07 AG 110M 5,476E + 06 TE-125M 4.773E + 05 TE-127M 1.480E + 06 TE-127 5.624E + 04 TE 129M 1.761E + 06 TE-129 2.549E + 04 TE-131 M 3.078E + 05 TE-131 2.054E + 03 TE-132 3.774E + 05 l130 1.846E + 06
*5ee note, page 3.0-20 Units - mrem /yr per pCi/m3 ,
ODCM, V.C. Summer, SCE&G Revision 13 (June 1990) 3.0 19
l TABLE 3 21 PATHWAY DOSE FACTORS FOR SECTION 3 2 2.2 (P,) Page 3 of 3 AGEGROUP (CHILD) ISOTOPE INHALATION 1131 1.624E + 07 l l-132 1.935E + 05 l l133 3.848E + 06 l l134 5,069E + 04 . j I135 7.918E + 05 i C5-134 1.014E + 06 l C5136 1.709E + 05 l C5137 '9.065E + 05 C5138 8.399E + 02 BA 139 5.772E + 04 B A 140 1.743E + 06 B A- 141 2.919E + 03 B A 142 1.643E + 03 LA 140 2.257E + 05 LA 142 7.585E + 04 C E-141 5.439E + 05 CE 143 1.273E + 05 CE 144 1.195E + 07 PR 143 4.329E + 05 PR 144 1.565E + 03 ND-147 3.282E + 05 Wa187 9.102E + 04 NP-239 6.401E + 04 NOTE: The P, values of Table 3.21 were calculated according to the methods of Reference 1, Section 5.2.1, for children. The values used for the various parameters and the origins of those values are given in Table 3.2-9 and its notes. Units - mrem /yr per pCi/m3 ODCM, V.C. Summer, SCE & G Revision 13 (June 1990) 3020
TABLE 3 2 2 PATHWAY DOSE FACTORS FOR SECTION 3 2.3 2 (R,)* Page 1 of 3 AGE GROUP (CHILD) l (N. A.) l (CHILD) ISOTOPE i INHALATION l GROUND PLANE j VEGETATION H3 l1125E + 03 (Total Body) 0 000E + 00 (Skin) 3.627E + 03 (Total Body) C-14 3.589E + 04 (Bone) 0.000E + 00 (Skin) 8.894E + 08 (Bone) NA 24 1.610E + 04(Total Body) 3 33E + 08(Skin) 3.729E + 05(Total Body) P-32 2 605E + 06(Bone) 0.000E + 00 (Skin) 3 366E + 09 (Borse) CR 51 1.698E + 04 (Lung) 5.506E + 06 (Skin) 6.213E + 06 (GI LU) MN-54 1576E + 06 (Lung) 1.625E + 09 (Skin) 6.648E + 08 (bver) MN 56 1.232E + 05 (GI LLI) 1.068E + 06 (Skin) 2 723E + 03 (GI-LU) FE-55 1.110E
- 05 (Lung) l 0.000E + 00 (Skin) ! 8.012E + 08 (Bone)
FE-59 1.269E + 06 (Lung) 3 204E + 08 (Skin) } 6 693E + 08(GI LU) CO-58 1.106E + 06 (Lung) 4.464E + 08 (Skin) l 3.771E + 08 (GI-LU) CO-60 7.067E + 06 (Lung) 2 532E + 10(Skin) ! 2 095E + 09 (GI LU) NI63 8.214E + 05 (Bone) 0.000E + 00 (Skin) l 3.949E + 10 (Bone) Ni65 8.399E + 04 (Gi-LU) 3.451E + 05 (Skin) 1211E + 03 (Gi LU) CU-64 3.670E + 04 (GI-LU) 6 876E + 05(Skin) 5159E + 05 (GI LU) ZN-65 9.953E + 05 (Lung) 8.583E + 08 (Skin) 2164E + 09 (bver) ZN-69 1.018E + 04 (Gi LU) 0.000E + 00 (Skin) 9.893E-04 (GI LU) BR-83 4.736E + 02(Total Body) 7 079E + 03 (Skin) 5 369E + 00(Total Body) BR 34 5.476E + 02(Total Body) 2.363E + 05 (Skin) 3.822E - 11(Total Body) BR IS 2.531E + 01 (Total Body) 0.000E + 00 (Skin) 0.000E + 00(Total Body) RB-86 1.983E + 05 (Liver) 1.035E + 07 (Skin) 4.584E + 08 (Uver) RB 88 5.624E + 02 (Uver) 3.779E + 04 (Skin) 4.374E - 22 (bver) RB-89 3 452E + 02 (Liver) 1452E + 05 (Skin) 1.642E 26 (Uver) ~ SR-89 2.157E + 06 (Lung) 2.509E + 04(Skin) 3.593E + 10 (Bone) SR-90 1.010E + 08 (Bone) 0 000E + 00(Skin) 1.242E + 12 (Bone) SR-91 1.739E + 05 (GI LU) 2 511E + 06 (Skin) 1.157E + 06 (GI-LU) See note, page 3.0 36 Reference 1, section 5.3.1, page 30, paragraph 1 explains the logic used in selecting these specific pathways. Critical organs for each pathway by nuclide in parentheses. Units - Inhalation and all tritium - mrem /yr per pCi/m3 Other pathways for all other radionuclides -m2
- mremlyr per pCi/sec ODCM, V C. Summer,5CES G Revision 13 (June 1990) 3 0-21
TABLE 3 2 2 (continued) i PATHWAY DOSE FACTORS FOR SECTION 3 2 3 2 (R,) Page 2 of 3 AGE GROUP l (CHILD) l (N.A.) l (CHILD) ISOTOPE l lNHALATION l GROUND PLANE l VEGETATION SR 92 2.424E + 05 (Gi LU) 8 631E + 05 (Skin) 1.378E + 04 (GI-LU) Y - 90 2.679E + 05 (GI-Lul) 5 308E + 03 (Skin) 6.569E + 07 (GI-LU) Y - 91 M J 2.812E + 03 (Lung) 1161E + 05 (Skin) 1.737E - 05 (GI LU) Y-91 l 2.627E + 06 (Lung) 1.207E + 06 (Skin) 2 484E + 09 (GI-l,U) Y - 92 2.390E + 05 (GI-LU) 2.142E + 05 (Skin) 4 576E + 04 (GI-LLI) Y 93 3 885E + 05 (GI LU) 2 534E + 05 (Skin) 4.482E + 06 (GI-LU) ZR-95 l 2 231E + 06 (Lung) 2.837E + 08 (Skin) 8 843E + 08 (GI-LU) ZR - 97 3 511E + 05 (GI LU) 3 445E + 06 (Skin) 1.248E + 07 (GI LU) NB-95 6.142E + 05 (Lung) 1605E + 08 (Skin) 2 949E + 08 (GI-LU) MO - 99 l 1.354E + 05 (Lung) 4 626E + 06(Skin) 1.647E + 07 (Kidney) TC 99M l4 810E + 03 (GI LU) 2109E + 05 (Skin) 5 255E + 03 (GI LU) TC 101 5 846E + 02 (Lung) 2 277E + 04(Skin) 4.123E - 29 (Kidney) R U - 103 6.623E + 05 (Lung) 1.265E + 08 (Skin) 3.971E + 08 (GI-LU) RU - 105 9 953E + 04 (GI LU) 7 212E + 05 (Skin) 5.981E + 04 (Gi-LU) RU 106 1.476E + 07 (Lung) 5.049E + 08 (Skin) 1.159E + 10 (GI-LU) AG - 110M 5 476E + 06(Lung) 4.019E + 09 (Skin) 2.581E + 09 (GI-LU)
- TE 125M 4.773E + 05 (Lung) 2.128E + 06 (Skin) 3 506E + 08 (Bone)
TE - 127M 1480E + 06(Lung) 1083E + 05 (Skin) 3 769E + 09 (Kidney) TE 127 5.624E + 04 (GI LU) 3 293E + 03 (Skin) l 3 903E + 05 (Gi-LU) TE - 129M 1.761E + 06 (Lung) 2,312E + 07 (Skin) 2 430E + 09 (GI-LU) TE - 129 2.549E + 04 (Gi-LU) 3.076E + 04 (Skin) 7 200E 02 (GI LU) i TE-131M 3.078E + 05 (Gi LU) 9.459E + 06 (Skin) 2.163E + 07 (GI-LU) TE-131 2.054E + 03 (Lung) 3 450E + 07 (Skin) 1349E - 14 (GI-LU) l TE - 132 3.774E + 05 (Lung) 4 968E + 06 (Skin) 3.111E + 07 (Gi-LU) I-130 1.846E + i(Thyroid) 6 692E + 06(Skin) 1.371E + 08 (Thyroid) Units - Inhalation and all totium - mrerh/yr per pCi/m3 Other pathways for all other radionuclides m2
- mremlyr per pCi/sec ODCM, V.C. Summer, SCE& G Revision 13 (June 1990) 3 0-22
TABLE 3 2-2 (continue) PATHWAY DOSE FACTORS FOR SECTION 3 2 3 2 (RJ Page 3 of 3 AGE GROUP ! (CHILD) l (N. A.) l (CHILD) ISOTOPE l INHALATION j GROUND PLANE l VEGETATION 1131 l 1624E + 07 (Thyroid) j 2 089E + 07 (Skin) 4 754E + 10 (Thyroid) 1-132 1935E + 05 (Thyroid) l 1452E + 06(Skin) 7 314E + 03 (Thyroid) 1-133 3 848E + 06(Thyroid) 2 981E + 06 (Skin) 8113E ,08 (Thyroid) 1-134 l 5.069E + 04 (Thyroid) 5 305E + 05 (Skin) 6.622E 03 (Thyroid) 1-135 l 7 918E + 05 (Thyroid) j 2 947E + 06(Skin) 9 973E + 06 (Thyroid) C5134 1014E + 06 (Liver) 8 007E + 09 (Skin) 2.631E + 10 (Liver) C5136 1709E + 05 (Liver) 1.710E + 08 (Skin) 2.247E + 08 (Liver) C5-137 ! 9 065E + 05 (Bone) 1201E + 10 (Skin) 2.392E + 10 (Bone) C5-138 ! 8 399E + O2 (Liver) 4102E + 05 (Skin) 9133E 11 (Liver) B A- 139 l 5 772E + 04 (GI LLI) , 1 194E + 05 (Skin) 2.950E + 00 (GI LLI) BA-140 j 1.743E + 06 (Lung) 2.346E + 07 (Skin) 2.767E + 08 (Bone) B A-141 i 2.919E + 03 (Lung) 4 734E + 04 (Skin) j 1.605E - 21 (Bone) B A-142 l 1.643E + 03 (Lung) 5 064E + 04 (Skin) 4.105E - 39 (Bone) LA-140 2.257E + 05 (GI-LLI) 2.180E + 07 (Skin) 3.166E + 07 (GI-LLI) LA-142 7.585E + 04 (Lung) 9.117E + 05 (Skin) 2.141 E + 01 (Gl-LLi) CE 141 5.439E + 05 (Lung) 1.540E + 07 (Skin) 4.082E 4 08 (GI-LLI) CE-143 1.273E + 05 (GI-LLI) 2 627E + 06(Skin) 1364E -+ 07 (GI-LLI) CE-144 1.195E + 07 (Lung) 8 042E + 07 (Skin) 1039E + 10(GI LLI) PR-143 4.329E + 05 (Lung) 0 000E + 00 (Skin) 1.575E + 08 (Gi LLI) PR- 144 1.565E + 03 (Lung) 2.112E + 03 (Skin) 3.829E - 23 (GI-LLI) N D- 147 3,282E + 05 (Lung) 1.009E + 07 (Skin) 9.197E + 07 (GI-LLI) W-187 9.102E + 04 (Gi-LLI) 2.740E + 06 (Skin) 5.380E + 06 (GI-LLI) NP-239 6.401E + 04 (Gi-LLI) 1.976E + 06 (Skin) 1.357E + 07 (GI LLI) U nits - Inhalation and all tntium - mremlyr per pCi/m3 Other pathways for all other radionuclides m2
- mremlyr per pCi/sec ODCM, V C. Summer. 5CE &G Revision 13 (June 1990) i 3.0 23
TABLE 3 2 3 1 PATHWAY DOSE FACTORS FOR SECTION 3 2.3 3 (R,)* l Page 1 of 3
.m. x+ l ~..n l 4, l ~..n ..n l .. . . n ~..n l .....n l - ..n y i _..._ n . - . .., i e. _ m . n . _ ..., i e. _ . i e. _ .., _. l..... )
H3 6 468E ,02 0 000t + 00 2157f + 03 0 000E
- 00 21571 + 03 0 000t ,00 4 398E ,03 0 000E + 00 i
l C 14 2 646t + 04 0 000f + 00 2 340E ,09 0 000( + 00 8189t + 08 0 000E + 00 2 340E + 09 0 0001 + 00 N A-24 1056E + M 1.385E + 07 1542t + 07 0 000E + 00 2 3001 37 0 000E + 00 1 851f + 06 0 000E + 00
. 1 P 32 2 030f + 06 0 000E + 00 1602E + 11 0 000E + 00 7 088E
- 08 0 000t 00 1924E + 11 0 000t 00 CR-51 1284f + 04 5 506E + 06 4 700t + 06 0 000t + 00 1729t + 05 0 000t + 00 5 641E + 05 0 000E + 00 M N-54 9 996E + 05 1625E + 09 3 900t + 07 0 000E + 00 1.118 E + 0 7 0 000E + 00 4 6801 + 06 0 000t + 00 MN 56 7168t + 04 1.068f + 06 2 862E + 00 0 000t + 00 0 0001 + 00 0 000E + 00 3 436E 01 0 000t + 00 7E 55 8 6S41 + 04 0 000f + 00 1.351 E + 08 0 0001 + 00 4 439E + 07 0 000E + 00 1757E + 06 0 000E + 00 f t 59 10151 + 06 3 204I + 08 3 919E + 08 0 000t + 00 3.384E + 07 0 000t + 00 5 0968 + 06 0 000E + 00 CO 58 7 770E + 05 4 464E + 08 6 0551 + 07 0 0001 + 00 8 8241 + 06 0 0001 + 00 7 251E + 06 0 000E + 00 CO 60 4 508E + 06 2.532E + 10 2.0981 + 08 0 000E + 00 71071 + 07 0 000E ,00 2 517f + 07 0 000E + 00 NI-63 3 3881 + 05 0 000t + 00 3 493E + 10 0 000t + 00 1.22 'I + 10 0 000E + 00 4192E + 09 0 000E + 00 NI-65 5 012 E + 04 3 451E + 05 3 0201 + 01 0 0001 + 00 0 0001 + 00 0 000E + 00 3 635E + 00 0 000E + 00 CU 64 1499E + 04 6 876E + 05 3 8071 + 06 0 000E + 00 7 934E -46 0 000E + 00 4 246E
- 05 l0000E+00 2N 65 6 4681 + 05 8 583( + 08 1904E + 10 0 000E + 00 5160E + 09 0 000t + 00 2.2 851 + 09 l0000E+00 2N 69 13221 + 04 0 000E + 00 3 855E-09 0 000t + 00 0 0001 + 00 0 000t + 00 3 581E 10 f0000f.00 SR 83 3 808E + 02 7 079E + 03 9 339t-01 0 000E + 00 0 000E + 00 0 000E + 00 1.1241 01 0 000f + 00 BR 64 4 004E + O2 2.3631 + 05 1.2561 22 0 000f + 00 0 000E + 00 0 000t + 00 1.52 7 E - 2 3 0 000t + 00 BR-85 2 044 E + 01 ~0 000E + 00 0 00E + 00 0 0001 + 00 0 000E + 00 0 000100 l 0 000E + 00 0 000E + 00 R8 86 1,904f + 05 1.035E + 07 2.234E + 10 0 ""Of + 00 2 827E + 08 0 000t + 00 2 671E + 09 0 000E + 00 R8-88 5 572E + O2 3 779E + 04 1874E 44 G , JO E + 00 0 0001 + 00 0 000E + 00 2 304E 45 0.000E + 00 R 8-89 3 206E + 02 1452t + 05 3 414E 52 0 000E + 00 0 000E + 00 0 000E + 00 4 0561 53 0 000E + 00 SR49 2.030t + 06 2.509E
- 04 1.258E + 10 0 000k + 00 1.280t + 09 0 000E + 00 2 6431 + 10 0 000E + 00 SR 90 4 088E + 07 0.000E + 00 1.216E + 11 0 000E + 00 l 4 2301 + 10 l 0 000E + 002 5531 + 11 0 000E + 00 SR-91 7.336t + 04 2 511E + 06 3.215E + 05 0 000E + 00 f 0 000E + 00 0 000E + 00 6 758E + 05 0 000t + 00 (PASTURE) (PASTORE) (F E E D) (PASTURE) (P A STURf)
- See note, page 3.0-36 Unit 5 -
Inhalation and all tritium . mrem /yr per pr Dm3 Other pathway 5 for all other radionuclids m2
- mrem /yr per pCi/Sec ODCM, V.C. Summer, SCE & G Revmon 13 (June 1990) 3021
1 TABLE 3 2-3 (continued) - l PATHWAY DOSE FACTORS FOR SECTION 3 2 3 3 (R,) Page 2 of 3
.a ao l -..n l 4.,
l - ..n l m..n l ...o l w ..n l ~..n l ~ . e. l von l + . .re= l aw=m.=t lanonwu l m cew we.? ! m an wa. ' as w w..r los mt wo l vietter.o. 5R 92 l 140Ci + 05 8 631E 05 l 5 0054 + 01 l00001+00 l 0 0001 + 00 l0000t+00 l 10541 + O2 l0000t+0 Y 90 2 688f + 05 5 308f .03 9 406E + 05 0 0001 + 00 l 2 335105 0 0001 + 00 1129E + 05 l0000E+00 Y 91M 2 7861 + 03 1161f + 05 5 1 8768 15 0 000t + 00 0 000f + 00 l 0 0001 + 00 2 2901 = 16 l0000t+00 Y-91 2 450t + 06 1.2076 + 06 5 251E + 06 0 000f + 00 6 324E + 05 0 000t + 00 6 3021 + 05 g000t+00 Y-92 1266f + 05 2.142 f . 05 l 1026E + 01 0 0001 + 00 ; O 0001 +00 0 000f 4 00 1234E + 00 0 0001 + 00 Y 93 1666E
- 05 2 534E + 05 1 776E+04 0 Ow . + 00 l 2 386161 0 000( + 00 2 046E + 03 0 000t + 00 2R 95 l 17501 + 06 2 8371 + 08 6 257f . 05 0 000t + 00 f1090t+05 0 000( + 00 9 910t + 04 0 000E
- 00 2R 97 1 400E +05 3 445E + 06 4 4468 + 04 0 0001 + 00 l 4 980E 35 l0000f+00 5 339E + 03 0 000t 00 NS 95 4 7881 + 05 1605E + 08 2 062f + 08 l0000E00 !1213E+07 f 0; 000t + 00 2 475E + 07 0 000f . 00 MO 99 1 348E +05 4 626E + 06 3106t + 08 l 0 000t + 00 j 1 5231 02 l0000t+00 l 3 731E + 07 0 000( + 00 TC 99M f 2 030E + 03 2109E + 05 1646f + 04 0 000t + 00 l 0 000f + 00 !00001+00 1 978E +03 0 000f + 00 TC 101 8 442i + 02 2 2 77E + 04 1423E 56 0 000t + 00 l0000(+00 0 000( + 00 6 530s-58 0 000f + 00 RU 103 5 516f .05 1265E + 08 1055E + 05 0 000f + 00 ! 7 573( + 03 0 000t + 00 12651 + 04 0 000E + 00 RU 105 l 4 844E + 04 7.212 E + 05 3 204 E + 00 0 0008 + 00 l 0 000t + 00 l0000t+00 3 8511-01 0 000f . 00 RU 106 l 1156E + 07 5 0496 + 08 1445E + 06 0 000t + 00 f 4 266E + 05 l 0 000t +17341 00 + 05 0 000E + 00 AG 110M 3 668f + 06 4 019E + 09 14616 + 10 0 0001 + 00 17521 + 09 l00001+00 f 3 984E +09 f0000E+00 TE - 12 5M 4 466E + 05 21281 + 06 1508E + 08 0 000t + 00 f 1799E + 07 l0000100 1 809E+ 07 0 000t + 00 TE 127 M 1 312i +06 1083E + 05 10371 + 09 0 000t + 00 l 2 0461 + 08 l 0 0001 + 12441 00 + 08 0 000f + 00 TE-127 2 436f . 04 l 3 2931 + 03 13596 + 05 0 000f . 00 1269f 65 0 0001 + 00 1594E + 04 0 000f + 00 TE - 129M 1680t + 06 2 312t + 07 1.392C + 09 0 000f + 00 7 559t + 07 0 0001 00 1672t + 08 0 000t + 00 Ti-129 l 2 632E ,04 1 3 076E + 04 2.187f 07 0 000f + 00 0 000E + 00 0 0005 + 00 2 624E 08 l0000(+00 TE 131M 19881 + 05 9 459E + 06 2 288E + 07 0 000f + 00 1653E 15 0 000f + 00 2 747E + 06 l0000f+00 Tf 131 8 218 E + 03 3 450! + 07 1 3848 30 0 000f + 00 0 000t + 00 0 000E + 00 1688t 31 j 0 000t + 00 TE 132 3 4021 + 05 4 968t + 06 6 5th + 07 0 000E + 00 l 1041t 01 0 000E 00 7 842E + 06 i 0 000E + 00 1 130 15968 + 04 6 6921 + 06 8 754E + 08 0 000f + 00 f 7115E 45 0 0001 + 00 1051t + 09 0 000f + 00 (PASTURf) (PASTU8f) j (F E E D) (PASTURt) (PASTURI)
U ru ts -
!nhalation and all tritium mremiyr per pCi/m3 Other pathways for all other radionuclides -m2
- mrem /yr per pCii5ec ODCM, V,C. Summer, SC E & G Revision 13 (June 1990) 3025
q TABl.E 3 2 3 (continued) 1 I' PATHWAY DOSE FACTORS FOR SECHON 3 2 3.3 (R ) t Page 3 of 3 j 1 aut omov+ -=# a n n in a l : =6 a% n chs a h ti ow+ an n ems a n r> vue s = 9 oms ag n v on l = a ato= ~lcaovmoeunel cast 0*was lvasw*utat j m m* ws lmwtutat f m oaf usa f vtostaron j 4 131 l 1484E + 07 2 089E + 07 10534 + 12 l 0 000f + 00 l 1567t + 080 0001 + 00 12441 + 12
?
l 0 000t + 00 l
)
1 132 16941 + 05 14521 + 06 11881 + 02 0 000t + 00 ! 0 000s + 00 0 0001 + 00 1638t + 02 0 000( + 00 j l I-133 3 556f + 06 2 981E + 06 9 601 t + 09 0 000f + 00 i 17?6t 22 0 0001 + 00 1 15 3 t + 10 0 000t + 00 1 134 4 4521 + 04 5 30$t + 05 6 402t 10 0 000f + 00 0 0001 + 00 0 000t + 00 1 0171-09 4 000t + 00 1-135 6 958t + 05 2 947t + 06 2 002t + 07 0 000t + 00 l 0 000t + 00 0 000E + 00 2 406t + 07 0 000f + 00 C5 134' 7 028f + 05 8 007E + 09 6 801E + 10 0 000t + 00 l 2191t + 10 0 0001 + 00 2 0401 + 11 0 000t + 00 l C5 136 1.345f + 05 1710t + 08 5 795f + 09 0 000t + 00 I 1729t + 07 0 000t + 00 1 744 t + 10 0 000t + 00 l C5-137 61181 + 05 1201t + 10 6 0241 + 10 0 000t + 00 2 0961 + 10 0 0001 + 00 1087f + 12 j 0 0001 + 00 l C5>138 8 764E + 02 4102t + 05 l 2180f-22 0 000f 00 j 00001 +00 0 0001 + 00 6 628f - 22 0 0001 + 00 BA 139 5 096! + 04 1194E + 05 2 8741-05 0 0001 + 00 ! 0 0006 + 00 0 000t + 00 3265f,06 l 0 000t + 00 BA*140 1596f 06 2 346E + 07 2 4101 + 08 0 000t + 00 i 6 409f + 05 0 000t + 00 2 893t + 07 l; 0 000t + 00 4 B A 141 4 7461 + 03 4 734I + 04 4 9161 44 0 0004 + 00 !. D000f ,00 0 000t + 00 $899E 45 l 0 000t + 00 l BAa142 1554f + 03 5 064f + 04 1049E 78 0 0008 + 00 l 0 000f + 00 0 0001 + 00 1 2591 79 0 0001 + 00 LA 140 1680E + 05 21801 + 07 18801 + 05 0 0001 + 00 4 563t 12 0 0001 + 00 2.2531 + 04 j 0 000t + 00 LA 142 5 950f .04 91176 + 05 1 0781 05 0 0001 + 00 0 000t + 00 0 0001
- 30 1278f 06 l 0 000f + 00 i ct 141 5.1661 + 05 1540f + 07 13664 + 07 0 000t + 00 f 7 008E
- 050 000E e 00 1640f . 06 0 0001 + 00 Cf 143 1.162f + 05 2 627f + 06 1536t + 06 0 000f ,00 l 1039( 14 0 000t + 00 1844E
- 05 f 0 000t + 00 CE 144 l 9 842E + 06 8 0421 + 07 1.3341 + 08 0 0001 + 00 l 3 7491 + 07 0 000f . 00 1601t + 07 l00001+00 PR 143 4 326f + 05 0 000! + 00 7 8451 + 05 0 000f 00 2 7711 + 03 0 000t + 00 9 407E + 04 0 000f 00 PR-144 4 2841 + 03 2112E + 03 1.1711 48 0 000t + 00 l03001+00 0 0001 + 00 f1259549 l0000t+00 ND 147 3 220t +05 1009E + 07 5.7431 + 05 0 000t + 00 l 6 902 ' + O2 0.000t + Or 6 8651 + 04 0 000t + 00 W .187 3 9625 + 04 2 740f + 06 2 501E + 06 0 000t + 00 5 275E v2 0 000t + 00 2 9831 + 05 0 000t + 00 NP-239 5 950t + 04 19768 + 06 9 400t + 04 0 0001 + 00 l 1025f.07 - 0 0001 + 00 11321 + 04 0 000f + 00 (PA5 TURI) (PASTURE) l (fi f D) g (PASTURE) (P ASTURE)
U nits - Inhalation and all tritium - mrem /yr per pCum3 Other pathway 5 for all other radionuclide5 -m2
- mrem!yr per pCi/Sec ODCM, V.C. Summer, SCE&G, Revi$ ion 13 Oune 1990) 3026 w
TABLE 3.2-4 PATHWAY DOSE FACTORS FOR SECTION 3 2.3 3 (R,)* Pag 1 of 3 f om f ou. act saas 4. . . f osoi f .ie, f f a .m o, 4 .m a mco,
., e. , j . . . . . i _ . . ... i e. _ . % +....j .m. . i . .., j .. _ . i + , . . -
H3 1 125[ + 03 0 000E + 00 14218 + 03 21181 + 02 14211 + 0J 2 543E + 01 3 2 899E + 03 3 62 71 + 03 C 14 3 589E + 04 0 000t + 00 1195E + 09 3 834f + 08 4181E + 08 4 601E + 07 1 1951 + 09 8 8941 + 08 NA 24 16101 + 04 1385E + 07 8 853E + 06 1725t 03 1321t 37 2 070t ' 04 10631 + 06 3,729t+ 05 P 32 2 605i + 06 0 000t + 00 f 7 7751 + 10 7411t*09 3 4401 + 08 8 8931 + 08 9 3351 + 10 3 3661 + 09 CF 51 1698f ,04 5 506E + 06 5 3984+06 4 661t + 05 1985E + 05 5 593E 04 6 4781
- 05 6.213t + 06 MN 54 15761 + 06 16251 + 09 2 0971 + 07 8 0111 + 06 6 012 E + 06 9 613t + 05 2 5171 + 06 6 648E + 08 MN 56 12321 + 05 1068E + 06 18651 + 00 2 8371 51 0 0001 + 00 2 9241 52 2 238E 01 2 723t + 03 f t 55 11108 + 05 0 000t + 00 1118E + 08 4 571t + 08 3 6731 + 07 5 486L + 07 1453E + 06 8 012E + 08 FE 59 1.269f + 06 3 204E + 0B 2 02 51 + 08 6 338E + 08 1749E + 07 7 605t + 07 2 633E + 06 6 693E + 08 CC 58 1 106 E + 06 4 464 + 08 7 080t + 07 9 596E + 07 1012t + 07 1 152 t + 0 7 8 487f e 06 3 7 71 t + 08 CO -60 7 067f + 06 2 5321 + 10 2 391E + 08 3 8386 + 08 8103t + 07 4 6051 + 07 2 8 701 + 0 7 2 095E + 09 Ne64 , 8 214! + 05 0 0001 + 00 2 9641 + 10 2 912t + 10 1036f .10 3 4951 + 09 3 557f + 09 3 9491 + 10 NI65 8 399E + 04 3 451E + 05 1 9091+01 4 061t 51 0 000t + 00 4 873t 52 J 298E 00 1211E
- 03 CU-64 3 670t +04 6 876E ,05 3 502 t + 06 1393t . 05 7 2998 46 1672t 06 3 9071 + 05 5159E + 05 2N 65 9 953t + 05 8 583E + 08 1 1011 10 1000E + 09 2 985t + 09 1200E 08 1.322t +09 2 1641 , 09 2N 69 1.018E + 04 0 0001 + 00 1 1231- 09 0 0001 + 00 0 000t + 00 0 000E + 00 104 3 t - 10 9 8931-04 BR 83 4 736E + O2 7 079E + 03 4 3991 01 9 519E 57 0 000t + 00 1.142 E - 57 5 1901 02 5 3691 + 00 BR-84 5 476t + 02 2.363E + 05 6 508t 23 0 000t + 00 0 000t
- 00 0 000t + 00 7 758t 24 l 3 822t 11 8R 05 [ 2 !,311 + 01 0 000t + 00 0 000f + 00 0 000t + 00 0 000l + 00 0 000t + 00 0 000E + 00 0 000t + 00 I
RB-86 19831 + 05 10351 + 07 8 804E + 09 5 8161 + 08 1 1141 + 08 6 9791 + 07 1.053t + 09 4 5841 + 08 R8-88 5 624E + 02 3 779E + 04 71501 45 0 0001 + 00 0 000t + 00 0 C00t + 00 8789f 46 4 374f 22 R8 89 3 452E + 02 14521 + 05 1.39 71 + 52 0 000t + 00 0 000t + 00 0 000t + 00 1659E 53 1642E 26 SR-89 2.1571 + 06 2 509E + 04 6 618f + 09 4 8158 + 08 6 730E + 08 5 778E + 07 13906 + 10 3 5931 + 10 SR 90 1010E + 0a 0 000E + 00 1.117E + 11 1040t + 10 3 88 71 + 10 12488 + 09 2 346E
- 11 1243E + 12
$R 91 1739E
- 05 2.511 E + 06 2.878E + 05 55.292t 10 0 000t + 00 6 351E 11 6 0$0! + 05 1.15 7 E + 06 (PASTURE) (P 45TURE l (f ttD) (PASTURE) (PASTURE)
*See note, page 3.0-36 Units -
Inhalation and all tritiurn - mrem /yr per pCi/m3 Other pathways for all other radionuclides -m2
- mrem /yr per pCi/SeC ODCM, V C. Summer, SCE &G Rusion 13 (June 1990) 3 0-27
TABLE 3 2 4 (contmued) PATHWAY DOSE FACTORS FOR SECTION 3 2 3 3 (R i ) Page 2 of 3
.ci o.u. co. l .... l ooi l m -o, l os o, l ou o.
o.... l a-o. l _ _ .... i. .o.... I e, _ . h. . - . i e., _.. t ., _ ., i . . . I m ,.. . 5R-92 2 4241 + 05 8 6316 + 05 4134E e 01 3 4921 48 ; o 000E+00 4191t - 49 I87061+01 13786 + 4 Y 90 2 679E +05 5 3081 + 03 9171( + 05 4 879i + 0$ 2 277i 05 5 855i + 04 1101( + 05 6 569i + 7 Y - 91 M 2 8121 + 03 1 1611 + 05 5 6221 16 0 0001 + 00 0 000f + 00 0 0001 + 00 6 3441 17 1 7371 5 Y 91 2 6274 + 06 1207E + 06
- 1994 + 06 2 400t + 08 6 2611 + 05 2 880f + 07 6 2401 + 05 / 484E ,9 Y 92 2 3901 + 05 2142t + 05 7 310E + 00 6 9598-35 l0000f.00 8 350t 36 8 791t 01 4 576f 4 Y 93 3 8856 + 05 2 534E + 05 1573E + 04 1 5471 07 9 134!-61 '
1 8571 08 1888t + 03 4 4821 + 6 2R+95 2 2314 + 06 2 837E + 08 8 7668 + 0% 7 3281 + 07 10541 + 05 8 643! + 8 61064 + 08 l 1160t + 05 ZR 97 3 511t + 05 3 445I + 06 41991 + 04 7 0158 01 1 4 7031 35 8 418f 02 5 042 f + 03 1248E + 7 N8 95 6142t + 05 1605f + 08 2 287t + 08 2 288E + 09 l 1346t + 07 2 673t + 08 2 747t + 07 2 949t + 8 MO 99 1.354E + 05 4 626E + 06 1738E + 08 2 4%t + 05 f8512103 2 9475 + 04 2 0864 + 07 164 71 + 7 TC 99M 4 810t ,03 21091 + 05 14 741 + 04 6 9151 18 f0000t+00 8 2981-19 1771t + 03 5 255E + 3 TC 101 58468+02 2 277E + 04 5 593E 58 0 0001 + 00 ; O 000t + 00 0 000t + 00 2 5661 59 4123 E -29 RU 103 6 623E + 05 1265E + 08 1106E + 05 4 009E + 09 l 7 9521 + 03 4 allt + 08 f1329t+04 3 9 71 t + 8 RU 105 9 9531 + 04 7.212E + 05 2 493E + 00 58851 25 7 0611 26 2 9978-01 5 981E + 4 l00004+00 RU 106 1476E + 07 5 0491 + 08 1437L + 06 6 902t + 10 4 243E + 05 8 282t + 09 1 7256 +05 1 159E+10 AG 110M 5 4768 + 06 4 019E + 09 16781 + 10 6 742E ,08 4 5768 + 09 8 0901 + 07 2 0131 + 09 2 5811 + 9
.c 125M 4 7731 + 05 2.128E + 06 7.377E + 07 ;
5 6901 + 08 8 802E + 06 6 828f + 07 8 8531 + 06 3 506t + 8 Ti 127M 1480t + 06 1083E + 05 5 932E ,08 ! 5 060t + 09 1.171 t + 08 6 0728 + 08 3 7696 + 9 l 7118E + 07 ft 127 5 624E + 04 3 2931 + 03 0 0001 + 00 1 1911 + 05 l 16074 08 l 1929 09 f 13968 + 04 f 3 903t + 5 Ti 129M 1761E + 06 2.312E + 07 7 961t + 08 l52451+09 4 3241 + 07 6 2941 + 08 9 563E + 07 2 461 + 9 TE 129 2 5491 + 04 3.076E + 04 7 96t 08 0 000E + 00 0 000t + 00 0 P00E
- 00 9 6411,09 7 204E 2 l
7E 131M 3.078E + 05 9 4591 + 06 2.J44E + 07 l98151+03 1621E-15 1 1781 + 03 l 2 0948 + 06 216)( + 7 TE 131 2 0548 + 03 3 450E + 07 84891 32 l 0000t 00 0 000t + 00 0 000t + 00 1 0361 32 , 1349E 14 Tl 132 3 7 74 E + 05 4 968E + 06 4 551E + 07 l 9 325E + 06 7 2721 02 1 1191 + 06 5 480! + 06 3 1111
- 7 4-130 18461 + 04 6.6921 + 06 3 8454 + 08 6 7581 04 3125E 45 8109E 05 4 617E + 08 13 71t + 8 (P45 TURI) (PA5TURE) (f f ED)
(P ASTURE) l (PA5TURt) Units - Inhalation and all tritium - mrem /yr per pCi m3 Other pathways for all other radionuclide5 m2
- mrem /yr per pCusec ODCM, V.C. Summer, SCE S G Rewsion 13 Oune 1990) 3028
- . . - - ~ - - - - - = _ - _ - - - . _- -- .
TABLE 3 2 4 (Continue! PATHWAY DOSE FACTORS FOR SECTION 3 2 3 3 (R,) Page 3 of 3
.a c.au - l .t ~
l , = . , l m o, l m .o, l ~. l oo, l oun l .mo, acron l ea. + r-o= lcaov=on+=ifcaswwwo l m ee wtat l cas tca vo j m cutuist l ca cu ws l aut+ nun F131 1624t + 07 2 089E + 07 4 J 3 31 + 81 5 503t + 09 6 448 t
- 0 7 6 604E + 08 5 2011 + 11 l 4 7541 + 10 F132 1935E ,05 14521 + 06 51298 + 01 2 4291 57 0 000t
- 00 2915E 58 7 07?f + 01 7,3141 + 03 F133 3 848E + 06 2 981E + 06 l 3 9451 + 09 1304t
- 02 7 2991 23 15641 + 01 4 7 3 71 + 09 8113t + 08 F134 5 0691 + 04 5 305t 05 3 624f 10 0 000( + 00 0 000t + 00 0 000t + 00 4 3868-10 6,6221 03 F135 7 918t + 05 2 947t + 06 8 607t + 06 1039E 14 0 000t + 00 1.247t 15 1034t + 07 9 973t + 06 C 5 < 134 1014t + 06 8 0074 + 09 3 7158 + 10 1513t
- 09 l 1197f + 10 1816i + 08 1115f
- 11 2 63 t t + 10 C5136 l 17091 + 05 17101 + 08 2 773E + 09 4 4761 *07 824 + 06 5 3114 + 06 8 344E ,09 2 247f .08 C5137 9 065E + 05 1201t*to 3 224t + 10 13341 + 09 1* + 10 1600E + 08 9 672E + 10 2 3921 + 10 CS 138 8 399E + 02 4102t + 05 5 5284 23 0 000t 00 0vv06 00 0 000f . 00 1 6811-22 9 1331 11 8 A 139 5 772t + 04 1 194t + 05 1231t - 05 0 000t + 00 0 0001 00 0 000t + 00 1 3981 06 2 950t + 00 BA140 1.743 E + 06 2.346f + 07 f 11711 + 08 4 384E + 0? 3 Ital + 05 52611+06 1406E + 07 f 2 7671 + 08 8 A 141 2 919 + 03 4 734f + 04 1 8941 45 0 0001 + 00 0 000t 00 j 0 000t + 00 2 2731 46 1605E 21 B A 142 16431 + 03 5 0641+04 1 2081 -79 0 0001 + 00 0 0001 + 00 0 0001 + 00 1 4501 80 4 1051 39 LA 140 2.257E + L5 2.180t 07 18941 + 05 5 492E + O2 4 596E -12 6 590t 01 2 2691 + 04 3166E + 07 LA 142 7.5858 + 04 91171 + 05 5 203t 06 0 300f + 00 f 0 0001 + 000 000t + 00 l 61661 07 2141t+01 CE-141 5 4391 + 05 1540E + 07 1361E + 07 1.382t + 07 l 6 980 05 1658E + 06 1633t
- 06 4 082t + 08 Cl 143 1.27 3 t + 05 2 6271 + 06 14888 + 06 2 516E + 02 3 020E + 01 1787f + 05 13641 + 07 f 1006E 14 Cl 144 1 1951 + 0 7 8 042E + 07 1326t + 08 1893E+08 3 727E + 07 2 271E + 07 1592t+07 1 ( 391 + 10 PH4143 4 329E + 05 0 000E + 00 7 754E + 05 3 609E + 0i 2 738E + 03 4 3 311 + 06 9 297E + 04 1 s75E + 08 PR 144 1.5651 + 03 2.1121 + 03 2 040E - 50 0 0004 + 00 0 0001 + 00 0 0001 + 00 2 3531 %1 ! 3 829t 23 ND 147 J282E+05 1009E + 07 5 7121 + 05 15051 + 0 7 6 8641 + 02 18054 + 06 6 8461
- 04 91971 + 07 W 187 9102E + 04 2 740f + 06 2 4201 + 06 2 7901 + 00 5103 E - 22 3 348f .01 2 8861 + 05 5 3801 + 06 l NP 239 6 401E + 04 19761 + 06 91381 + 04 2 J32E + 03 j 9 336t 08 2 6791 + 02 1 100t + 04 ! 13571 + 07 (PASTURI) (PASTURI) (f E E D) - (P A STUHt) (P A 5TURE )
l l Units - Inhalation and all tritium - mrem /yr per pCi/m3 Other pathways for all other radionuClide5 m2
- mremlyr per pCi/5eC ODCM, V C. Summer, SCE&G Reo5 ion 13 Oune 1990) 3029 l
l
TABLE 3 2 5 PATHWAY DOSE F ACTORS FOR SECTION 3 2 3 3 (R,)* Page 1 of 3 l c n .u. l m..ue, ! e n..u., j o n ..u .. j cn..u., j m..u.,
.u .. . 1 l s o ..w e, l
_, .....~. 1 . . , . . . , i < e m . .- . imem...,j ..m._. ~ . .., j me .. . ..m. H3 1272i e 03 l 0 000t e00 8 9931 02 l1754ie02 l 8 993i 02 2104i e 01 18351e0l 2 3421e03 C 14 2tiXte04 0 000t + 08 4 859i e 08 2 0401
- 08 1 700t 08 l 2 448i
- 0 7 4 859i e 08 3690te08 N A 24 l 13711 e 04 13851+07 4 2554 06 1 0841 03 6 3478 38 1 3018 04 5 1101 e05 l 2 3891 05
+,
P 32 18ket + 06 0 0001
- 00 3153t .10 3 9318 e 09 13958 e 08 4 7178 +08 3 7851 e10 l 1 6088+09 CR 51 2 096l e 04 }$506406 8 3871
- 06 9 471t 05 3 0854 e 05 1137f 05 1 0061 06
.. : +
- f. 10?
MN$4 19841 e 06 ! 1 625Ie 09 2 875t + 07 .l 1436l 07 8 2401 e06 l1723te06
- _. 3 450t e 06 l 9 340t ._
MN % $ 144 E e 04 i 1 0681
- 06 4 8568 01 03021-52 0 0008 + 00 t 9628 53 58298-02 9 4511 e 02
% 12401 e 05 I O0001+00 4 4541 *07 2 382t e 08 1463t 07 2 859t
- 07 5 7901 *05 3 2596 + 08 18 59 152pt e 06 3 204: e 08 2 8614 , 08 l1171te09 2 4701
- 07 l1405108 3 720t 06 98951e08
+
CO $8 1 3448*06 4 abat 08 1095t . 08 19428 *08 1596l e 07 2 3301
- 07 13131 e 07 6 0341 e 08 CO 60 8 7?ot + 06 25321 + 10 36211 08 7 boot . 08 122 71 e 08 9 1201 07 4 3454 eof a 2388 + 09 NI63 5 8008 +C5 0 000t
- 00 1 182 t 10 1 $198 e 10 41308 e 09 l 18231 e 09 1 419t e09 1 606l 10 t
Ni 65 36721e04 3 Allt + 05 4 6921 *00 1 3051 51 0 0001 e 00 1 % 68 -52 5 6471 01 3 966l *02 CU 64 6144l + 04 6 8761 05 32931e06 i 713t 05 6 863t 46 2 072t - 06 3 673t
- 05 6 465L
- 05 2N 65
- 240t
- 06 8 583t + 08 7 3154 e09 8 6881
- 08 1983t e 09 10411 + 08 8 7791 08 14 711 e 09 2N 69 15841+03 0 0001 00 17608 11 0 000t 00 0 0006
- 00 0 0001 e 00 1 6354 12 2 0671 05 BR 83 3 4401 e 02 7 0791
- 03 1 7908 01 5 0668 57 0 0006 00 6 0791-58 2112t-02 2 9111 00 BR 84 4 3281
- 02 2 3631 + 05 2 8771 23 l0000te00 0 000t 00 ! 0000t*00 3 4291 24 l 2 2'ill 11 dN 85 1 8321 01 0 0001 + 00 0 0001 + 00 0 000t
- 00 0 0004 + 00 0 000 00 0 0001 00 0 0001 + 00 h8-86 1 904l+05 1035E e 07 4 746t e 09 4 1011 *08 6 0061 + 07 4tille07 5 675[ + 08 2 7721 e08 F8 88 5 4%i e 02 3 779E e 04 3 8861 45 0 000t + 00 0 0001
- 00 0 000t 40 4777: 46 3168t 22 RB 89 3 520I
- 02 1452E*05 7 957E 53 0 000t 00 0 000t e 00 0 0001 e 00 94541 54 12471-26 SR 89 2 4161+06 2 5091 e 04 2 674l ,09 2 Salt 08 2 719f eDe 3 0541 e 07 5 617L e 09 1 Sill e to SR 90 1080t + 08 0 000t 00 6 6121 e 10 8 0491 e 09 2 301t .10 9 6598 + 08 1389t e 11 ! 7 5071 e 11 5R 91 2 592t e OS 2 Stil
- 06 2 409f 05 57941 10 0 0001 e 00 6 953t 11 }50641+05 1291L
- 06
} :pa5tupti tea 5tual) (s tio) tealiupri (pa5tual)
- See note, page 3 0 36 Units -
Inhalation and all tritium mrem /yr per pCum3 Other pathways for all other radionuclides m2
- mrem /yr per pC1/5ec ODCM, V C. Summer, SCE &G Revision 13 (June 1990) 30 30 !
TABLE 3 2 5(contmued) PATHWAY DOSE FACTORS FOR SECTION 3 2 3 3 (R,) Page 2 of 3 eu .xe l gn m.., l .... l muu., l mua., l gnua., ! anuu.4 muu., anua., l !
.~, u...se i..,e.... ! . . . ! - . . . . . i .. , m. . . .....,,j .e _ .
I j e,m... SR 92 l11921e05 , 8 611t e 05 2 2 7 71
- O t 2 5161 48 3 Otti -49 4 76 i e 01 l 10121
- 04
_ {00001e00
- to 5 592I e 05 5 308I + 01 l1074Ie06 7 4701 + 05 l 26661 05 l 8 565i
- 04 1289i e 05 1025i e 08 Y ,91M J 2001,03 11614 + 05 51J 91 18 0 0001 e 00 0 0001 , 00 0 0001 e 00 6 2601 19 2 2851-07 7 91 4 9368e 06 12071 + 06 6147f + 06 3 9101 08 719 71 e 05 4 691f e 07 7 700E e 05 f { 2121
- 09 v-92 16481 + 05 2 telt e 05 2 elet 00 3 5221 -35 0 000t + 00 4226f-16 34021 01 f 2 3604 04 7 93 5 7921 05 21341 + 05 1312( e 04 1688f (7 7 6201 61 2 0268 08 1511t
- 03 4 983t e 06 28-95 2 6881 06 28371 00 1201t
- 06 10921 + 09 1 5851 05 1310f
- 08 1441f+05 12511 + 09 JR-97 6 3041 + 05 3441+06 4 2254 + 04 9 2311 01 4 7328-35 11081-01 5 0131
- 03 , 16738e07 N8 95 7512te05 1 6051 e08 3 3381 e 08 42518 *09 19634 e 07 51011 e 08 4 0088
- 07 f4551t08 MO 99 2 6881 05 4 6268+06 1023t e OS 1892t *05 50131 03 2 2 701 + 04 12281 e 07 1293t
- 07 fc 99M 61281 + 01 21091 + 05 10551 + 04 6 "11 it l00001+00 77661 19 12671*03 5 0111 + 03 TC 101 6 672t
- 02 2 2 771 + 04 1 3431,58 0 000t
- 00 0 000t + 00 0 0001 + 00 1 5081 59 3 2291 29 RU 103 7 8321 + 05 12658 + 08 1513t
- b5 71628+09 1086t e 44 85951+08 18154 + 04 5 706t 08 RU.105 9 0401 + 04 7 2121 + 05 12631 + 00 3 9001 25 0 000t 00 4 6801-26 1519f - 01 4 0391 + 04 RU 106 16081+07 5 0491 + 08 17991 + 06 1 1301 + 11 5 312t + 05 13561 + 10 2 1598 +05 14641+10 AG 110M 6 752t 06 4 019t + 09 25591+10 13451 09 6 9828 + 09 1614f + 08 J071t 09 4 0311 + 09 f f 125M 5 360 , 05 2128f + 06 8 8631 07 8 9411 + 08 10581 + 0 7 10738 + 08 10641 + 07 4 375l
- 06 fl - 12 7 M i 6568+06 1083t + 05 3420teot 38161+09 6 7531 + 07 4 5801 08 4 105I + 07 2 2361 09 ft 127 8 0801 +04 3 2938 +03 9 5721 + 04 1 6891 08 0 0001 00 2 0274 09 1 tilt + 04 4 180t*05 ff 129M 1976E 06 2 3128 + 07 4 6021 + 08 3 9664 09 2 5001 + 07 4 7591+08 5 5288 07 15148 + 09 it 129 3296Ee03 3 076f + 04 2 8346 09 0 0001 + 00 0 000t + 00 0 0004 + 00 3 4338 10 3 9161 03 T4 131M 6 2081 + 05 9 459f + 06 2 529t + 07 14471 + 04 18271 15 17361 + 03 3 0361 e 06 3 2481 + 07 131 l 2 3368 + 03 3 450t + 07 2 8798 32 0 0001 00 0 000t 00 0 0001 + 00 35151 13 6 099E 15 ft 132 4 6321 + 05 4 9688 + 06 8 5818 + 07 2 300t + 07 1 3711 01 2 760t ,06 10331 + 07 7 Sist 07 l 130 1 4881 06 6 6921 + 06 1742t + 08 4 0058 04 1 4161 45 48061 05 2 0921 e 08 8276t 01 (P A 5TUR() (PA$ fuel) Ullo) (P A 5 f uRI) (PA5TURE) l Unit 5 -
Inhalation and all totium - mrem /yr per pCum3 Other pathways for all other radionuchde5 -m2
- mrem /yr per pCi/sec ODCM, V C. Summer, SCE&G Revmon 13 (June 1990) 3 0-31
TABLE 3 2 5 (Continued) PATHWAY DOSE FACTORS FOR SECTION 3 2 3 3 (R,) Page 3 of 3 i em.- l cn..u., l .... l nn eu., l a n..u.. l c n.eu., l vmeue. l en..u., { on..ue,
.,- ......-, j ._... j ., o. ,e,e nom . . , j .em_ e i .e m . . e, i . . _ , ,e , e j , , m . e s.
t til 144,48 e07 l 2 0894 + 07 21951 e 11 36458 +09 f 3 266t e 07 4 1751 96 l 2 6141 e 11 l J 1404 e 10
, . y-- +. _ ~~ e I 112 1 Sill e OS 14$21
- 06 2 2421 e01 1 1898 $7 0 0001 e 00 1 6678 Sel10928e01 } 42628e01
-t t 1 13) 29204 *06 29818e06 1 6748 +09 7 2341
- 01 J0966 2) a680t 00 2 0091 e 09 l 4 4874 e og 1 134 3 9521
- 04 % 10$1 e 0% 1 Stil 10 0 000t e 00 0 000t e 00 0 0006 e 00 1915f to )8541 03 i 1)s 6 :Osi + 0s 1 9471 06 , 3 777: + 06 5 96) .is o wn e 00 7 156 16 A uei e 06 5 :32 e 06
+ 4 Cl 134 1 1288 e06 0 0071 e 09 l 2110f e 10 31f 09 7 adil e 09 1 4774
- 08 6 till e 10 t 67i1 e 10 l
C5+136 1 9361 e0% 1 7101 04 1 7591
- 09 l1 '11 e07 4 2491
- 06 4 40$t 06 $ 2921 *09 1 70sl ce j C1 137 84601e05 1201I e 10 1 fall e10 j 9 i, l
- 08 61971 e 09 11%I e 08 4__ ,
f S 1421 e 10 l 13461 e 10 C$ 138 8 $60 e O2 41024 e 06 31491-JJ 0 0001
- 00 0 000t + 10 ) v $76l 2) 6 9)St 11 l 4'O0001*00 ,
BAeik9 6 464l e 03 11941 eOS 7 7411 07 0 0001 e 00 0 0001
- 00 0 0006
- 00 f67941 08 24031 01
.e. q_ _ __. _ _ _ _
l BA 140 J 0321 e 06 2146t
- OF 7 4831 e 0 7 16614 e 07 1 990t 05 4 1961e06 8 981I e 06 2 1304 e08 4
8 A 141 328hl e01 4 7348 + 04 770)t 46 0 000t
- 00 0 0001 00 0 0004 00 9 2448 4? 86998 22 SA 142 1 9121 e0) % 064f + 04 5 010f 80 0 0006 e 00 0 0001
- 00 0 0006
- 00 6 0121 81 % 6111 39 L A - 140 4 872t e OS 2.1801 + 07 2 2918 + 0$ 8 6491 e 02 5 5604 - 12 104)( + 02 2 7454
- 04 $ tool e 07 L A 142 12001 e04 9 1178
- 05 46111 07 0 C,08 e 00 0 000f e 00 0 0001
- 00 $46%I-Ob 2 %29t*00 ct 141 6136L + 0S 15404 + 07 1696f e 07 2 2521 +07 6 7001 oct 2 7038 +06 2 0161
- 06 5 404l eos Cl 14) J 552*OS 26271 *06 1 671( *06 3 69%Ie02 t1101 14 4 4)48 e 01 2 0064 e 06 2 0401
- 07 Ct 144 13361+07 8 0424 07 165St e os 1 0691 08 4 650t
- 07 37066+07 1 9661 07 13261 e to
+
PR 14) 4 till
- 0% 0 000f + 00 9 55)t o ch h 817(
- 07 13?4f + 0J 69606+06 1 1461
- 05 7 2 3104 + 0p PR 144 1752t e 03 21128 + 0) 12J81 51 0 0001 + 00 0 0004 00 0 0004 00 1Alti-$4 30978 26 ND 147 37201 +0S 10098 + 07 71161 + 0% 24%3t*07 85521 02 2 9421 06 e $301
- 04 1424t e en
# 197 1 7651 e05 )7404e06 2 6461 e 06 3 9891 e 00 % $791 22 4 7871 01 31$51 e 0% 7 8194 e Ub <
NP 239 1 3201
- 05 1976t e 06 1060f e 05 1 3871 01 1 0831 07 4 064 f + 02 1 2761 04 2 0978 e07 l (PAlfuRil (P A Uunt ) (illD) (P Altuhl) (P a n f uk t )
( 1 1 l Units - Inhalation and all trittum - mrem /yr per pCi/m3 Other pathways for all other radionuclides m?
- mrem /yr per pCusec l
ODCM, V C Summer SCE &G Reesion 13 (June 1990) 3032 i
T ABLE 3 2-6 PATHWAY DO5E F ACTORS FOR SECTION 3 2 3 3 (R,)') Page 1 of 3
.u .. . l .a -
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'See note, page 3.0-36 Units - Inhalation and all tritium - mremlyr per pCom3 Other pathways for all other radionuclides -m2
- mrem:yr per pCilseC ODCM, V C. Summer, SCE 50 Reosion 13 (June 1990) 3033
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t-f ABLE 3 2 6(continued) W t A/AY DO5E F ACTOR $ FOR SECTION 3 2 3 3 (R,) Page 2 of 3 e m.+ e l . . ~. n l .. . l 4e .: ., , l ..s o. l . . .m o l .m , l ...m. l .. n _, i ... .~. i em ., e ... i e_. ! . . .. . . i m . ., . j ~ .... j ._ . j .s.r . m., 5h 92 4 1041
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- 08 %6914 05 7 4 7 71 e 0 7 % 672t + 05 48141 09
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Units - Inhalation and all tritium mremlyt per pCom3 Other pathways for all other radionuclides -m2
- mremlyr per pCusec ODCM, V.C. Summer. SCE & Ci Reosion 13 (June 1990) 3034
TABLE 3 2 6(continued) PATHWAY DOSE F ACTORS FOR SECTION 3 2 3 3 (R,) Page 3 of 3
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- 07 6 2861 + 02 4 722f e 06 6 2731 04 1 853! 08 W 18 7 1 $521 + 05 2 740( + 06 17968 + 06 5 9121 +00 3 787f 22 7 094t - 01 2 let + 05 1046f . 07 NP 239 11921+05 19761+06 7 4091
- 04 5 1521
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i Units - Inhalation and all tritium rnremlyr per pCi/m3 Other pathways for all other radionuclides m2
- mremlyr per pCuteC ODCM, V.C. Summer, SCE &G Revision 13 (June 1990) 3035
NOTE. The R, values of Table 3 2-2 through 3 2 6 were calculated in accordance with the methods of Section 5 31 of Reference 1. Columns in those tables marked " Pasture" are for freely grazing animal 5 (fp = f, = 1). Columns marked " Feed" are for animals fed solely locally grown stored feed (f,,a f, = 0) The values used for each parameter and the origins of the values are given in Table 3.2-9 and its notes. . 5 i l l l i I l l i, ODCM, V.C. Suinmer, SCE&G. Revision 13 (June 1990) 3036 E
Table 3 2 7 CONTROLLING RECEPTORS, LQCATIONS, AND PATHWAYS * ' DISTANCE AGE ORIGIN SECTOR (METERS) PATHWAY GROUP (FOR INFORM ATION ONLY) N'* 6,100 Vegetation Child Vegetable Garden NNE" 5,300 Vegetation Child -Vegetable Garden NE 4,500 Vegetation Child Vegetable Garden 4,500 Grass / Cow / Meat Child Grazing Beef Cattle ENE 2,600 Vegetation Child -Vegetable Garden 2,600 Grass / Cow / Meat Child Grazing Beef Cattle F 1,800 Vegetation Child Vegetable Garden ESE 1,800 Vegetation Child Vegetable Garden SE 2,400 Vegetation Child Vegetable Garden SSE 4,300 Vegetation Child Vegetable Garden 5** 6,300 Vegetation Child -Vegetable Garden SSW** 5,500 Vegetation Child Vegetable Garden SW" 5,300 Vegetation Child Vegetable Garden WSW 3,100 Grass / Cow / Meat Child Grazing Beef Cattle W 4,300 Vegetation Child Vegetable Garden 3,500 Grass / Cow / Meat Child Grazing Beef Cattle WNW*
- 7,700 Vegetation Child Vegetable Garden NW *
- 6,600 Vegetation Child -Vegetable Garden NNW 4,800 Vegetation Child Vegetable Garden 4,800 Grass / Cow / Meat Child Grazing Beef Cattle See note on the following page for the method used to identify these control-t ling receptors.
If a cow were located at 5.0 miles (8,000 meters) in this sector, an infant consuming only its milk would receive a greater total radiation dose than would the real receptor listed. However, such an infant would not be the Maximum Exposed Individual for the site. ODCM, V. C. Summer, SCE &G: Revision 14 (December 1990) 3.0 37
NOTE: The controlling receptor in each sector was identified in the following way. Receptor locations and associated pathways were obtained from the August 1990 field survey. A child was assumed at each location, except that where a milk cow was listed, an infant was assumed. X/O' for each candidate receptor was obtained by interpolation of values in Table 6.1 10 of Reference 5: 57d' for each ) candidate receptor was obtained by interpolation of values in j Table 6.113 of Reference 5. Expected annual releases of each nuclide were taken from Table 5.2 2 of Reference 5. The pathway dose factors given above in Tables 3.2 3 and 3.2 4 were then used with the referenced values in the methodology of Section 5.3 of Reference 1 to compute total annual doses at each candidate receptor site for the pathways existing at that site. The controlling j receptor for each sector was then chosen as the candidate receptor j with the highest total annual dose of any candidate receptor in the i given sector. Alllisted pathways are in addition to inhalation and ' ground plane exposure, l l ODCM, V. C. Summer, SCE &G: Revision 14 (December 1990) 3.0 38 l
. ~ - .. .- .-__ .. - . . - - _ - . - - . _ . . . _ ~ . . .-- . _ . . . -
, Table 3.2 8 ATMOSPHE RIC Dl5PE R510N PAR AMETE RS F6FC6RTROLLING RTGI4Dh [OCATTdNP DISTANCE SECTOR k"7D' 07D' (MILES / METERS) N 1.5 E-07 7.0 E 10 3.8/6,100 NNE 2.5 E 07 1.1 E 09 3.3 -,3 . NE 3.7 E-07 1.8 E 09 2.8/4,500 ENE 1.1 E 06 5.8 E 09 1.6/2,600 E 2.2 E-06 1.2 E 08 1.1/1,800 ESE 2.2 E 06 8.4 E 09 1.1/1 800 , SE 1.6 E 06 5.8 E 09 1.3/2,400 , SSE 3.0 E 07 1.0 E 09 2.7/4,300 5 1.7 E-07 3.7 E-10 3.9/6,300 SSW 2.0 E-07 6.4 E 10 3.4/5,500 SW 2.6 E 07 1.0 E 09 3.3/5,300 WSW 6.4 E 07 3.2 E-09 1.9/3,100 W 2.2 E 07 9.2 E 10 2.7/4,300 W 3.2 E 07 1.5 E 09 2.2/3,500 ; WNW 5.9 E 08 2.2 E 10 4.8/7.700 NW 9.6 E 08 4.1 E 10 4.1/6,600 NNW 1.8 E-07 9.8 E 10 3.0/4,800 Annual average relative dispersion and deposition values for the receptor locations in Table 3.2 7. Values were obtained by interpolation in Tables 6.1-10 and 6.1-13 of Reference 5. Those tables are based on one year (1975) of meteorological readings and the FSAR dispersion model (ground-level release, sector averaged model, with open terrain recirculation factors, dry depletion by Figure 3.31, and using decay with a half life of 8.0 days). As a result of the
- analysis described in the note to Table 3.2 7, the location of the maximum exposed individual for the site was identified as being the vegetable garden at l
l 1.1 miles in the E sector. Therefore, the site RTQ' and TilD' (Section 3.2.3.2 and following) are those from this table for that location. ODCM, V. C. Summer, SCE &G: Revision 14 (December 1990) 3.0 39
Table 3.2 9 Page 1 of 4 PARAMETER 5_USED IN DOSE F ACTOR _C&CULATIO_NS Origin _ of Value Paramelej Value Table in Section of b't"' R G 1 109 NUREG-gg)-- Specik c
. . . F o r P, * *
- DFA, Each radionuchde E9 Note 2 BR 3700 m /yr 8 E5
"*For Ri(Vegetation)
r Each element trpe E1 Y, 2 0kg/m' E-15 Aw 5.83 E 7 sec S 3.1.3
'DFL, Each age group and radio- E 11 thru Note 2 nuclide E 14 Ua' Each age group E5 i
f, 10 5.3.1 5 t, 8 6 E 4 4 seconds E 15 Uf Each age group E-5 f ,, - 0.76 5,3.1,5 t ,, 518 E + 6 seconds E 15 H 8 84 gm/m 8 Note 1
*"For Ri(Inhalation)*"
BR Each age group E5 DFA Each age group and nuclide E 7 thru Note 2 i E-10 i i l l ODCM, V.C Summer. SCE &G Revmon 13 (June 1990) 3040 l l
n Table 3 2 9 , Page 2 of 4 PARAMETERS USED IN DOSE FACTOR CALCULATIONS Oriain of Value Parameter Value Table in Section of Site. R G 1 109 NUREG- SP'C'I'C 013T
* * *For R Plane)* * *(Ground SF 0.7 E 15 DFG, Each radionuclide E6 t 4.73 E + 8 sec 5.3.1 2
* * *For Ri (Grass / Animal / Meat)* *
- 0, (Cow) 50 kg/ day E3 .;
0, (Goat) 6 kg/ day E3 U,, Each age group E5
-~
Aw 5.73 E-7 sec ' 5 3.1.3 F,(Botl') Each element E1 l r Each element type E 15 DFl, Each age group and nuclide E 11 thru Note 2 E 14 f,, 1.0 Note 3 f, 1.0 Note 3 Y ,, 0.7 kg/m 3 E 15 t, 7.78 E + 6 sec E 15 Y, 2.0 kg/m 2 E 15 l tt 1.73 E + 6 sec E 15 H 8 84 gm/m 3 Note 1 ODCM, V C. Summer, SCE &G Revision 13 (June 1990) 3041
Table 3.2 9 Page 3 of 4 PARAMETERS USED IN DOSE FACTOR (,AL(_ULATIONS O._rfgin of Value Parameter Value Section of Table in NUREG-Site-R G 1 109 gyp Specific
*"For R Note 4 (Grass /Ahimal/ Milk)*'
Or (Cow) 50 kg/ day E3 Or (Goat) 6 kg/ day E3 U ,, Each age group E5 Aw 5 73 E 7 sec ' 5313 F. Each element E1&E2 r Each element type E-15 DFL, Each age group and nuclide E-11 thru E- Note 2 14 Y ,, 0.7 kg/m*' E-15 t, 7 78 E + 6 sec E 15 Y, 2.0 kg/m 2 E 15 t, 1.73 E + 5 sec E 15 f ,, 1.0 Note 5 f, 1.0 Note 5 f, 00 Note 5 f, 0.0 Note 5 H 8.84 gm/m 3 Note 1 ODCM, V C Summer, SCE &G Revmon 13 (aune 1990) 3012
Table 3.2 9(Continued) l Page 4 of 4 NOTES
- 1. Site specific annual average absolute humidity. For each month, an average absolute humidity was calculated from the 7 years of monthly average temperatures in Table 2.3 49 of Reference 4 and the 5 years of monthly average dew points in Table 2.3-64 of Reference 4. The 12 monthly vplues were averaged to obtain the annual average of 8.84 gm/m 3. (Section 5.2.1.3 of Reference 1 giver a default value of 8 gm/m 3.)
- 2. Inhalation and ingt stion dose factors were taken from the indicated source.
For each age group, for each nuclide, the organ dose factor used was the highest dose factor for that nuclide and age group in the referenced table.
- 3. Typically beef cattle are raised all year on pasture. Annualland surveys have indicated that the small number of goats raised within 5 miles typically are used for grass control and not food or milk. Nevertheless, the goats were treated as full meat and milk sources where present, despite the fact that their numbers cannot sustain the meat consumption rates of Table E4 of Reference 3.
- 4. According to the August 1990 land use census, dairy cattle possibly graze at 4.9 miles in the West sector if dairy cattle graze at this location, the dose to an infant consuming milk from these animals would be less than the dose received by the critical receptor identified for the sector. No other milking activity within five miles of the plant was identified. These values are included for reference only.
- 5. *fwo columns of R,'s were calculated one for cows kept exclusively on local pasturo (f, = f, = 1), and one for cows kept exclusively on locally 9to.vn stored feed (f, = f, = 0). See the note on page 2.0 37.
ODCM, V. C. Summer, SCE &G: Revision 14 (December 1990) 3.0 43 l
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3.3 Meteorolocecal Model for Dose Calculations 3 3.1 Meteorological Model Parameter) Section of Term Definition Instial Use b = height of the containment building. (3321) D'
= deposition rate for ground level re- (3.3.2.2) leases relative to the distance f rom the com.ainment building (from Figure 3 3 3). ,
DiQ = the sector averaged annual average (3.322) relative deposition f or any distance in a given sector (m i) i = wind speed class. The wind speed classes are given in Table 4A of Reference 10 as 13,4 7,812, 1318,19 24, and > 24 miles per hour N a total hours of valid meteorological (3.3.2.1) data. n = number of hours rneteorological (3.3.3.1) conditions are observed to be in a given wind direction, wind speed class i, and atmospheric stability class J. n = number of hours wind is in given direction- (33.2.1) l l r = distance from the containment building (33,21) j to the location of interest for dispersion calculations (m) AT/AZ = temperature dif ferential with vertical (3321) separation ('K/100m). T = terrain recirculation factor, Figure (3321) l 3.3 4. u, = wind speed smidpoint of wind speed (3.32.1) class i) at ground level (m/sec). X/O = the highest annual average relative (33.2.1) concentration given sector. (sec/m at anE). distance in a 6 = plume depletion factor at distance r (3.331) from Figure 3.31. ' ODCM, V. C, Summer, SCE &G: Revision 13 (June 1990) 3.0 45 i e - - . . - , , . - - a s ,c.. ,, . - , , , ,,- , - , - - , - , - , ~ . . - -
.tlon of Term Defimtiotj in..eal Use o, a vertical standard deviation of the plume (3.321)
, (in meters), at distance r for ground level releases under the stabihty category indicated by AT/ 32, from Figure 3 3 2 2 032 = (2/n)'i divided by the width in radians of a (3321) 22 5* sector (0 3927 radians) 2 55 = the inverse of the number of radiansin a 22 5" sector (3.3'22) e t (22 5 )(0 OT75 RadianV) 332 Mettorological Model 3321 Atmosphenc dispersion for routine venting or other routine gaseous effluent relcases is calculated using a ground level, wake corrected form of the straight kne flow model X/O = the sector averaged annual average relative concentra-tion at any distance in the given sector (seum') n
= .: 03 6 r ~~V (52) v Nru =r, where:
2.032 = (2/n)' ' divided by the width ei, radians of a 22 V sector (0.3927 radians). Sa plume depletion factor at distance r for the appropriate , stabihty class from igure 3 31. r i = Wind speed class. The wind speed classes are given in Table 4A of Reference 10 a513,4 7,812,1318,19 24, and > 24 miles per hour. n = number of hours meteorological conditions are observed to be in a given wino direction, wind speed class i, and atmospheric stability class J. ODCM, V. C. Summer, SCE &G: Revision 13 (June 1990) 3.0 46
N z total hour 5 of valid meteorological data r e distance from the containment building to location of interest (m) u, = wind speed (midpoint of wind speed c' ass i) at ground level (m/sec) 1, 2 i A .. t, s s, , . .i o,'4 6%nii .., ( v 3 o , ) [53) where. o, = vertical standard deviation of the plume (in meters) at distance r for ground level releases under the stabihty category indicated by AT/ AZ, from Figure 3.3 2. T = terrain recirculation f actor, from Figure 3 3 4 n = 3.1416 b = height of the containment building (50 9m) AT/AZ = temperature differential with vertical separation ("K/100m). 3322 Relative deposition per unit area for all releases is calculated for a ground level release. l D/Q = the sector averaged annual average relative deposition l at any distance in a given sector (m #). l
= 2 55 Dgn (54) rN where.
D, a deposition rate for ground level releases relative to distance (r) from the containment building (from Figure 33-3). ODCM, V. C. Summer, SC E &G : Revision 13 (June 1990) 3.0 47 l
2 55 = the inverse of the number of radiansin a 22 5 sector 1 (22 5')(6T05 Radiansr) n = number of hours wind is in given direction (sector). N = total hours of valid meteorological data. i
's.
ODCM, V. C. Summer, SCE &G Revision 13 (June 1990) 3048 l
i 3 FIGURE 3.31 Plume De pletion Effect for Ground Level Releases (6) (All Atmospheric Stability Classes) Graph taken from Reference 8, Figure 2 m R i g I
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ODCM, V.C. Summer, SCE&G Revision 13 (June 1990) 3.0 49 o
.,,m.__,_.. _ _ . . .__ . _ , . _ . . . . . , . . . _ . . . ~.. . . . _ . , _ _ . , _ .
FIGURE 3,3 2 Vertical Standard Deviation of Materialin a Plume (S1) (letters denote Pasquill Stability Classes) Graph taken from Reference 8 Figure 1 i 1000 , , , l , i / )~ l / /
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0.1 1.0 10 100 PLUME TRAVEL DISTANCE (KILOMETERS) Temperature Change Pasquill Stability with Height AT/AZ (*K/100m) Cateaories Classification
< 1.9 A Extremely Unstable 1.9 to -1.7 B Moderately Unstable 1.7 to 1.5 C Slightly Unstable 1.5 to 0.5 D Neutral 0.5 to 1.5 E Slightly Stable 1.5 to 4.0 F Moderately Stable
> 4.0 G Extremely 5 table ODCM, V.C. Summer, SCESG. Revision 13 (June 1990) l 3.0 50 l
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FIGURE 3.3 3 Relative Deposition for Ground Level Releases (09) (All Atmospheric Stability Classes) Graph taken f rom Reference 8. Figure 6 10-3 10-4 'y i y N '
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FIGURE 3.3 4 Open Terrain Recirculation Factor Graph taken from Reference 7, Figure 2 8
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4.0 R ADIOLOGIC AL ENVIRONMENTAL MONITORING Sampling locations as required in section 1.4.1 of the ODCM Specifi-cations are described in Table 4 0-1 and shown on Figures 4 01 and 4.0 2. As indicated by the ditto (") marks in the table, entries in the sampling frequency and analysis frequency columns apply to all samples below the entry until a new entry appears . ;
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RADIOLOGICAL ENVIRONMt:NTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1
* "' Sampling and Sample' to(ations Type & Frequency Criteria for Selection MUDir of Analysis p, Collection Frequenty Location of Sample Number & Location and/or Sample ===
26 Onsite Gamma isetop+( and tritium A) 2 Indstator samples to be tak en within the Quarterly grab samphng 7 V. Ground 27 Onsste analyses quarterly 7 Water exclusion boundary and in the direction of potentially af f ected ground water supp!**s Quarterly grab samphng 7 16 201 W Gamma isotop4( and tritium B) 1 Control sample from unaffected location analyses quarterly ' I Monthly grab samphng 5 28 2 4 55E Menthly59amma isotop ( VI Drinhng A) I !ndicator sample from a nearby publi( ground and 9:oss beta analyses and Water water supply source quarterly 7(omposite foe tritium analyses Mont hly (omposite 17 2475 Monthly 5 gamma esotop*( B) 1 Indicator (firwshed water) sample from the and gross beta analyses and ! nearest downstream water supply samphng quarterly 7(omposite f of triteum anaf yses 39 14 0 55E Monthly 5gamma isotop.( 1 Controf (finished water) sample from an Monthly compcsste C) and gross beta analyses and unalfe(ted water supply sampling quarterly 7(omposne for tritium analyses ODCM, V.C. Summer, SCEandG: Revision 13 (June 1990) 4.0-6 _ _ _ _ _ - l
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR ST.. TION TABLE 4.0-1 Enposure Criteria for Selection Sampling and Sample 1 Locations Type & Frequency
# *'Y of Sample Number & Location Collection Frequency Location Mi/Dir of Analysis end/or Sample * *NGESTION:
Vll. Milk 4 A) Samples from milking animals in 3 lo(ations Semimontbly when animals i .> i . Gamma isotopic and 1-131 within 5 km having the highest dose potential are on pastures , monthly ~ . analysis semimonthlye when
- if there are none then 1 sample itom rnilkmg other times 5 .. w animals are on pasture, l animals in each of 3 areas between 5 to 8 km ~ . - , monthly other times 5 l distance where doses are cal (ulated 1o be ..~,.u
- greater 1han 1 mrem per yez.r10 B) 1 Control sample to be tak en at the location of Semimonthly when a 'emals 16 201 W Gamma isotopic. and 8 131 l a dairy > 20 miles destance and not in the most are on pastureB, monthly analyses semimonthlyewhen prevalent wind dire (tson / other times 5 animals are on pasture, monthly other timey l
C) 1 Indicator grass (torage) sample to be 1ak en at Monthly when 6 1 0E5E Gamma esotopec l one of the 1, stations beyond but as (lose to the available 5 ex<lusson boundary as pra(ticable wher e the highest of fsste R(torsal ground level (oncentrations are anticipated 2 D) 1 Indicator grass (forage) sample to be tak en at Monthly when 4- % -4 Gamma esotopic the location of Vil( A) above when animals are available 5 . . .mm t on pasture. -..,-o-
+. .n..
. . . . ~ . .
ODCM. V.C. Summer. SCEandG: Revision 13 Oune 1990) 4.0-7
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 E we p Criteria for Selection Sampling and Sample 1 Locations Type & Frequen(y
,gg,3 pg of Sample Number & Lo(ation Collection F requent y to(ation Mi/Dir of Analysis l
E) 1 Control grass (f orage) sample to be tav en at Monthly when 16 201 W Gamma isotop ( thelocation of Vil(B)above. available 5 Vill Food A) 2 samples of beoadleaf vegetation grown en the Moritbly when available 5 6 1 0E5E Gamma tsotop+c on edible Produc ts 2 nearest of fsite locations of highest (altulated 8 15 E rJE portion annual average ground level D/O if milk sampl ng as not performed within 3 6 m or of i milk sampling is not performed at a location within 5-10 km where the doses are (akulated to be greater than 1 mrem /yrl0 B) 1 Conteof sample for the some foods tak en at a Monthly when available 5 18 1655 Gamma isotop.c on edible lo(ation at least 10 miles distance and not en portion the most prevalent wind direction it milk sampling is not periormed within 3 k m or if milk sampling is not periormed at a location within 5-8 6 m where the doses are calculated to be greater than 1 mrem /yr 10 1M Fish A) 1 Inde(ator sample to be tab en at a locat*on en Semtannual9 (ollection of 233 0 3-5 Gammo isotopic on edibte the following spece types ei the upper reservoir portions semaannua!!i* available. bass. tweam. (rappie (atfish. (arp, forage fish (shod) ODCM, V.C. Su mmer, SCEa ndG: Revision 13 (June 1990) 4.0-8
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Sampling and Sampiet Locations Type & F requen(y Criteria for Selection of Analyses a&way Collection Frequency Location Mi/Dir of Semple Number & Location and/or Sample 213 1-3 Gamma esotop,( on ed ble 1 Indicator sample to be taken at a location en Semiannual 9 collection of B) portrons sermannually 9 the iower reservoir the io!!owing spece types if available; bass; bream, (rappie, (atfish,(arp; forage fish (shad) 24i 5 5-6 5 Gamma isotop;( on edible 1 Indicator sample 1o be ia6 en at a location en Semiannual 9 (otlec tion of C) 1he following spece 1ypes if portrons semsannually 9 the upper te ervoir'snon-fluctuating available. bass; bream, reoeational area crappee, (atfish,(arp, forage fish (shad) 223 30 0 farJW Gamma isotopic on edible 1 Controf sample to be tan en at a lo(ation on Semiannual 9 (ollection of D) the following spece typesif portions sermannually 9 the receiving rever suf froently far upstream su(h that no ef f e(ts ol pumped storage available: bass; bream, operation are antiopated (rappee; catfish,(arp, forage fish (shad) AQUATIC: semiannual grab sample 9 233 0 5 ESE Gamma isotop< X Sediment A) 1 Indicator sample io N ta6 en at a location in the upper reservoir 244 5 5 f4 Gamma esotop4( 1 Indicator sample to be tak en at a lo(ation in Semiannual grab sampfe 9 f B) the upper reservoir's non-f!uttuating recreational area ODCM, V.C. Su mmer, SCEandG: Revision 13 (June 1990) 4.0-9
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Criteria for Selection Sampling and Samplet Locations Type & Frequency End/or Sample I SamP! e Number & Lo(ation Collection Frequency to(ation Mi/Dir ( f Analfsis C) 1 Indicator sample to be tak en on the shoreline Semiannual grab sample 9 213 2 7 55W Gamma esotopg of the lower reservoir. D) 1 Controf sample to be tab en at a lo(at.on on - Semiannual grab sample
- 221 30 0 PJrJW Gamma esotop.(
the receiving rever suf ficently far upstrearr such that no ef fects of pumped storage operation are anticipated l l l l ODCM, V.C. Summer, SCEandG: Revision 13 (June 1990) t 4.0-10
RAD!OLOGICAL ENVIRONMENTAL MONITORifJG PROGRAM VIRGIL C. SUMMER IJUCLEAR STATION TABLE 4.0-1 NOTES (1) location numbers refer to Figures 4.0-1 and 4.0-2. (2) Sample site locations are based on the meteorological analysis for the period of record a presented in Chapters 5 and ' 6, V.C. Summer Operating License Environmental Report. (3) Though generalized areas are noted for simplicity of sample site enumeration, airborne, water and sediment sampling is done at the same location whereas biological sampimg si es are gencratized areas m order to reasonably t assure availability of samples. (4) Milking animal and garden survey results will be analyzed annually. Should the survey indicate new dairymg activity the owners shall be contacted with regard to a contract for supplying suffioent samples if contractual arrangements can be made, site (s) will be added for additional milk sampling up to a total of 3 Indicator tocations (S) Not to eneed 35 days. (6) Time composite samples are samples which are collected with equipment capable of collecting an aliquot at teme intervals which are short (e 9. houriy) relative to the compositing period (7) At least once per 100 days. l (8) At least once per 18 days (9) At least once per 200 days (10) The dose shall be calculated for the maumum organ and age group, usmg the guidance / methodology contamed m Regulatory Guide 1.109, Rev.1 and the parameters partuular to the site ODCM, V.C. Su mmer/SCEandG: Revision 13 (June 1990) 4.0-11
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l ODCM, V.' C. Su mmer, SCEand G Revision 13 (June 1990) 4.0 12
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Appendix A Worked Examples of Monitor Setpoint Calculations and Dose Calculations A. RM LS, RM L7 and RM_ L9 Given: V = 5100 gal Nuclide Concentrations: fr = 100 GPM H3 = 2 70E-2 uCi/ml . Fop = 2.1 E6 GPM Mn 54 = t 5' 'i 7 uCi/ml* fe, = 60 GPM Fe-55 = A 3,E-6 uCi/mi Ft = 2.33E6 GPM Fe 59 = 5.38E-7 uCi/ml* Cir/MPC, = 8.73 E-6 Co 58 = 5.83E 7 uCi/ml* itk = 1.25 hr Co 60 = 2.76E-6 uCi/ml* Sr 89 = 6.50E 8 uCi/ml Sr90 = 1.74E-7 uCi/mi Tc-99m = 2.10E-7 uCi/ml* Sb 124 = 5.49E-7 uCi/ml* Sb 125 = 3.24E 6 uCi/ml* l-131 = 3.83E 5 uCi/ml* l-133 = 5.92E 8 uComl* Xe 133 = 1.12 E-4 uCi/ml* Xe 133m = 8.46E 7 uCi/ml* La-140 = 1,77E-7 uCi/ml*
= make up ECg i
- 1) Monitor Setpoint Calculations The method outlined in section 2.1.2 by which these monitor setpoints are calculated is as follows (see Section 2.1.1 for definitions of terms)*
a) The minimum recirculation time shall be: t, = 2V/f, ! = 2 (5100 gal) /100 gal / min l- = 102 min l ODCM, V.C. Summer, SCE&G Revision 13 (june 1990) Appendix A 1 l
t b) The isotopic concentration to be released is obtained from the sum of the measured concentrations. Y
-- a C=T Cg ==
+C+C a s
+ C a
+C
/
i I
= 1.60E 4 uCi/ml + 0 + 2.39E 7 uCi/ml + 2.70E-2 uCi/ml +
4.35E-6 uCi/ml
= 2.71E-2 uCi/ml c) Once isotopic concentrations have been determined, these salues are used to calculate a Dilution Factor, DF.
C DF = T '
, SF
~ SIPC s <
C* C* C' CI C' DF = T , + + , + SF
- StPC SfPC SIPC- SIPC '
s a a SIPC( e
\- 3.59E- 7 5 38E-7 5 38E-7
= + +
IE-4 6E-5 lE-4 2.76E- 6 2.1 E- 7 5.49 E- 7
+ + +
5E-5 6E-3 2E-5 3.24 E- 6 3 83E-5 5 92E-8
+ + + z.
IE-4 3E-7 IE-6 1.12 E- 4 8.46E-7 1.77 E - 7
+ + + )+0 2E-4 2E-4 2E-4 .
65E-8 1.74 E - 7 4.35E-6 27E-2
+( + i+ + + 0.5 l- 3E-6 36-7 8E-4 36-3
= 138 / 0.5 = 276 i
l l ODCM, V.C. Summer, SCE &G: Revision 13 (June 1990) l Appenda A 2
d) The maximum permissible discharge flow rate, fg is now calculated. F +f F f, = -f = f li>rF,,>>f,, 2 LEG gpm + 60 gym 276
= 7600 gym ,
and, C-F = ( 0 9 ) F, # 1 - )
= t 912 33E6 t1 -0 00000873)
= 2 I E6 g p m e) The dilution flow rate setpoint, F,is established at 90 percent of the expected available dilution flow rate:
F = (0.9) F,
= (.9)2.33E6 gpm
= 2.1E6 gpm i
The flow rate monitor setpoint for the effluent stream shall be set at the selected discharge pump rate (normally the maximum discharge ! pump rate or zero). f) The radiation monitor setpoints is now determined based on the values of EC;, F, and f which were specified. The monitor response is primarily to gamma radiation, therefore, the actual setpoint is cased - on EC,. l l-t 1
- ODCM, V.C. Summer, SCE8G Revision 13 (June 1990)
Appendix A 3 l
l i The setpoint concentration, c,is determined as follows: es1c,iA> e A = l~ l l 7600 epm
,~
sogpm *
= 127 If A < 1, No release may be made. Reevaluate the alternatives pre-sented in Step d).
If A il 1, Calculate c and determine the maximum value for the actual monitor setpoint (cpm) from the monitor calibra-tion graph. c s 1 c,tA > = ti coE--4 ucvmin1273 e
= 2.03E-2 uCi/ml c L cpm equivalent to 2.03E-2 uCi/mi
- Reading from Figure 2.1-1 yields:
C h 12,000 cpm g) Within the limits of the conditions stated above, the specific monitor setpoints (in uCi/ml) for the three liquid radiation monitors RM L5, RM-L7, and RM- L9 are determined as follows: RM-L5, Waste Monitor Tank Discharge Line Monitor: csy 1 c, y(A> a RM-L7, Nuclear Blowdown Monitor Tank Discharge Line Monitor: c5 a 1 C, a* ODCM, V.C. Summer, SCE8G Revision 13 (June 1990) Appendu A-4
RM L9, Combined Liquid Waste Processing System and Nuclear Blow-down Waste Effluent Discharge Line Monitor The monitor setpoint on the common line, c c, should be the same as the setpoint for the monitor on the active individual discharge line (i e., cy, or c, as determined above): C(, s AfAX (Cy ,rg i Liquid Radwaste Discharge Via Industrial and Sanitary Waste System (RM-LS) The RM-L5 setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm should an inadvertent high concentration release occur.
- 2) Dose Calculation The dose contribution from all radionuclides identified in liquid effluents released to unrestricted areas is calculated using the expression * -
D=Y A V at, C, F, i e=1 where: f as 60gpm g _ _ 4 (F,n1> 2 33E6 gpm
= 2.6E 5
=
1 I (8 96Hl.25H2 7E-2H276) t #90sul 25n3 59F-7H276) + tl43n t 25i (4 35E-6)(2.6E-5) + (1260Hl.25H5.38E-T H2 6E-5) + (339H125H5 83E-7 n2 6E-5)
+ 1958 H125)(2 76E-6H2 6E-5) + tl370n125 n6 SE-8H2.6E-5) + (288,000n125)
(174E-7H2 6E-5) t (1.06H125H2 IE-7 n2 6/.-51 + (95Hl 25n5.49E-7H2.6E-5i + (36 Sul 256 43 24E-6) 2 6E-5) + 1486HI .!5i>3 53E-SlI2 6E-5i + (89 7nl 25i ODCM, V.C. Summer, SCE 8 G Revision 13 (June 1990) AppendixA 5
. . - - . . . . - _ . - -. - _ ~ . - . - -- - - . . -._..-_v.
(5 92E- sit 2 6E-5) + (0 0476Hl .25)(177E- 7n2 6E-5, i
- = 1.025E 5 mrem (to the Whole Body)
- Dose calculation example was done only for Whole Body, Bone, Liver, Thyroid, Kidney, Lung and GI LLI also must be done to address all dose receptors.
B. RM-L3, RM L8, RM-L10 and RM L11 Normal Mode Given: Nuclide Concentrations; ; fr = 100 GPM H3 = 3.71E 2 uCiimi Fcd = 3.114E5 GPM l-131 = 1.04E-5 uCuml* fs d = 250 GPM l133 = 6.14E 7 uCuml* Fd = 3.46ES GPM Cs 134 = 1.47E 6 uComl* l Ci r/MPCi = 8.54E 6 Cs-137 = 1.73E 6 uComl*
.itk = 2.5 hr
= make up ECg
- 1) Monitor Setpoint Calculations The method outlined in section 2.1.4.1 by which these monitor setpoints are calculated is as follows:
a) Totalisotopic concentration and the Dilution Factor are calculated as in steps a) and b) of example A. 1 C, = [, C, + C, + C, + C, + C7
. a
= 1.42E 5 uCi/mi + 0 + 0 + 3.71E-2 uCi/ml + 0
= 3.71E-2 uCi/mi
.C DF = 5 i 4
+ SF y 1.0 4 E - 5 6 14 E - 7 147K-6 173E-6
- 3.0E-7 1.0E - 6 9OE-6 2OE-5 ODCM, V.C. Summer, SCE&G Reasion 13 (June 1990)
Appendu A 6
- . . . - . . _ _ . . - . - _ . . . . . ~ . . . . ~ . . . - - - . . ~~ - _ . . - _ _ . _ , _.
3 71 E-2
+ 0 + 0 +- 0 + +05
= 4 8/0 5 = 96 b) The maximum permissible effluent discharge flow rate, f,, is now calculated.
F3 , + /j, F 3-I, = s pp = g ihr F,, ? > f s , 3 l14ES GI I + 250 GP$1
= 3,24 5 Gl'3f and, c
F, = to 91 F, I I - , ,, , l
= (0.9)(3.46E5 GPM)(1-8,54E 6)
= 3.11E5 GPM If f,2 f, releases may be made as planned.
c) The dilution flow rate setpoint for minimum flow rate, F, is established at 90 percent of the expected available dilution flow rate: F = (0.9)(F,)
= 3114E5 GPM Flow rate monitor setpoints for the Steam Generator Blowdown effluent stream shall be set at the selected discharge purup rate (normally the maximum discharge pump rate) f,, chosen in Step '3) above.
d) The Steam Generator Monitor setpoints are specified based on the values of S C,, F,and f which were specified. The raonitor response is I primarily to gamma radiation, therefore, the actual setpoint is based ODCM, V.C. Summer, SCE8G. Revision 13 (June 1990) Appendix A-7
on S Cg. The monitor setpoint in cpm which corresponds to the calcu-lated value c is taken from the monitor calibration graph. The setpoint concentration, c,is determined as follows: e s [ C, 4Hi e B = fd, oo _ 3.245 GPAf
~ 250 GPAl
= 13 If B e 1, Calculate c and determine the maximum value for the actual monitor setpoint (cpm) from the monitor talibration graph.
If B < 1,- No release may be made. Reevaluate the alternatives presented in step b). es c,e = n 42E-5 acvm/n13)
= 1.85E-4 uCi/ml c a cpm equivalent to 1.85E-4 uCi/ml
- Reading from Figure 2.1 1 yields:
l C i 105 cpm l l l e) The Turbine Building Sump and Condensate Demineralizer Backwash monitor setpoints are established independently of each other and without crediting dilution. They are based on the measured radio-nuclide concentradons of the effluent stream and are to ensure l compliance with the limits of 10CFR 20, Appendix B, Table 11, Column l 2 prior to discharge. l l For each effluent stream, a concentration factor CF must be l calculated, measuring the nearness 'of approach or the undiluted ODCM, V.C. Summer, SCE&G. Revision 13 (June 1990) Appendix A-8 l
waste stream to the specified limiting condition of the Maximum Permissible Concentration. That is, e' CF % ~ SF
-- \tPc, if CF a 1 calculate c and determine the actual monitor setpoint (cpm) from the calibration curve.
If CF > 1, no release may be made via this path. The release must either be delayed or diverted for additional processing. Because of spurious alarms, these remedial steps may be required if the monitor setpoints are only near the actual concentrations being released. f) Within the limits of the conditions stated above, the specific inonitor setpoints (in uCi/ml) for the two Steam Generator Blowdown monitors RM L3 and RM L10 and the setpoints for RM L8 and RM-L11 are now calculated. Because they are primarily sensitive to gamma radiation, their setpoints will be based on the concentrations of gamma emitting radionuclides as follows: For RM L3, Steam Generator Blowdown Discharge initial monitor, and for RM-L10, Steam Generator Blowdown Discharge final monitor: cy. reg s 1C 3 tlO a For RM L8, Turbine Building Sump Discharge Monitor: c,. s 5 C, ,. - C F , e For RM-L11. Condensate Demineralizer Backwash Discharge Monitor:
\
c.5 f _C o
- CF o ,
l ODCM, V.C. Summer SCESG Revision 13 (June 1990) Appendn A 9
- 2) Dose Calculation The dose contribution from all radionuclides identified in liquid effluents released to unrestricted areas is calculated usmg the expression * -
D=Y .t \ a t , C,, F,
# est where: ,
I, d . 2500 eprn
. (Fa it1) 3 46ES epm
= 7 2E 4 l
=[ 18 96112 SH3 7E-2H7 2E-4) + e486H2 Sul 04K-5H7 2E-41 + 489 7h2 51 i
l
~l (614E-7H7.2E-4) + iS89.000 n2 5HI 47E- 6117 2E-41 + #348,000)
(2.5H173E-6H7 2E-41
= 3.25E 4 mrem (to the Whole Body)
- Dose calculation example was done only for Whole Body, Bone, Liver.
Thyroid, Kidney, Lung and Gl LLI also must be done to address all dose receptors. C. RM A3 and RM-A4 Given: Nuclide Concentrations: X/O = 5.3E 6 sec/m3 Kr-85m = 1.1 E-6 uCi/ml Fv = 481 cc/sec Kr-88 = 3.5E-7 uCi/ml
.itk = 0.6 hr Xe 131m = 3.9E-6 uCi/ml Xe 133 = 8.5E-4 uCi/ml Xe-133m = 1.2E-5 uCi/ml Xe 135 = 5.1E-5 uCi/ml' l131 = 6.73E-8 uCi/ml (0.6 hr)(3600 sec/hr)(481 cc/sec) = 1.04E6 cc Cv = 3.54 cpm (summed Noble Gas concentrations, used Example Noble Gas Calibration Curve, Figure 3.1 1).
ODCM, V.C;5ummer, SCE8G Reviuon 13 (June 1990) Appendu A 10
. _ - . _ - . . - -. . _. . - . . . .. - =- --. . _-. . . - - -_ _-
- 1) Monitor Setpoint Calculations The method outlined in section 3.1.2 by which the Station Vent Noble Gas Monitor setpoints are calculated is as follows:
a) Determine the count rate per mrem /yr to the total body (R t ) R, = C,, ! (Ly s :F 1( E, X i 3 SE4 cpm / (53E-6 U H481 I){ti 17E3HI IE-6)+ tl 47E4 n3 5E-76 3 sw m
+ t915El H3 9E-6)+ 12 94E2 W8 $E-4 i + t2 S I E2H 12E-5;+ il 81 E3)i5.l E- 51)
= 3,5E4 /t5.sE-6 H481 no 3524 l
l l l = 3.9E7 cpm / mrem /yr b) Determine the count rate per mrem /yr to the skin (Rs) R, = Cy / (LQ ( F,. H t I., + 1 1 31, I X,,)
= 35 E4 / 5.3E-6 H481) it2.8E3 M I 1E-6)+ tl 9E4 H3 5E-7)+ <646; (3 9E-6)+ t3.9E2KB 5E-41+ tl 4E3 h l 2E-5s + i4 E3x5 I E- 51)
= 3.5E4 /t5 3E-6 x48180 566i
= 2.4E7 cpm / mrem /yr i
c) Determine the count rate at the alarm setpoint level. This will be less j than or equal to the lesser of: l l ODCM, V.C. Summer, SCESG Rev6 ion 13 (June 1990) l Appendia A 11
(0:25)(Rt )(500 mrem /yr), or (0,25)(R 5)(3000 mrem /yr) - (0.25)(3.9E7)(500) = 4.9E9 cpm (0.25)(2.4E7)(3000) = 1.8E10 cpm so use 4.9E9 cpm d) If two simultaneous releases out of the main plant vent should o'ccur, calculate the setpoint for each type of release and use highest setpoint obtained.
- 2) Dose Calculation a) Unrestricted Area Boundary Dose Rate (Section 3.2.2)
D, = X:q [ K, Q, (mrem lyrs 4
= 5 3E-6 [ (1.2E3)t481 Hl .l E-6) + il 5E4:14 81 )(3.5 E- 7 )
+ (9 2Eli(481 H3.9E-61 + 42 9E2)(48148 SE-4i + (2 5E2H481i (1.2E-5 + d l.8E3H481 H5 I E- 5)
= _ 9E-4 mrem /yr to total body.
D = XiQ (L, + 1.1 M nQlD, = X/Q K, Q, (mrem lyr)
= 5.3E-6 [ 11460 * (1.lH12E3 el 481H1 lE l2370 + tl.Inl.5E4 1 (481 H3.5E-7) + [476 + t i 1 n15F2il(481 H3 9E-6i + 1306 + (1. I) 13 5E2il(481u8 5E-41 + 1994 - il i H3.3E2ilt4811tl 2E-Si +
11860 + (1.lH19E3sl(481 H5 I E-5) l i ODCM, V.C. Summer, SCE8G: Revision 13 (June 1990) ! Appendix A-12 l l-l
= 2.1E 3 mrem /yr to skin.
D, = XIQ P,if,
= (5.3E-6)($.624E + 7)(6.73E 8)(481)
= 2.8E-3 mrem /yr (Orcan Dose Rate) ,
b) Unrestricted Area Dose to Individual (Section 3.2.3) , D = 3.17E- 8 M, fTQ Q
= 3.17E-8 [ ((1.2E3)(5.3E-6)(1.1 E-6)(1.04E6) + (1.5E4)(5.3E-6)(3,5E-7) s (1.04E6) + (1.6E2)(5.3E-6)(3.9E-6)(1.04E6) + (353)(5.3E-6)(8.5E-4)(1.04E6)
+ (327)(5.3E- 6)(1.2E- 5)(1.04E6) + (1.9E3)(5.3E-6)(5.1E- 5)(1.04E6))
= 5.6E-8 mrad y air dose.
Op = 3.17E- 8 N,YIQ Q', l l
= 3.17E-8 [(2.0E3)(5.3E-6)(1.l E- 6)(1.04E6) + (2.9E3)(5.3E-6)(3.5E-7) l l (1.04E6) + (1.1E3)(5.3E-6)(3.9E-6)(1.04E6) + (1.lE3)(5.3E-6)(8.5E-4)(1.04E6) 1
+ (1.5E3)(5.3E-6)(1.2E-5)(1.04E6) + (2.5E3)(5.3E-6)(5.lE-5)(1.04E6))
= 1.82E 7 mrad air dose.
D, = 3.17E- 8 R g Wg Q'
= 3.17E-8 [ [(1.624E7)(2.2E-6)(6.73E-8)(1.04E6) + (2.089E7)(1.2E-8) u ODCM, V. C. Summer, SCE&G: Revision 14 (December 1990)
! Appendix A 13 (
l (6.73E-8)(1.04E6) + (4.754E10)(1.2E-8)(6.73E-8)(1.04E6))
= 1.35E 6 mrem individual dose due to radioiodines and radionuclides in particulate form with 13 > 8 days.
D. RM A10 Given: X/Q = 5.3E-6 sec/m3 Kr-89 = 1E-5 uCi/ml .
- 1) Monitor Setpoint Calculation Permissible release conditions for the Waste Gas System are defined in terms of both radionuclide concentration and waste gas flow rate (using previous nuclide concentrations).
a) The maximum permissible flow rate is set on the same basis but include the engineering safety factor of 0.5. The RM-A10 setpoint level 5,is defined as: So<.1.Sc b) The maximum permissible waste gas flow rate f, (cc/sec) is calculated from the maximum permissible dose rates at the site boundary according to: f,5 the lesser of f, or f, f, = the maximum permissible discharge rate based on total body dose rate. 0.25 X 500 mremlyr G7Q X 1.5) (Xad UI
= (0.25)(500) / 5.3E-6)(1.5((1E-5)(1.66E4))
= 9.47E7 cc/sec ODCM, V. C. Summer, SCE&G: Revision 14 (December 1990)
Appendix A 14
l f, = the maximum permissible discharge rate based on skin dose rate. 0 25 X 3000 mremSr (L Q Hl.5) Xgi t, + 1.1 M,)
= (0.25)(3000) / 5.3E-6)(1.5((1E-5)(101E4 + 1.1(1.73E4)))
= 3.24E8 cc/sec so f u __- 9.47E7 ccsec c) When a discharge is to be conducted, valve HCV-014 is to be opened until:
(a) the waste gas discharge flow rate reaches (0.9)(f ) or (b) the count rate of the plant vent noble gas monitor RM A3 approaches its setpoint, whichever is reached first. E. Alternative Methodology for Establishing Conservative Setpoints (using previous nuclide concentrations) A more conservative setpoint is calculated to minimize requirements for adjustment of the monitor as follows:
- 1. For a plant vent:
R ,' = conservative count rate per mremlyr to the total body (Xe-133 detection, Kr-89 dose).
= C ' + [(X/Q) (K,, g) (X,') (F,)],
Note: C,'is based on tha pven Kr-89 concentration being applied to the Example Noble Gas Monitor Calibration Curve, Figure 3.1-1.
= 3.3E4 cpm / ((5.3E-6)(16,600)(8.5E-4)(481)
= 9.2E5 cpm R ,' = count rate per mrem /yr to the skin.
ODCM, V.C. Summer, SCE & G Reeon 13 (June 1990) j Appendix A 15
= C,* + [(X/0) ((L,, ,, + 1.1 M,, g)) (X,') (F,)]
= 3.3E4 cpm /((5.3E 6)(29,130)(8 SE-4)(481)
= 5.2E5 cpm
- 2. For the waste gas decay system:
f,' = the conservative maximum permissible discharge rate based on Kr 89 total body dose rate. ,
= (0.2 5) (D,,) + [(X/Q) (1.5) (X,')(K,, g)J
=
(0.25)(500)/(5.3E-6)(1.5)(918E 4)(1.7E4)
= 101E6 cc/sec f,' = the conservative maximum permissible discharge rate based on Kr 89 skin dose rate.
= (0 2 5) (D ss ) + [(X/0) (1.5) (X,') (L,, g + 1,1 M,, g))
! = (0.25)(3000)/(5.3E-6)(1.5)(9.18E 4)(2.9E4)
= 3.54E6 cc/sec i
l l l l l l l ODCM, V.C. Summer, SCE&G Revision 13 (June 1990) Appendix A-16 _ -. . _ _ _}}