ML20128C701
| ML20128C701 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/23/1974 |
| From: | Benaroya V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Muller D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9212040511 | |
| Download: ML20128C701 (29) | |
Text
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Y' boeket No. 50-263 Daniel Muller, Assistant Director for Environ:nental Projects, L PRVIROTENTAL TECIINICAL SPECIFICATIONS FOR }DUTICELLO MUCLEAR CENITJ. TING TLANT Plant Names Monticello, Unit 1 ,
Licensing Stage: OL
' )
Docket N aber 50-253 ' , . t
. Responsible Branch EPB $4 _
Ecrtueeted Cort)etion Datet Description of Responnet Environwrotal Technical Specifications ;
nevtcu Statunt Couplete T.nclosed are the Environmental Technical Speciftentione for the radioactive vaste treatrent and nonitoring nectionn for l'anticello i 4
Nucient Generating Plant. Unit 1. , ,
b VictorDenarhya, Chief Effluent; Treatment Systems Branch Directorate of Liccueing ,
Enclosure . ,
1 Ao stated ' N DISTRIBETION:
Docket (50-263) <
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PROPOSED TEC11MICAL SPECIFICATI0::S FOR
- MONTICELLO NUCLEAR CENERATI!:G PLANT
, DOCKET h*1DBER 50-263 t-j 2.4 LntITING CO::DITIO :S FOR OPERATION i Radioactive Discharnes_
f Objective: To define the limits and conditions for the controlled
- release gf radioactive materials in liquid and gascous effluents to the envjrons to ensure that these releases are as low as practicabic.
These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural background exposures. ,
The release rate for all effluent discharges shall be within the
, limits specified in 10 CFR Part 20. ..
To ensure that the releases of radioactive natorial above background to unrestricted areas will be as low as practicable as defined in Appendix I to 10 CFR Part 50, the following design objectivcs apply:
For liquid wastes:
- a. The an,nual dbsc above background to the total body or any organ of an individual from all reactors at a site should not exceed 5 mica
'in an unrestricted area. t
- b. The annual total quantity of radioactive materials in liquid -
, vaste, excluding tritium and dissolved gases, discharged from' each reactor should not exceed 5 Ci. .
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e mme
, s
- 2 '-
s For gaseous vastes: .
- c. The annual total quantity of noble gases above background
,- discharged from the site should result in an air dose due to gamma radiation of less than 10 mrad, and an air dose due to i
l beta radiation of less than 20 mrad, at any location near ground 1evel which could be occupied by individuals at or beyond the boundary of the site. ,
- d. The annual total quantity of all radiciodines and radioactive material in particulate forms above background from all reactors '
at a site should not result in, an annual dose to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 mrem. -
- e. The annual total quantity of iodine-131 discharged from each
\ reactor at a site should not exceed 1 Ci.
2.4.1 Specifications for Liquid Maste Discharnes
- n. The concentration of radioactive materials released in liquid wastes from all reactors at the site shall not exceed the values specified in 10 CFR Part 20, Appendix B. Table II, Column 2, for' unrestricted areas.
- b. The release rate of radioactive materials in liquid vastes, excluding tritium and dissolved gases, shall not exceed 10 ,
C1/ reactor / calendar quarter.
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- mm -
.mm m -
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- c. The release rate of radioactive materials in liquid wastes, i 1
excluding tritium and dissolved gases, shall not execed 20 Ci/ reactor in any 12 consecutive months. I
- d. During release of radioactive wastes, the effluent control monitor shall be set to alarm and to initiate the automatic closure of the waste discharge valve prior to exceeding the limits specified in 2.4.1.a above. -
- e. The, operability of the automatic isolation valves in the liquid discharge line shall be demonstrated quarterly.
- f. The equipment installed in the' liquid radioactive vaste system shall be maintained and shall be operated to process radioactive liquid wastes prior to their discharge when the projected cumulative e
release rate will exceed 1.25 Ci/rcactor/ calendar quarter. -
excluding tritium and dissolved gases.
- g. The maximum radioactivity to be contained in any liquid radwaste tank that can be discharged directly to the environs shall not exceed 10 C1, excluding tritium and dissolved gases.
- h. When the release rate of radioactive materials in liquid wastes.,
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excluding tritium and dissolved
- gases, exceeds 2.5 Ci/ reactor / calendar quarter, the licensee shall make an investigation to identify the causes of such release rates, define and initiate a program .
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'of action to reduce such release rates to the design objective levels listed in Section 2.4, and report these actions to the
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. ,- Consission within 30 days from the end of the quarter during which the release occurred.
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2.4.2 Specifications for Liouid Unste Sampling and Monitoring
- a. . Plant records shall be maintained of the radioactive con-centration and volume before dilution of 11guid vaste intended .
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, for discharge, and the average dilution flow and length of
- time over which each discharge occurred. Reports and sample analyses results shall be submitted in accordance with
- Section 5.6.1 of these specliications. Estimates of the total error associated with each reported value shall be 4
- included .
- b. Prior to release of each batch of liquid waste, a sample shall be i
taken from that batch and analyzed for the concentration of each e
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, .5-j significant garma energy peak in acco'rdance with Table 2.4-1 to -
g debonstrate conpliance with Specification 2.4.1 using the flow i
rate of the stream into which the waste is discharged during the n
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l i period of discharge. ,
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- c. Samplin'g and analysis of liquid radioactive vaste shall be 9
j performed in accordance with Tabic 2.'4-1. Prior to taking ,
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samples from a monitoring tank, at'least two tank volumes shall 4
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- be recirculated. ' -
j . d. The radioactivity in liquid wastes shall be continuously mdaitored and. recorded during release. ,henever W these monitors are inoperabic for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, two independent f samples of each tank to be discharged shall be analyzed and two -
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plant personnel shall independently check valving prior to the -
j \ discharge. If these monitors are inoperabic for a period exceeding i
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, no release from a liquid waste tank shall be made and.
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j any release in progress shall be terminated. ,
- e. 'The flow rate of liquid radioactive vaste shall be measured and recorded during release.
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< . f. All liquid radwaste effluent radiation monitors shall be calibrated h at least quarterly by means of a' radioactive source which has been
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' calibrated to a National Bureau of Standards source. Each monitor '
shall also have a functional-test monthly and an instrument chech .
l prior to making a release. _ _ _
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Bases: The release of radioactive materials in liquid waste to -
unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part 20 and should be as low as practicable in accordance with the requirements of 10 CFR Part 50.36a. These specifications provide reasonabic assurance that the resulting annual exposure to the total body or any organ of an individual in an unrestricted area will not exceed 5 mrem. At the same time, these specifications permit the ficxibility of operation, compatibic with considerations of, health and safety, to assure. that the public is provided a dependable nource -
of power under unusual operating c,onditions which may temporarily result in releases higher than the design objective icvels but still within the concentration limits specified in 10 CFR Part 20. It is expected that by using this operational ficxibility under unusual .
Y operation conditions, and exerting every effort to keep icvels of radioactive material in liquid vastes as low as practicabic, the annual releases will not exceed a small fraction of the concentration limits .
specified in 10 CFR Part 20.
The design objectives have been developed boced on operating experience taking into account a combination of variables including defective fuel, primary system leakage, and the performance of the various waste treatment systems, and are consistent with Appendix I .
to 10 CFR Part 50.
I:
- 7- .
Specification 2.4.1.a requires the licensee to limit the concentration of radioactive materials in liquid wastes from the site to 1cvels specified in 10 CFR Part 20, Appendix B. Table II, Column 2, for unrestricted areas. This specification provides assurance that no
- member of the general public will be exposed to liquid containing radioactive materials in excess of limits considered permissibic under the Commission's Rules and Regulations using the guidelines given in
Specifications 2.4.1.b and 2.4.1.c establish the upper limits for the release of radioactive materials in liquid effluents. The intent of these Specifications is to permit the_ licensee the ficxibility of operation to assure that the public is provided a dependabic source g of power under unusual operating conditions which may temporarily result in releases higher than the levels normhlly achievable when the plant and the liquid waste treatment systems are functioning as designed. Releases of up to these limits will result in concentrations of radioactive material in liquid wastes at small percentages of the limits ,specified in l'O CFR Part 20. .
Specifications 2.4.1.d and 2.4.1.e require that suitable equipment to control and monitor the releases of radioactive materials in liquid wastes are operating during any period these releases are taking place consistent with the requirements of 10 CFR Part 50, Appendix A.
Design Criterion 64.
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Specification 2.4.1.f requires that the licensee maintain and operate i the equipment installed in the liquid vaste systems to reduce the
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release of radioactive materials in liquid effluents to as low as i
- practicabic consistent with the requirements of 10 CFR Part 50.36a.
Normal use and maintenance of installed equipment in the liquid i vaste syster. provides reasonabic assurance that the quantity released j will not exceed the design objective. In order to keep releases of radioactive materials as low as practicable, the specification-requires, as a minimum, operation of equipment whenever it appears that the projected cumulative discharge rate vill exceed one-fourth of this design objective annual quantity during any calendar quarter.
Specification 2.4.1.g limits the amount of radioactivity that may be
( inadvertently released to the environment to an amount that vill not exceed the Technical Specification limit.
In addition to limiting. conditions for operation listed under Specification 2.4.1.b and 2.4.1.c the reporting requirements of Specification 2.4.1.h delineate that the licensee shall identify the cause whenever the release rate of radioactive materials in liquid wastes exceed one-half the design objective annual quantity during
- any calendar quarter' and describe the proposed program of action to ,
reduce such release rate to design objective icycls on a ticely basis.
This report must be filed within 30 days following the calendar quarter in which the release occurred.
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The ' sampling and monitoring requirceents given under Specification 2.4.'2 provide assurance that radioactive materials in liquid vastes are properly controlled and monitored in conformance with the require-ments of Design Criteria 60 and 64. These requirements provide the data for the licensec and the Commission to evaluate the plant's performance relative to radioactive. liquid vastes released to the I
environment. Reports on the quantitics of radioactive materials '
released in liquid wastes are furnished to the Comission according
. to Section 5.(s.1 of these Technical Specifications in conformance with Regulatory Guide 1.21. On the basis of such reports and any additional information the C'ommission may obtain from the licensee
. or others, the Commission may f rom time-to time require the licensee v to take such action as the Conmission deems approprint
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The c'nv.ironmental release points to be monitored in Section 2.4.2 include all the monitored relcase points as provided for in. Tabic t-2.4.3.
k l 2.4.3 Specif.$ cations for caseous vaste Discharpes
(
- a. ,
(1)' The release rate limit of nobic cases shall be:
Qs 5E+y 2E B
+Q v 28 E'y + 64 E 6 <{ .
. L-m -
- where Q, = release rate from main stach in C1/sec (elevated release) f i"-.-.-- . . . . . _ .. . . . , . - __ _ . , , _ .
~ni f
t 10 ,
4 Qy a release rate from vents in Ci/sec (ground relcane) l .
-n..- _ , . , _ _ _ _
i = the individual nuclido, n = total,nuclides.
E = the avercge ga m a energy per disintegration Eg = the average beta energy per disintegration
! Refer to Tabic 2.4-5 for EY and ESvalues to be used.
'. ...3 (2) The release rato limit of all radioiodines and radioactive
. materials in particulate form with half-lives greater than eight days, released to the environs as part of the cascous vastes shall be: )
. 1
. 4.9 x 10 4 Q s
+ 2.7x 10 6 q gi y_
where Q, = release rate from the main stack in ci/sec (as k ,
cicvated release)
Q y = relcase rato from the vents in Ci/sec (ground release) , ,
- v. b. (1) The average release rate of nobic gases during any calcudar
, quarter shall be .
Qs 31 E + 6.2 E +0 180 E + 200E < *1 i Y 8 7 Y S i+n (2) The average release rate of nobic gases during any 12 consecutive conths shall be:
l 1 . . . .
Q s 63 Ey + 13 E g +Qy 350 E7-1400 E <{
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l (3) The average release rate of all' iodine,s and radioactive 4 materials in particulate form with half-lives ~ greater than eight days during any calendar quarter shall be 6.1 x 10 5 g,+ 3,4x 10 7gy s1 ,
(4) The average release rate of all iodines and radioactive '
' materials in particulate form with half-lives greater 'than eight days during any period of 12 consecutive months shall be: .
- 1. 2 x 10 6 q + 6. 7 x 10 7 Q sl s y f.
i -
, , (5) The amount of iodine-131 released during any caletidar quarter chall not c::cced 2 C1.
(6) The acount of iodine-131 released during any period of 12 consecutive months shall not exceed 4 Ci.
- c. Should the conditions of 2.4.3.c(1), (2) or (3) listed below exist, the licensee shall make an investigation to identify the causes of the release rates, define and initiate'a program of action _to reduce the release rates to design objective levels listed in
- Section 2.4 and report these actions to the Commission within 30 days from t;he end of the quarter during which the releases occurred.
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- 12 i (1)' If the average release rate of nobic gases during any calendar i -
-quarter is: .
4 -
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-Q s
125 Ey + 25 E S
+Qv 700 E + 800 E y 6 5{
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(2) If.;the average release rate of all iodines and radioactive
- , materials in particulate form with half-lives greater than 4 ,
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- eight days during any calendar quarter is
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2.5 x 10 E s+'1.4x 10 Q, >l
-(3) If heamountof' iodine-135re1casedduringanycalendar ~
c y . ..
t quarter is greater than 0.'5 C1. ,
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/ 1'he offgas monitor shaLL be operating and set to alarm and to 5
initiate the automatic closure of the waste gas discharge valve
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prf c r to execeding the limits specified in2.4.3.a(1) above.The operabilit: of tha automatic isolation valves'shall be demon-9 strated quarterly.
- c. i 12 the offgas monitor'is not operating, a shutdown shall be-initiated so that :the . reactor will be in the hot shutdown .
condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
- f. If the release rate of noble gases from the main condenser-vacuum system is:
+
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4 13 . .
-- -+- .
f.
. . .. . - ~
l .'. Q, 63 Eg+13 E -
g 1+n for a period of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, notify the Conmission in writing within 10 days, identifying the causes of activity. The
. report should include the flow rate of the offgas *~ m the main O * '
condenser vacuum system, and the' activity measured downstream of i .
the main condenser vacuum system' prior to holdup, and at a point i
upstr.eam of the point of release.
! g. The reactor containment shall be purged through the standby gas treatment system. -
j
- h. At least two hydrogen monitors in the offgas line downstream of the recombiners shall be operable during power operation. If the hydrogen concentration reaches an alarm set point of four .
\
percent by volume, the of fgas flow shall be stopped by closing .
the valves downstream of the recombiners. Whenever either.of these hydrogen monitors are not inoperable during power operatien, a program shall be initiated to. bring the activity releases within two percent of the limits specified in 2.4.3a(1), and grab sdeples shall be taken and Analyzed for hydrogen concentration each snift. Calibration-of the monitoring system shall be-performed weekly.
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!. .e 2.4.4 Specifications for Caseous Was'te Samoling and Monitorine *
- a. Plant records shall be maintained and reports of tae campling and analysis results shall be submitted in accordance with Section
'} 5.6.1 of these Specifications. Estimates of the total error associated with each reported value should be included.
- b. Gaseous releases to the environment, except as noi;d in Specifi-cation 2.4.4.c below, shall be continuously monitored for gross
., radioactivity and the flow measured and recorded. -Whenever.these monitors are inoperabic, grab sampics shall be taken and analyzed daily for gross radioactivity. If these monitors are inoperable for more than seven days, these releases shall be terminated.
- c. An isotopic analysis shall be made of_a representative sample of 4
gase.ous activity, excluding tritium, at the discharge'of the steam jet air ejectors and 'at a point prior to dilution and
- discharge.
(1) within one month of initial criticality, (2) at least monthly there,after, (3) follouing each refueling' outage, (4) . 'f the gaseous vaste monitors indicate an increase of greater than 50% in the steady state fission gas release after factoring out increases due to power changes.
- d. All vaste gas monitors stall be calibrated at least quarterly by means of a known radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shc11 have a functional test at least monthly ar.d ca instrunant chech at least daily.
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- e. Sampling and analysis of radioactive material in gaseous vaste,
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particulate form, and radioiod'ine shall be performed in accordance I
' with ' Table 2.4-2. ' ,
! l Bases: The release of radioactive materials in gaseous wastes to i !
t i unrestricted 4
areas shall not exceed the concentration limits ll >
specified in 10 CFR Part 20, and shoul'd be as low as practicable 4
! in accordrnce with the requirements of 10 CFR Part 50.36. These d
specifications provide reasonable assurance that the resulting annual
)
I air- dose due to ga=na radiation will not exceed 10 mrad, and- an-i annual air dose due to beta radiation will not exceed 20 mrad
! from noble gases, and that the annual dose to any organ of an J .
, individual from iodines and particulates will not exceed 15 mrem.
1 .
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At the sane time these specifications permit the flexibility of
- l operation, compatible with considerations of health and safety, to
- assure that the public is provided with a dependable source of-power I -
i
- under unusual operating conditions which may ' temporarily resu'lt in i
releases higher than the design objective levels but still within l
the concentration -limits specified in 10 CFR Part 20. It is expected I
that using this operational flexibility under unusual. operating conditions, . and by exerting every effort to keep levels of radio-e 1 active material in gaseous wastes'as. low as practicable, the annual '
releases will not exceed a small fraction of the concentration-limits specified in 10 CFR Part 20. These efforts should include t -
consideration of meteorological conditions during releases.
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1 There is a reduction factor of 243 by which the maximum permissible- i concentration of radioactive iodine in air should be reduced to allow i
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{ for the- grass-cow-milk pathway. , , , _
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This factor has been derived for radioactive j
i
! iodine, taking into account the milk pathway. It has been applied i
t I, to radionuclides of iodine and to all radionuclides-in particulate i 6 l
form with a half-life greater than eight days. - The factor is not
. 1 . .
appropriate for iodine where milk is not a pathway of exposure or
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for the other radionuclides.
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The design objectives have been developed based on operating experience taking into account a combination of system variables q including defcetive fuel, primary system leakage, and the perfor- ,
mance,of the yarious vaste treatment systems.
i '
j For Specification 2.4.3.a(1) -dose calculations have been made for-
- the critical sector. These calculations consider site meteorol.ogy, buoyancy characteristics, and radionuclide. content of the effluent
- of each unit. Meteorological calculations for offsite locations
' vere performed, and the most critical one was selected to set-the release rate. The controlling distance is 756 meters to the
. south-southeast.
- i i -
- The gamma dose contribution was determined using the equation 7.63 in Section 7-5.'2.5 of -Meteorolonv and Atomic Energy ---1968. The-releases from vents are considered-to be ground icycl releases which
- could result _in a beta doge from clo'ud submersion._ The beta dose- .
= _ . -
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m - A E e l .
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contribution was determined using Equation 7.21, as described in '
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Section 7-4.1 of Meteorology and Atomic Enerry - 1968.
The beta 4;
dose contribution was determined on the basis of an infinite cloud t
passage with semi-infinite geometry for a ground level release (submersion dose). The beta and gamma components of the gross e,
radioactivity in gaseous effluents were combined to determine the allowable continuous release rate. Based on these calculations, a continuous release rate of gross radioactivity in the amount specified in 2.4.3.a(1) will not result in offsite annual doses above back-ground in excess of the limits specified in 10 CFR Part 20.
The average gamma and beta energy per disintegration used in the equation of specification 2.4.3.a(1) will be based on the' average composition of gases determined from the plant vent and ventilation exhausts. The average energy per beta or ga==a disintegration.for those radioisotopes determined to be present from the isotopic analyses are given in Table 2.4-5. Where isotopes are identified that are not listed in Tabic 2,4-5 the gamma energy are determined from Table of Isotopes, C. M. Lederer, J. M.-Hollander, and I. Perlman, Sixth Edition, 1967 and the beta energy shall be as given in USNRDL-TR-802, II. Spectra of Individual Necatron Ecitters (Beta Spectra), O. Hogan, P. E. Zigman, and J. L. Mackin. .
For Specification.2.4.3.a(2), dose calculations have been made for the critical sectors and critical pathways for all radiciodines and
. radioactive material in particulate form with half-lives greater than .
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- 'em o e.no 4 w-u ee e e u% =g % . e em w += ==-ae es. eeg - ep - ee.eg-a e _ _e-***
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eight days. The calculations consider site meteorology for these releases.
For radioiodines and radioactive material in particulate form with l
half-lives greater than eight days, the critical location'for
! ground releases is the SSE sector at a distance of 756 meters
~0 sec/m3 for the dose, due to inhalation.
where the X/Q is 4.4 x 10 i The critical location for elevated releases is the SSE sector at a distance of 756 meters where the X/Q is 1.2 x 10-7 sec/m3 for
- the doseg due to inhalation. The nearest milk cow is located
, in the NW sector at a distance of 2410 meters where the X/Q is 1.1 x'10-6 sec/m3 for ground releases, and 2.0 x 10-8 3cc/m3
+
i for elevated releases. The grass-cow-milk-child thyroid chain I
is controlling.
The assumptions used for these calculations are: (1) onsite meteorological data for the most critical 22.5 degree sector; l
(2) credit for building wake; and (3) a reconcentration factor of 243 was applied for possible ecological chain effects from
'" radioactive iodine and particulate releases where applicable.
Specification 2.4.3.b establishes upper limits for the releases of noble gases, iodines and particulates with half-lives greater than eight days,_ and iodine-131 at_ twice the design objective annual e
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- quantity during any calendar quarter, or four times the design l ;_
objective annual quantity during any period of,12 consecutive months, f
The intent of this specification is to permit the licensee the
[
f fiexibility of operation to assure that the public is provided a L depend 6'le source of power under unusual operating conditions which _
i i
- may' temporarily result in higher releases tha. the objectives. ,
In addition to the limiting conditions for operation of Specifica- ,
tions 2.4.3.a and 2.4.3.b, the' reporting requirements of 2.4.3.'c
~
j . delineate that the cause be identified whenever the release of
\'. . gaseouseffluentsexceedsone-halfthedesignobjectiveannuak-i f ,
quantitp during any calendar quarter, and describe the N eposed
)
- \ program of action to reduce such release rates to the design-I -
objectives.
l -
Specification 2.4.3.d and 2.4.3.e are in accordance with Design
+ ,
! Criterion 64.
- j .
I Specification 2.'4.3.f'is to monitor the performance of the core. A t,
i sudden increase in the activity l'vels e of gaseous releases may be
. the result'of defective fuel. Since core performance is of utmost-i' -
importance in the resulting doses from accidents, a report must be
? filed.within 10 days'following the specified increase in gaseous radioactive releases. .
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Specification 2.4.3.g requires that the primary containment atmosphere receive treatment for the recoval of gaseous iodine and particulates prior to its release.
Specification 2.4.3.h requires that hydrogen concentration in the system shall be monitored at all times. .
~ '
The sampling and monitoring requirements given under Specification 2.4.4 provide assurance that radioactive materiala released in gaseous vastes are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64. These require-ments provide the data for the licensee and the Co==ission to evaluate the plant's performance relative to radioactive vastes released to the environment. Repor s on the quantities of radio- -
, \
active caterials released in gaseous effluents are furnished to the Commission on the basis of Section 5.6.1 of these Technical Specifications and in conformance with Regulatory cuide 1.21. On the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Commission may from time -
. to time require the licensee to take such action as the Commission
\
deems appropriate.
The environmental release points to be monitored in Section 2.4.4
, include all the monitored release points as.provided for in Table 2;4-4.
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~l 2.4.5 Specifications for Solid Waste Handlink and Disposal i ,
- s. Measurements shall be made to determine or estimate the total g curie quantity and principle radionuclide composition of all radioactive solid vaste shipped offsite, b '. Solid vastes in storage and' preparatory to shipment shall bc
, monitored and packaged to assure-compliance with 10 CFR Part 20,'
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- 10 CFR Part_71, and 49-CFR Parts 171-178. .
1 -
1
- c. Reports of the radioactive solid vaste shipments, volumes,'
s i principle radionuclides, and total curie quantity, shall be submitted in accordance with Section 5.6.1.
Bases: The requirements for solid radioactivc waste handling and i
disposal given under Specification 2.4.5 provide assurance that solid g
radioactive materials stored at the plant and shipped offsite are proper'ly control'1cd, monitored, and packaged in conformance with -
i 10 CFR Part 20, 10 0FR Part 71, and 49 CFR Parts 171-178. These h requirements provide the data for the licensee and the Commission to evaluate the handling and storage facilities-for_ solid radwaste,
', and to evaluate the environmental impact of offsite shipment and storage. Reports on the quantitics and amounts of the radionuclides,
~
and volumes-of the shipments, shall: be furnished to the Commission
.accordingtoSection556.1oftheseTechnicalSpecifications. 'On the basis of such reports and any- additional information the Commission-f may obtain from.t'hc licensco or others, the Commission may from time to timo require the licensee to take such action as the Commission
' deems appropriate.. * '
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- s Tabic 2.4-1
. . RADIOACTIVE LIQUID SAMPLING AND ANALYSIS
- 4 j Liquid Sampling Type of Detectable =
Source Frequency Activity Analysis Concentra ion?'
l (uC1/ml) 3) 4
'A. Monitor Tank Releases Each Raleave -Individual' Camma Emitters 5 x 10-7 (2).
s One Batch /ffonth Dissolved Cases 10 ,
- Weekly Composite ( } Ba-La-140, I-131 10-6 . .
-8 Sr-89 5 x 10 Monthly C mposite( 10 j 11- 3 1*
~
Gross a 10
-8
' Quarterly Composite ( Sr-90 5 x 10 Weekly (4) -6 -
B. Prima-y Coolant I- ;1, I-133 10
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. , Table 2.4-1 (Continued)
NOTES:
8 (1) A composite sample is one in which the quantit, of liquid sampled is proportional to the quantity of liquid waste discharged from the plant..
(2) For certain mixtures of gat =a emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sat:ple in much greater concentrations. Under these circumstances, it will be more appro-priate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured. .
(3) The detectability limits for activity analysis are based on technical feasibility and on the potential significance in the l environment of the quantities released. For some nuclides, lower i
detection limits may be readily achievable and when nuclides are measured below the stated limits, they should also be reported.
(4) 'Ine,pouer level and cleanup or purification flow rate at the sa=ple time shall also be reported. ,
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Table 2.4-2 P_XIOACTIVE CASEOUS WASTE SAME T'G AND ANA' YSIS Gaseous Sumpling - -
Type of Detectable Source Frcquency Activity Analysis- ,Concentra pns (uC1/ml)
A. Containment Purge Releases Each Reiease Individual Camma Emitters 10- (3)
H-3 -6 ,
10 B. Environmental Release Points Monthly Individual Camma Emitters 10~ )(
, (Cas Sampics)
-6 i
Weekly (Charcoal Sample) I-131 10
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Monthly (Charcoal Sample) I-133. I-135 10
-10 Weekly (Particulates) Individual Gamma Emitters (at least for Ba-La-140, I-131 10-11 (4)
Monthly. Composite Sr-89 1011 (Particulates)
Cross a -11 10 Quarterly Composite Sr-90 10
-11 (Par,ticulates) ' ~
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- 4 Table 2.4-2 (Continued)
NOTES:
(1). The above detectability limits for activity analysis are based on technical feasibility and on the potential significance in the
- environment of the quantities released. For some nuclides, lower detection limits may be readily achievable and when nuclides are measured below the stated limits, they should also be reported.
(2) Analyses shall also be performed following each refueling, startup or similar operational occurrence which could alter the mixture of radionuclides. .
(3) For certain mixtures of gamma emitters, it may not be possible to ,
measure radionuclides at levels near their sensitivity limits Uhen other nuclides are present in the sample at much higher levels. .
Under these circumstances, it will be more appropriate to caletlate the levels of such radionuclides using observed ratios with those radionuclides which are measurable.
(4) When the sverage daily gross radioactivity release rate exceeds that
'Ei ven in .4.3.c. (1) or where the steady state gross radioactivity .
release rare ine cases by 50% over the previous corresponding power icvel steady state release rate, the iodine and particulate collection
\ device shall be removed cnd analyzed to determine the change in ,
,.. iodine-131 and part?culate release rate. The analysis shall be donc daily following such change until it is shown.that a pattern exists which can be used to predict the release rate; after which it may revert to weekly nampling frequency.
(5) To be representative of the average quantities and concentrations of radioactive materials in particulate form released in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent stream or in proportion to the volume of each batch of efflucnt releases. Prior to analyses, all samples taken for the composite should be thoroughly mixed in order for the composite sample to be representative of the average effluent release.
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-l Table 2.4-3 .
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BWR-Liq'uid Waste System i l Location of Process and Effluent Monitors and Samplers Required By Technical Specifications
~
Measurement' -
G b j >
Continuous Sample Gross ' Dissolved Isotop:
- Process Stream or Release Point Monitor Station Activity. I Gases Alpha, H-3 Analys
- '
- 1tigh Purity Waste Sample.. (Test) Tank X X X X X X i i '
Floor Drain Waste Sample (Test) Tank X X .
X X. X- X-
> Chemical Waste Sample (Test) Tank- X X X F 1X i X X ;
- Detergent Waste Collector Tank X X X X X X
. t Primary Coolant System X X ,
s Liquid Radwas'te Discharge Pipe " X X
^
-[
4
- .. Service . Water Discharge Pipe . X X f Outdoor Storage Tanks - Dikes or Retention -i Ponds X X -
X i .
j ' Emergency' Core Ccoling System X. X, . ?
Nuclear Closed Cooliqg System X' X '
I ;
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- In some BWR's the 111gh Purity Waste System may not have's vaste sample (test) tank. The processed liqu'id l l will be routed directly to the condensate storage tank or to -the floor drain waste sample -(test) tank. !
- In most 'BWR's the contents of the detergent waste collector tank re:e sempted, analyzed and then filtered prior to -
release through the liquid'radwaste' discharge pipe. The detergent waste system must' be designed with either a ,
j split ' tank or 'two separate collection or sample- (test) tanks to p,c cmit isolation of .the tanks for mixing, sampling and analysis prior to release.
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, .g Table 2.4-4 BWR-Caseous Waste System-location of Process and Effluent Monitors and Samplers Required B'y Technical Specifications _,
Crab "***"#***"
Continuous Sample .
Process Stream or Release Point Monitor Station NG I Part H-3 .
~
Condenser / Air Ejector (downstream) X X X-Main Stock X X X X X X Building Ventilation Systems Reactor Containment Building. X X X X X X
~
- Radwaste Euilding X X X X X X
- Turbine Building X X -X X X X
- Fuel llandling & Storage Building X X X X X X
- Auxiliary Building X X X X X X 8 N
X X X X X X - N
- Mechanical Vacuum Pump ' s
- Turbine Gland Sedl Condenser -X X X X X X i
- If any or all of the building ventilation systems are routed to a single release point the need for a -
continuous monitor at the individual building exhaust discharge point to the main exhaust duct is eliminated.
One continuous monitor at the final release poi'nt is sufficient.
- Normally the offgases from the mechanical vacuum pump will be discharged downstream of the turbine gland seal condenser vent and the need for individual monitors on each systen. is eliminated. One continuous monitor at Cie final release point is sufficient.
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i Table 2.4-5 i
[ AVERAGE ENERGY PER DISINTEGRATION
- i. .
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Isotope -E , mev/ dis (Ref) E , mov/ dis (3) (Ref) g i , Kr 83m O.00248 (1) . O.0371 (1) i i
Kr-85 0.0022 (1) -
0.250 -
(1)
- Kr-85m 0.159 (1-) 0.253 (1)~
! Kr-87 0.793 (1) 1.32 (1) i '
[ Kr-88 1. 9' 5 ' (1) -O.377, . (1) ,
} Kr-89 2.22 (2). 1.37 (2) l -
j- Kr-90 2.10 (2) 1.01 (2)-
i .
l Xe--131m 0.0201 (1) ,
0.143 (1) i j Xe-133 - 0,0454 (1) 0.135- '(1)
\, Xe-133m O.042 (1) 0.19- (1)
- i
' Xe-135 0.247 (1) . O.317 (1)
! Xe-135m- O.432 (1) O.095" (1) l- Xe-137 0.194 (1)- - 1 . 6 41 (1)
Xe-138 1.18 (1) .O.611 (1)
(1) ORNL-492}, Radioactive Atoms - Supplement I, M. S. Martin, November' 1973. ,
(2) NEDO-12037, ." Summary. of Gar.ma and Bata Emitters and Intensity Dcta; ..-
-H. E. Meek, R. S. Gilbert, January 1970. (The average S cnergy was computed from the maximum energy using the ICRP II equation, not q the 1/3 value assumption used-in this reference.)'
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(3) The average-B energy includes conversion electrons. I a ,
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- M"- M9fu$ e embe*P e-es+4 . e g. g - - e
- es M , he e g aw - gn.g e -- a summe . S.< -m+w e + w mustwe M ' * * ' " ' #
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