ML19253B971
| ML19253B971 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/23/1974 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19253B968 | List: |
| References | |
| NUDOCS 7911010759 | |
| Download: ML19253B971 (24) | |
Text
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PROPOSED TECHNICAL SPECIFICATIONS FOR THREE MILE ISLAND, UNIT 1 Docket Number 50-289 2.4 LIMITING CONDITIONS FOR OPERATION Radioactive Discharges Objective: To define the limits and conditions for the controlled release of radioactive materials in liquid and gaseous effluents to the environs to ensure that these releases are as low as practicable.
These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural back-ground exposures.
The release rate for all effluent discharges shall be within the limits specified in 10 CFR Part 20.
To assure that the releases of radioactive material above background to unrestricted areas be as low as practicable as defined in Appendix I to 10 CFR Part 50, the following design objectives apply:
For liquid wastes:
a.
The annual dose above background to the total body or any organ of an individual from all reactors at a site should not exceed 5 mrem in an unrestricted area.
b.
The annual total quantity of radioactive materials in liquid waste, excluding tritium and dissolved gases, discharged from each reactor should not exceed 5 C1.'
1504 173 For gaseous wastes:
c.
The annual total quantity of noble gases above background discharged from the plant should result in an air dose due to gamma radiation of less than 10 mrad, and an air dose due to 7911010 7 @
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. beta radiation of less than 20 mrad, at anf location near ground level 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.
c.
The annual total quantity of iodine-131 discharged from each reactor at a site should not exceed 1 C1.
I j
2.4.1 Specifications for Liquid Waste Discharges a.
The concentration of radioactive materials released in liquid wastes from all reacto:s at the site shall not exceed the values specified in 10 CFR Part 20, Appendix B, Table II, Column 2, e
for unrestricted areas.
b.
The release rate of radioactive materials in liquid wastes, excluding tritium and dissolved gases, shall not exceed 10 Ci/ reactor / calendar quarter.
c.
The release rate of radioactive materials in liquid wastes, excluding tritium and dissolved gases, shall not exceed 20 Ci/ reactor in any 12 consecutive months.
d.
During release of radioactive wastes, the effluent control monitor shall be set to alarm and to intiate the automatic closure of the waste discharge valve prior to exceeding the limits specified in 2.4.1.a above.
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, c. The operability of the automatic isolation valves in the liquid discharge lines shall be demonstrated quarterly.
f.
The equipment installed in the liquid radioactive waste system shall be maintained and shall be operated to process radioactive liquid wastes prior to their discharge when the projected cumulative release rate will exceed 1.25 Ci/ reactor / 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 Ci, excluding tritium and dissolved gases, h.
When the release rate of radioactive materials in liquid wastes, excludies tritium and dissolved gases, exceeds 2.5 Ci/ca.lendar quarter, the licensee shall make an investigation to identify the causes for such release rates, define and initiate a program of act;on to reduce such release rates to the design objective levels listed in Section 2.4, and report these actions to the Commission within 30 days from the end of the quarter during which the release occurred.
2.4.2 Specifications for Liould Waste Sampling and Monitoring Plant records shall be maintained of the radioactive concentration a.
and volume before dilution of liquid waste intended for discharge and the average dilution flow and length of time over which each discharge occurred.
Plant records shall be submitted in accordance with Section 5.6.1 of these specifications.
Estimates of the error associated with each reported value shall be included.
1504 175
- T
, b.
Prior to release of each batch of liquid waste, a sample shall be taken from that bateit and analyzed for the concentration of each significant gamma energy peak in accordance with Table 2.4-1 to demonstrate compliance with Specification 2.4.1 using the flow rate into w!'ch the waste is discharged during the period of discharge, c.
Sampling and analysis of liquid radioactive waste shall be performed in accordance with Table 2.4-1.
Prior to taking samples from a monitoring tank, at least two tank volumes shall be recirculated, d.
The radioactivity in liquid wastes shall be continuously monitored and recorded during release. Whenever these monitors are inoperable 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 samples of each tank to be discharged shall be analyzed and two plant personnel shall independently check valving prior to the discharge.
If these monitors are inoperable for a period exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, no liquid waste tank shall be releaced and any release in progress shall be terminated.
e.
The flow rate of liquid radioactive waste shall be measured and recorded during release, f.
All liquid radiation monitors shall be calibrated at least quarterly by means of a radioactive source which has been calibre:ed to a National Bureau of Standards source.
Each monitor shalt also have a functional test nonthly and an instrument check prior to making a release.
1504 176
,T
- J Bases:
The. release of radioactive materials in liquid waste to unrestric " 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 reasonable 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 flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20.
It is expected that by using this operational flexibility under unusual operation conditions, and exerting every effort to keep levels of radioactive material in liquid wastes as low as practicable, 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 based on operating experience taking into._ account a combination of variables including 1504 177
3
. defective fuel, primary system leakage, primary to secondary system leakage and the performance of the various vaste treatment systems, and are consistent with Appendix I to 10 CFR Part 50.
Specification 2.4.1.a requires the licensee to limit the concen-tration of radioactive materials in liquid wastes from the site to levels 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 permissible w.?er the Commission's Rules and Regulations using the guidelines given in Regulatory Guide 1.21.
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 flexibility of operation to assure that the public is provided a dependable source of power under unusual operacing conditions which may temporarily result in releases higher than the levels normally achievable when the plant and the liquid waste treatment systems are functionin,.s designed. Releases of up te these limits will result in concentrations of radioactive material in liquid waster at small percentages of the limits specified in 10 CFR Part 20.
Specifications 2.4.1.d and 2.4.1.e reauire that auftable aculprent to control and monitor the releases of radioactive materials in liquid 1504 178
O, s vastes are operating during any period these releases are taking place consistent with the requirements of 10 CFR Part 50, Appendix A, Design Criterion 64.
Specification 2.4.1.f requires that the licensee maintain and operate the equipeent installed in the liquid waste systems to reduce the release of radioactive materials in liquid effluents to as low as practicable consistent with the requirements of 10 CFR Part 50.36a. Normal use and maintenance of installed equipment in the liquid waste system provides reasonable assurance that the quantity released will not exceed the design objective.
In order to keep releases of radioactive materials as low as practicable, the specification requires operation of equipment whenever it appears that the projected cumulative discharge rate will 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 will not exceed the Techaical 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 exceeds one-half the design objective annual quantity during any calendar quarter and describe the proposed program of 1.504 179
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- action to reduce such release rate to design objective levels on a timely basis. This report must be filed within 30 days following the calsndar quarter in which the release occurred.
The sampling and monitoring requirements given under Specification 2.4.2 provide assurance that radioactive materials in liquid wastes are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64. These requirements provide the data for the licensee and the Commission to evaluate the plant's performance relative to radioactive liquid wastes released to the environment.
Reports on the quantitia.s of radioactive materials released in liquid wastes are furnished to the Commission according to Section 5.6.1 of these Technical Specificatiens in conformance with Regulatory Guide 1.24.
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 paints to be monitored in Section 2.4.2 include all.the monitored release points as provided for in the Final Safety Analysis Report.
2.4.3 Specifications for Gaseous Waste Discharges a.
(1)
The release rate limit of noble gases shall be:
1504 180
+-~is-4
. IQ
[49.6 5
+ 94.35 g] 11 gy where ( = release rate from the unit vent in Ci/sec (ground release) i = the individual nuclide EY = the average gamma energy per disintegration 58 = the average beta energy per disintegrat..on Refer to Table 2.4-3 for E and E values to be y
g used.
a.
(2) The release rate limit of all radiciodines and radioactive materials in particulate fr_m with half-lives greater than eight days, released to the environs as part of the gaseous wastes shall be:
6 2.33 x 10 g iy y
where ( = release rate from the unit vent in C1/sec (ground release) b.
(1) The average release rate of noble gases during any calendar quarter shall be:
IQ
[310 E
+
295 E ) 1 1 1y gg b.
(2) The average release rate of noble gases during any 12 consecutive months shall be:
IQ
[6205
+ 589 E ] 11 gy 1
gg 1504 181
'<w' b.
(3) The average release rate of all iodines and radioactive materials in particulate form with hali-lives greater than eight days during any calendar quarter shall be:
7 2.92 x 10 Q 5,1 y
b.
(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:
7 5.83 x 10 Q 5,1 y
b.
(5)
The amount of iodine-131 released during any calendar quarter shall not exceed 2 Ci.
b.
(6)
The amount of iodine-131 released during any period of 12 consecutive months shall not exceed 4 C1.
Should the conditions of 2.4.3.c(1), (2), or (3) listed below c.
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 Ce mission within 30 days from the end of the quarter during whier the releases occurred.
c.
(1)
If the average release rate of noble gases during any calendar quarter is:
IQ
[1240 E
+ 1178 E} ] >l 4
g}
gy f *.
,3
^
c.
(2)
If the average release rate of all iodines and radioactive materials in particulate form with half-lives greater than eight days during any calendar quarter is:
8 1.17 x 10 q 31 y
c.
(3)
If the amount of iodine-131 released during any calendar quarter is greater than 0.5 C1.
d.
During the release of gaseous wastes from the primary system waste gas holdup system the effluent monitor shall be operating and set to alarm and to initiate the automatically closure of the waste gas discharge valve prior to exceeding the limits specified in 2.4.3.a above. The operability of the automatic isolation valves shall be demonstrated quarterly.
e.
The maximum activity to be contained in one waste gas storage tank shall not exceed 90,000 curies (considered as Xe-133).
2.4.4 Specifications for Gaseous Waste Sampling and Monitoring a.
Plant records shall be maintained and reports of the sampling and analysis results shall be submitted in accordance with Section 5.6.1 of these Specifications.
Estimates of the error associated with each reported value should be included.
b.
Gaseous releases to the environment, except from the turbine building ventilation exhaust and as noted in Specification 2.4.4.c, shall be continuously monitored for gross radioactivity and the 1504 183
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flow measured and recorded.
Whenever these nonitors are inoperable, grab samples 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.
During the release of gaseous wastes from the prinary system waste gas holdup system, the gross activity monitor, the iodine collection device, and the particulate collection device shall be operating.
d.
All waste gas.)nitors shall 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 shall have a functional test at least monthly and instrument check at least daily, e.
Sampling and analysis of radioactive material in gaseous waste, particulate form, and radiciodine shall be performed in accordance with Table 2.4-2.
Bases:
The release of radioactive materials in gaseous wastes to unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part 20, and in accordance with the requirements of 10 CFR Part 50.36a.
1504 184
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These specifications provide reasonable assurance that the resulting annual air dose due to gamma radiation will not exceed 10 mrad, and an 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 individual from iodines and particulates will not exceed 15 mrem. At the same time these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided with a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20.
It is expected that using this operational flexibility under unusual operating conditions, and by exerting every effort to keep levels of radioactive material in gaseous vastes as low as practicable, the annual releases will not exceed a small fraction of the concen-tration limits specified in 10 CFR Part 20.
These efforts should include consideration of meteorological conditions during releases.
There is a reduction factor of 243 by which the maximum permissible concentration of radioactive iodine in air should be reduced to allow for the grass-cow-milk pathway.
(The factor is 1220 for the grass-goat-milk pathway).
This factor has been derived for radioactive iodine, taking into account the milk pathway.
It has been applied to radionuclides of iodine and to all radionuclides in particulate form with a half-life greater than eight days.
The factor is not appropriate for iodine where milk is not a pathway of 1504 185
,s s exposure or for the other radionuclides.
The design objectives have been developed based on operating experience taking into account a combination of system variables including defective fuel, primary system leakage, primary to secondary system leakage, and the performance of the various waste treatment systems.
For Specification 2.4.3.a(1) dose calculations have been made for the critical sector.
These calculations consider site meteorology, buoyancy characteristics, and radionuclide content of the effluent of each unit. Meteorological calculations for offsite locations were performed, and the most critical one was selected to set the release rate.
The controlling distance is 755 meters to the ESE.
The gamma dose contribution was determined using the equation 7.63 in Section 7-5.2.5 of Meteorology and Atomic Energy - 1968.
The releases from vents are considered to be ground level releases which could result in a beta dose from bloud submersion. The beta dose contribution was determined using Equation 7.21, ar described in Section 7-4.1 of Meteorology and Atomic Energy - 1968.
The beta dose contribution was determined on the basis of an infinite cloud passage with semi-infinite geometry for a ground level release (submersion dose).
ihe beta and gamma components of the gross radio-activity in gaseous effluents were combined to determine the allowable continuous release rate. Based on these calculations, a continuous 1504 186
~
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release rate of gross radioactivity in the amoun: specified
- i. 2.4.3.a(1) will not result in offsite annual doses above background in excess of the limits specified in 10 CFR Part 20.
The average gamma and bete 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 gamma disintegration for those radioisotopes determined to be present from the isotopic analyses are given in Table 2.4-3.
Where isotopes tre identified that are not listed in Table 2.4-3, 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 Negatron Emitters (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 eight days. The calculations consider site meteorology for these releases.
For radiciodines and radioactive materials in particulate form, the controlling sector for unit vent releases is the ESE Sector at a distance of 755 meters (X/Q = 6.5 x 10~
sec/m ) for the dose due to inhalation. The nearest milk cow is located in the ESE sector at 1504 187
,m
/ a distance of 2410 meters. The applicable X/Q at the nearest 3
milk cow is 9.6 x 10- sec/m.
The grass-cow-milk-child thyroid chain is controlling.
The assumptions used for these calculations are:
(1) onsite meteorological data for the most critical 22.5 degree sector; (2) credit for building wake; and (3) a reconcentration factor 243 was applied for possible ecological chain effects from radioactive iodine and particulate releases.
Specification 2.4.3.b establishec 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 quantity during any calendar quarter, or four times the design objective annual quantity during any period of 12 consecutive months.
The intent of this specification is to permit the licensee the flexibility of operation to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in higher releases than the objectives.
In addition to the limiting conditions for operation of Specifications 2.4.3.a and 2.4.3.b, the reporting requirements of 2.4.3.c delineate that the cuase be identified whenever the release of gaseous effluents e ceeds one-half the design objective annual quantity during any x
calendar quarter and describe the proposed program of action to reduce such release rates to the design objectives.
1504 188
s
. Specification 2.4.3.d requires that suitable equipment to monitor and control the radioactive gaseous releases are operating during any period these releases are taking place.
Specification 2.4.3.e limits the maximum offsite dose above background to below the limits of 10 CFR Part 20, postulating that the rupture of a waste gas storage tank holding the maximum activity releases all of the contents to the atmosphere.
The sampling and monitoring requirements given under Specification 2.4.4 provide assurance that radioactive materials released in gaseous wastec are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64.
These requirements provide the data for the licensee and the Commission to evaluate the plant's performance relative to radioactive wastes released to the environment.
Reports on the quantities of radioactive materials 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 Guide 1.21.
On the '.. sis of such reports and any additional information on 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 the Final Safety Analysis Report.
1504 189
~
s
. Specification 2.4.4.b excludes monitoring the turbine building ventilation exhaust since this release is expected to be a negligible release point. Many PWR reactors do not have turbine building enclosures. To be consistent in this requirement for all PWR reactors, the monitoring of gaseous releases from turbine buildings is not required.
2.4.5 Specifications for Solid Waste Handling and Disposal a.
Measurements shall be made to determine or estimate the total curie quantity and major radionuclide composition of all radio-actfve solid waste shipped offsite, b.
Solid wastes in storage and preparatory to shipment shall be monitored and packaged to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 171-178.
c.
Reports of the radioactive solid waste shipments, volume s, major radionuclides, and total curie quantity, shall be submitted in accordance with Section 5.6.1.
Bases:
The requirements for solid radioactive waste handling and disposal given under Specification 2.4.5 provide assurance that solid radioactive materials stored at the plant and shipped offsite are packaged in conformance with 10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 171-178.
These requirements provide ti.e data for the licensee and the Commission to evaluate the handlin; and storage facilities for solid radwaste, and to evaluate the environmental impact of offsite shipment and storage.
Reports on the quantities, principle 1504 190
~
m
. isotopes and volumes of the shipments, are furnished to the Commission according to Section 5.6.1 of these Technical Specifications.
On the basis of such reports and any
' ' '.t ional information the Commission may obtain from the licensee or others, the Commission may from tiLie to time require the licensee to take such action as the Commission deems appropriate.
1504 191
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. Table 2.4-1 (Continued)
NOTES:
(1)
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged from the plant.
(2)
For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations.
Under these circumstances, it will be more appropriate to calculate the concentrations of such radio-nuclides using measured ratios with those radionuclides which are routinely identified and measured.
(3)
The detectability limits for activity analysis are based on the technical feasibilir and on the potentisi 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.
(4)
The power level and cleanup or purification flow rate at the sample time shall also be reported.
1504 193
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a oa a
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g p
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nc n
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u l u T
i n a
u y
ya y
yc rc C
l e T
P l
lS y
l y
li ei A
pu h
h l
h l
ht t t O
mq h
h t
t s k
t k
t r rr I
ae c
c n
na e
n e
na aa D
S r a
a o
oG e
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oP uP A
F E
E M
M(
W M
W M(
Q(
R r
e o
s t
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g j
e y
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R a
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s P
r l
e e
i a
D s
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t a
n n
s e
e r
e a
l m
es m
s G
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se n
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i ns a s oc e
a ea t t er t
k t
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n n
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a o
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T C
CR EP A
B C
D
T
. 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 ILmits, 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 when other uuclides are present in the sample at much higher levels.
Under these circumstances, it will be more appropriate to calculate the levels of such radionuclides using observed ratios with those radionuclides which are measurable.
(4)
When the average daily gross radioactivity release rate exceeds that given in 2.4.3.c(1) or when the steady state gross radio-activity release rate increases by 50% cuer the previous curresponding power level steady state release rate, the iodine and particulate collection device shall be removed and analyzed to determine the change in iodine-131 and particulate release rate. The analysis shall be done 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 sampling frequency.
(5)
Monthly and quarterly composites are the sum of all samples taken during the period.
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N..,.
Tab 1Le 2,4-3
/NERAGE ENERGY PER DISINTEGRATION (3)
Isotope
_Ey, mev/ dis (Ref.)
_E8, mev/ dis (Ref.)
Kr-83m 0.00248 (1) 0.0371 (1)
Kr-85 0.0022 (1) 0.250 (1)
Kr-85m 0.159 (1) 0.253 (1)
Kr-87 0.933 (1) 1.32 (1)
Kr-88 2.18 (1) 0.377 (1)
Kr-89 1.95 (2) 1.58 (2)
Kr-90 2.10 (2) 1.01 (2)
Xe-131m 0.0201 (1) 0.143 (1)
Xe-133 0.0454 (1) 0.193 (1)
Xe-133m 0.042 (1) 0.19 (1)
Xe-135 0.247 (1) 0.317 (1)
Xe-135m 0.432 (1) 0.095 (1)
Xe-137 0.194 (1) 1.64 (1)
Xe-138 1.18 (1) 0.611 (1)
(1) ORNL-4923, Radioactive Atoms - Supplement I, M.S. Martin, November 1973.
(2) NED0-12037, Summarv of Gama and Beta Emitters and Intensity Data.
M.E. Mack, R.S. Gilbert, Janaary 1970.
(The average 8 energy was ccm-puted from the maximum energy using the International Comittee on Radiological Protection II equation, not the 1/3 value assumption used in this reference).
(3) The average S energy includes conversion electro;.s.
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