ML20054E751
| ML20054E751 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/07/1982 |
| From: | Lahti G COMMONWEALTH EDISON CO., SARGENT & LUNDY, INC. |
| To: | |
| Shared Package | |
| ML20054E730 | List: |
| References | |
| NUDOCS 8206140140 | |
| Download: ML20054E751 (30) | |
Text
a .
UNITED STATES OF AMERICA Os NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In The Matter of )
)
)
COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL
) 50-455 OL
)
(Byron Nuclear Power Station, )
Units 1 & 2) )
AFFIDAVIT OF CERALD P. LAHTI The attached questions and answers constitute my testimony in the above-captioned proceeding. The testimony is true and accurate to the best of my knowledge, information and belief.
Gerald P. Lahti Subscribed and swggn to before me this / day of June, 1982 l
,uv . mum Notary Public My commission expires January 14, 1983 O
B206140140 B20607 PDR ADOCK 05000454 G PDR
BYRON STATION TESTIMONY OF GERALD P. LAHTI Q. State your name and present occupation.
A. My name is Gerald P. Lahti. I am Assistant Division i
Head of the Nuclear Safeguards and Licensing Division in charge of Shielding and Radiological Safety at Sargent &
Lundy in Chicago, Illinois.
Q. Briefly state your educational and professional qualifi-cations.
A. I received a BSCE in Civil Engineering from Wayne State University of 1959. I rects .ved a MSE (Nuclear Engineering) from the University of Michigan in 1960, and completed additional part time course work in Mechanical and Nuclear Engineering at the University of Delaware, Case Western Reserve University, and the University of Toledo. From 1960 to 1963 I was employed by E. I. duPont deNemours & Co., Inc., and mathematically analyzed and designed polymer transfer systems and extrusion dies.
From 1963 to 1973 I was a member of the National Aeronautics and Space Administration (NASA) staff at Lewis Research Center, Cleveland, Ohio. There, I evaluated radiation hazards and designed radiation shields for nuclear reactors j considered for power or propulsion systems in space vehicles.
In 1968 I assumed supervisory responsibilities in this area.
In 1973 I joined Sargent & Lundy and have been employed in the Shielding and Radiological Safety Section continuously since that time.
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The Shielding and' Radiological Safety Section designs and
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evaluates all radiation shielding and other radiation protection features incorporated in nuclear power plant design. I am also responsible for assessing the radiological kmpact of radio-nuclides released during normal and abnormal power plant operations.
I am a Registered Professional Engineer in the State of Illinois and a member of the American Nuclear Society and Health Physics Society. I am a past Chairman of the ANS's Radiation Protection and Shielding Division.
Q. What have been your duties and responsibilities with respect to Commonwealth Edison Company's Byron Nuclear Station?
A. I have supervised the in-plant radiation shielding designed for the Byron Station. In addition, the offsite radiological dose assessments made for radioactivity expected to be released in effluents under normal conditions (FSAR, Chapter 11; ER, Chapter
- 5) and under accident conditions (FSAR, Chapter 15; ER, Chapter 7) were made under my supervision.
I have also testified before the ASLB on matters relating to the impact of radioactivity in Byron Station effluents.
This testimony was offered on August 26, 1975 in support of the construction permit (CP) application. The testimony analyzed I
expected radioactive emissions from the Byron Station and con-cluded that the Byron station design would meet Appendix I to 10CFR50 design objectives.
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O. To which contention is this testimony addressed?
() A. To portions of Contention 2. In that contention intervenors claim that because residents of the DeKalb-Sycamore and Rockford areas are subject to radioactive emission ~ from nuclear power stations other than nearby Byron l consideration should be given to the cumulative impact of projected routine releases
- from those other plants and Byron for this area. The re-evaluation should be performed to ensure applicable limits (e . g . , 10CFR part 20, 10CFR part 50 Appendix I, and 40CFR part 190) are not exceeded in this area.
Q. What regulatory criteria did you consider in analyzing the cumulative routine releases from Byron and other nuclear power plants in Illinois?
A. There are two. Amendments to 10CFR, part 50, Appendix I which sets forth the NRC's decision regarding numerical guides for design objectives and limiting conditions for operation to meet the criterion "as low as practicable"* for radioactive materials in effluents from light-water-cooled nuclear power reactors, were published by the Commission on May 5, 1975 (40FR 19439) and became effective on June 4, 1975.
The-EPA regulation 40CFR part 190 states that the total annual dose commitment to any member of the public
! in the general environment from all uranium fuel cycle sources i
(except those.specifically excluded by the regulation 1 is limited
- "as low as practicable" was changed to "as low as reasonably
(~} achievable," December 19, 1975 (40FR58847) effective
(> January 19, 1976.
to 25 millirems to any organ but the thyroid, and 75 millirems
() to the thyroid.
Q. Briefly desertbc the degign objectkyc4 get (ogth in Appendix I to 10 CFR 50 which you considered.
A. The limiting design objective was that the total quantity of radioactive material released from each reactor to the atmos-phere will not result in an estimated annual external dose to any individual in an unrestricted area in excess of 5 millirems to the total body.
Q. Are there further design objectives for airborne emissions which you considered in analyzing the cumulative effect of routine releases from Byron Station and other nuclear power plants in Illinois?
A. Yes. On September 4, 1975 (40FR 40816), Appendix I was.
amended to allow construction permit applicants, whose applications were docketed on or after January 2, 1971 and prior to June 4, 1976, the option of dispensing with the cost-benefit analysis required by Paragraph II. D of Appendix I, if the proposed radwaste systems and equipment satisfied the " Guides on Design Objectives" proposed by the regulatory staff in the rulemaking proceeding on Appendix I (Docket RM-50-2, February 20, 1974) Commonwealth Edison has elected to dispense with the cost-benefit analysis described in the Annex to Appendix I and demonstrated that O
Byron's proposed radwaste systems and equipment satisfied
{} the " Guides on Design Objectives" referrred to above.
Two design objectives of this rulemaking, included in the Annex to 10CFR part 50, Appendix I, are that 1) the total quantity of radioactive material released from all nuclear reactors at the site will not result in a calculated annual air dose at or beyond the site boundary in excess of 10 millirads; or 2) releases will not result in doses to individuals in the unrestricted area in excess of 5 millirems to the total body. Thus, the most limiting design objective for Byron Station is that releases not result to doses to individuals in the unrestricted area in excess of 5 millirems to the total body.
Q. Define the term " unrestricted area" as used in your prior answer A. Unrestricted area is a term defined in 10CFR 20.3 (a) (17) ; it is any arca access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation. The restricted area boundary for the Byron station is described in the FSAR subsection 2.1.1.3 and shown in figure 2.1-6.
Q. Describe the relationship between Appendix I and 40CFR Part 190.
A. Appendix I to 10CFR50 (and its Annex) describes the regulations applicable to each indivlaual-nuclear power plant's routine O
emissions; 40CFR190 considers the cumulative effects, not only
(_) of all power plant operations, but also of other aspects of nuclear fuel cycle operation.
Q. Why did you not consider 10CFR Part 20 in your analysis?
A. 10 CFR part 20 limits to persons in the unrestricted area correspond to 500 mrem / year to the whole body or its equivalent.
(See 10CFR20.105). Meeting the more stringent Appendix I limits automatically assures that both 10CFR20.105 limits and 10CFR20.l(c)
"as low as is reasonably achievable" criteria are met.
Q. Describe what analysis, if any, you made of liquid radioactive effluents from the Byron Station.
A. Routine liquid radioactive releases from Byron Station are directed, after extensive treatment, to the Rock River.
A dose assessment for radioactivity in liquid effluents was made for an individual involved in various activities in the station discharge canal. (See ER, Table 5.2-7). However, downstream from Byron, the Rock River does not flow past either the DeKalb-Sycamore area or the Rockford area.
Accordingly, any analysis of liquid radioactive releases was not within the scope of the contention.
Q. What conclusions were reached regarding the conformance of Byron Station to Appendix I criteria when the application for a construction permit was granted?
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A. The NRC Staff's'and CECO's proposed joint findings of fact
() and conclusions of law (September 16, 1975) determined that the Byron station complied with 10CFR50.34 (a) ; that is effluents met the"as low as practicable"'(now "as low as reasonably achievable"]
criteria; offsite doses were less than Appendix I criteria. The calculations demonstrated compliance with the Annex to Appendix I. Calculations are reported in the PSAR (Sections 11.2.5 and :
11.3.5), ER (Section 5.3), and the NRC's FES (Section 5.3)
Q. Have further analyses of routine radioactive emissions from Byron Station been made since the construction permit was granted?
A. Yes. Current reassessments made in support of the OL application (see for example ER Table 3.5-4 and 5.2-2 attached hereto as Exhibits 1 and 2) show an even lower release than originally estimated at CP Stage.
Q. Why have the calculations changed since the CP Stage analysis Stage?
A. Lower values are due to evolving refinements in dose assessment and source term generation methodology. No additional equipment for treating radioactive effluents has been installed at Byron.
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Q. Has anyone other than Sargent and Lundy made a re-analysis
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(_) of routine releases from the Byron Station?
A. Yes. The NRC has also reassessed the impact of expected releases from the Byron plant. Their evaluations appear in Appendix C to the Byron Final Environmental Statement (NUREG-0848, November 1981) and again conclude that the Byron Station will meet both 10CFR50, Appendix I and 40CFR part 190 criteria (see Section 5.9.3.2).
Q. Describe the calculations you have made to determine the radiological doses which would be experienced at the DeKalb-Sycamore and Rockford areas from routine atmospheric releases at the Byron site.
A. Radiological doses due to radioactivity released in gaseous effluents are proportional to an atmospheric disper-sion parameter called X/Q (Chi-over-0). This parameter is determined from actual site meteorological data and is reported from the Byron station in its FSAR (Table 2.3-40) and ER (Table 2.3-44). The site meteorological data has been established by monitoring at the site which has taken place since June 1, 1973; this program is described in Section 2.3.3 of the FSAR.
The further one is from the release point, the lower X/Q (and radiological dose) is.
The DeKalb-Sycamore area is about 30 miles ESE of the Byron site. ER Table 2.3-44 gives the following data for the ESE direction.
t 1
Distance X/Q x_-
0.5 miles 35 X 10 - 8 30 miles 0.38 X 10 - 8 Therefore, one may conclude that the doses 30 miles distant will be about 1/100 that observed at 0.5 miles (i . e . , the site boundary). ER Table 5.2-4 shows maximum near-site dose to be less than 1 millirem per year (and less than 10CFR part 50, Appendix I or Annex criteria). Therefore, one would expect doses of the order of 0.01 millirem per year in the DeKalb-Sycamore area due to Byron station operation. The same calculation results in doses of about 0.02 millirems per year in the Rockford area. The expected natural background dose throughout Northern Illinois (including the DeKalb-Sycamore and Rockford areas is about 135 millirems per year (ER Table 5.2-11).
Q. Have you made calculations for the other nuclear power plants operated or under construction by Commonwealth Edison Co.?
A. Yes. Similar calculations may be made for other " neighbor-ing" stations. These stations and their distance from Rockford and DeKalb-Sycamore are as follows:
Approximate Distance, Miles Station Rockford Sycamore-DeKalb LaSalle 72 48 Br aidwood 87 57 l Dresden 75 45 Quad Cities 74 83
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Zion 66 58 1
Q. Describe the calculations made to determinc the routine n
(_) radiological releases from the operating plants described in your previous answer.
A. Calculated offsite radiological doses for CECO operating plants are shown in Exhibit 3. Shown are annual offsite doses calculated for each station (Dresden, Quad-Cities, and Zion) for a person near the site boundary. The doses are calculated by combining the actual, measured releases of radioactivity in effluent streams with the meterological conditions for that time period using NRC-approved dose assessment methodologies and data.
The plot for the Dresden station is dominated by releases from Unit 1, an older unit shut down for decontamination and cleaning in late 1978. .The significant reduction in doses in both Dresden and Quad-Cities in 1975 was achieved by the addition of additional gaseous radwaste treatment systems (Dresden Units 2 and 3; Quad-Cities Units 1 and 2). The reduction of offsite doses at'the Zion station was due to discontinuance of the practice of purging the reactor containment continuously while at power.
The above notwithstanding, the average calculated annual whole body doses in the 1970s were less than 8 millirems (Dresden, 3 units), 2 millirems (Quad-Cities, 2 units) and 1 millirem (Zion, 2 units). Currently, Dresden annual doses are also less than 2 millirem / year. All are less than Appendix I to 10CFR50 design objectives.
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Q.
What is the anticipated annual whole-body doses from the
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(_) operation of Braidwood Station and LaSalle Station?
A. As shown in the Braidwood ER(Table 5.2-4). and La Salle ER (Tabic 5.2-6) maximum offsite doses from radioactivity in gaseous effluents are expected to be less than 1 mrem / year, This maximum is at or near the site boundary, As described earlier, these doses decrease with increasing distance from the plant.
Q. Have you calculated the cumulative dose effect in the DeKalb-Sycamore and Roci. ford areas from the operation of Byron, Braidwood, Zion, Dresden, Quad-Cities and La Salle stations?
A. Yes, although they are more distant than 30 miles from the Rockford and DeKalb-Sycamore areas (see table, page 9) , let us assume that Dresden, Quad-Cities, Zion, Braidwood and La Salle are each 30 miles distant from that area and that the value of X/0 and dose is lower than the site maximum for each plant by a factor of 100, as I calculated previously.
Using actual releases of the 1970s, then the postulated average annual dose in the DeKalb-Sycamore area due to Dresden/ Quad-Cities / Zion operations would be less than 0.11 mrem / year which, again, is insignificant compared to the natural background level of 135 mrem / year and the 40 CFR190 limit of 25 mrem / year. Using releases of the 1980s this dose would be about 0.05 mrem / year. To the figure of
.05 mrem per year, we should add 0.03 mrem / year for the con-tribution of Byron, Eraidwood and LaSalle, making a total O
-A2-postulated cumulative dose of almost 0.08 mrem in the
() Sycamore-DeKalb area.
Q. What consideration, if any, have you given to routine radioactive airborne' releases from aspects of the nuclear fuel
. cycle other than operation of the nuclear power plants?
A. 40 CFR Part 190 specifies certain additional sources.
The sources of radioactivity addressed in 40CFR 190 are as follows:
- a. operations of milling of uranium ore,
- b. chemical conversion of uranium,
- c. isotopic enrichment of uranium,
- d. fabrication of uranium fuel,
- e. generation of electricity by a nuclear power plant using uranium fuel, and
- f. reprocessing of spent uranium fuel.
All of these activities (except e) are distant (over 100 miles) from the CECO service area. By virtue of that distance, they will contribute negligibly to radiological dose in the DeKalb-Sycamore and Rockford areas.
Q. Are there sources of routine atmospheric radioactive emissions other than those identified by 40CFR Part 1907 A. Yes. There are two other sources of radioactivity in this service area.
Q. Describe the first source and its contribution to radiological doses from routine releases from that source.
A. The radiation dose associated with the transport of low-O l
level radioactive waste and spent fuel is excluded from con-O sideration by the requirements of 40CFR190, This subject has been reviewed in References 1 and 2, and summarized again in Reference 3 (Chapter IV, Section G) . The expected dose associated with transportation is 3.4 X 10-3 mrem / person /
reactor / year.
O. Describe the second source and its contribution to radio-logical doses from routine releases from that service.
A. The radiation doses associated with releases of radioactivity by independent spent fuel storage facilities (ISFSF) should also be considered in determining compliance with 40CFR190.
One ISFSF is being operated with CECO's service area; that is, 4
! General Electric Company's Morris Operations Plant at Morris,.
Illinois, adjacent to CECO's Dresden station. Minute quantities of Kr-85 and other radioactive particulates are released in air-born effluents. A dose assessment was performed using typical release data provided by the General Electric Company and the CECO Offsite Dose Calculation Manual (ODCM) environmental dose assessment models for Dresden. (The effluent is released through a 300-foot vent stack; hence the similarity between the two facilities). The estimated annual airborne releases from the GE Morris Operation plant are 6 X 10-4 uCi of tritium, 55 X 10-6 uCi of Kr-85, 3 uCi of Co-60, and 1 uCi of Cs-137. The maximum whole body dose from these redionuclides is 4 X 10-4 mrem /yr.
So long as this plant remains spent fuel storage facility
, and does not reprocess the fuel, dose contributions to any person living in its vicinity will be negligible and not considered ,
further in the 40CFR190 analysis.
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Q. What is your conclusion regarding the compliance of
( emissions from the Byron Station with the two regulatory critoria you identified earlier?
A. Because of their distances from the sources, radiological doses in the DeKalb-Sycamore and Rockford area are expected to meet 10CFR50 Appendix I-criteria for individual stations and 40CFR190 criteria for cumulative effects, due not only to all neighboring stations but all other aspects of the nuclear fuel cycle as well.
Q. In reaching that conclusion, have you considered the comments of R. Mattson, an employee of the NRC, regarding the possibility that in measuring compliance with 40CFR Part 190, the sum of all doses at multiple reactor sites may exceed 25 mrem per year?
A. Yes. I have reviewed those comments which were submitted to the United States Environmental Protection Agency when 40CFR Part 190 was under consideration by that agency. Mr. Mattson was alluding to sites which may contain more than 4 reactors, particularly BWR's with tl.e radioactive isotope N-16 in the turbine systems, and postulating a member of the public at the boundary of this site continuously. As pointed out earlier in this testimony the dose rate due to airborne radioactivity in gaseous effluents drops off rapidly with distance. Similarly the dose rate due to sources of radioactivity confined within the containment and other structures at a nuclear power plant drops off rapidly. Mr. Mattson's comment is intended to be O
applicable to a site boundary locale, not a distant locale
() such as Sycamore-DeKalb or Rockford.' As I stated earlier, the Byron station will comply with both Appendix I and 40CFR190.
O. At my request, have you calculated what the dose effect in the DeKalb-Sycamore and Rockford areas would be, if releases of airborne radioactivity comparable to those experienced following the accident at the Three Mile Island Unit 2 reactor were to occur at Byron?
A. Yes. The Kemeny commission report on the Accident at TMI-2 cites the following estimates of releaces of radioactivity and corresponding doses.
Release: 2.4 million curies, primarily noble gases Maximum individual dose: less than 70 millirem (individual within a few miles of the plant)
Collective man rem dose to population within 50 miles: 2800 man rem Population within 50 miles: 2,163,000 persons Average. dose to persons within 50 miles: about 1 millierem Average dose to persons within 10 miles: 6.5 millirem Since the DeKalb-Sycamore and Rockford plants are more than 30 miles from the Byron site, doses in the DeKalb-Sycamore areas would have been less than 5 millirem if the TMI-2 accident had occurred at the Byron station.
Q. In making your calculations of dosca from airborne radioactive' emissions from the Byron Station, have you taken
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account of the fact that there may be~ days during a year when the routine airborne radioactive' emissions exceed the average annual dose specif.i.ed in Appendix I to 10 CFR Part 50?
A. Yes. Plant releases are required to meet 10CFR50, Appendix I requirements as assessed on an annual average basis.
Short term increases are permitted by regulations and plant Technical Specifications (part of the plant operating license) but do require licensee actions as specified in the Technical Specifications; nevertheless the annual average requirements are to be met. Moreover, operating experience at other nuclear power plants (which I expect to also occur at Byron Station) demonstrates that while there may be short-term
" higher-than-average" releases for a day or several days, there are also many days with "7.ower-than-average" releases resulting in an acceptable average annual release rate. In any event, these
" higher-than-average" releases were taken into account in estimating the releases from the Byron Station. The releases cited in the Byron FSAR were determined using the NRC's GALE computer code. Its modeling is in large part empirical, drawing from actual licensee effluent release data. Further information may be found in the report documenting the GALE code (reference 4) and its predecessor, WASH-1258 (reference 5) which summarizes some of the documentation entered into the Appendix I rulemaking hearing (RM-50-2).
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Q. T would now like to turn to the issue of whether the radio-() Active release assessment for the Byron Station has in fact considered all radionuclides likely to occur in the liquid effluents from Byron Station. Where are those effluents listed in the FSAR?
i A. They are found in Byron Station FSAR Table 11.2-1 titled
" Expected Annual Average Releases in Liquid Effluents" which I prepared. An amended Table 11.2-1 will be submitted
. in the near future as an ?SAR amendment. The numerical data which will be incorporated in Table 11.2-1, as amended, appears as Exhibit 4.
Q. Does that table include values for the releases from the
]
radionuclides Zr-95, Nb-95 and Ce-144?
A. Yes.
Q. Are there radionuclides other than those listed on FSAR 4 Table 11.2-1 which are a byproduct of the fission process which will take place in the nuclear reactors at the Byron
. Station?
A. Yes.
j Q. Why are such radionuclides not included in Table 11.2-1 and the dose assessment for liquid radioactive releases
. from the Byron Station?
A. In order for a specific radionuclide to appear in a liquid.
l effluent stream, several things must happen. First, the fission
- product (or actinide) must migrate from the oxide fuel pellet
through the cladding into the~ reactor coolant, A h4gh
\/ fission product inventory does not necessarily mean a high coolant activity. Similarly, radionuclides which appear as corrosion products must be actually dissolved in the coolant before they need to be analyzed. Each radionuclide demonstrates a different characteristic in this regard. Then the specific pieces of liquid radwaste equipment (filters, domineralizers, evaporators, et cetera) used to process various waste streams each demonstrate a preferential cican-up capability for different elements. All of these parameters have been empirically determined by measurements made in operating nuclear power plants and are documented in the GALE Code report and WASH-1258; these data are then the input to the GALE computer code. Further the design-basis reactor coolant radioactivity concentrations are drawn from the American National Standards Institute Standard ANSI N237-1976/ANS-18.1,
, "American National Standard Source Term Specification".
Radiodecay of these into their respective daughter products is then tracked by the Oak Ridge National Laboratory Computer program, ORIGEN, (ref. 7) included as part of the GALE Code.
ORIGEN has complete and extensive libraries of information which permit following the radiodecay in all of its detail.
Accordingly, I believe the GALE Code adequately represents all of the phenomena with regard to radionuclide formation, transport within the plant, and clean up prior to discharge O
-- ~, - - - - - -
into an effluent stream. The Gale Code was used in analyzing liquid effluents from the Byron Station. The results of that analysis demonstrates that radionuclides other than those listed in Table 11.2-1 will never enter an effluent stream, for the reasons discussed above.
Q. Have radionuclides with release rates of less than .00001 Ci/yr been considered in Table 11.2-17 A. No.
Q. Why not?
A. Such radionuclides are not considered in Table 11.2-1 for practical convenience. As a demonstration,.the GALE Code was run for the Byron plant with cut-offs of 10-5, 10-6 and 10-7 ci/ year for the minimum value to appear in the output table. The results are given in Exhibits 4, 5, and 6, respectively.
(Exhibit 4 is the same as FSAR Table 11.2-1 as revised to date) .
As the screening value is lowered, more nuclides appear on the list. The impact, though of these added nuclides in a dose
! assessment (FSAR Table 11.2-3, Exhibit 7), shows no change in the estimated doses calculated to 3 significant figures (i.e.,the added dose due to those nuclides onsidered as "all others" adds negligibly l to dose already calculated, justifying not considering them) .
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REFERENCES
() 1. WASH-1248, " Environmental Survey of the Uganium Fuel Cycle," April 1974,
- 2. WASH-1238, " Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants,"
December 1972.
- 3. NUREG-0002, " Final Generic Environmental Statement on the use of Recycled Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors," August 1976.
- 4. NUREG-0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE CODE) ," April 1976.
- 5. WASH-1258, " Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," July 1973.
- 6. Staff Reports to the President's Commission on the Accident at Three Mile Island; " Reports of the Public Health and Safety Task Force," (page 9), October 1979.
- 7. ORIGEN "The ORNL Isotope Generation and Depletion Code," ORNL-4628, May 1973.
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4 TABLE 3.5-4 EXPECTED ANNUAL AVERAGE RELEASE OF AIRBORNE RADIONUCLIDES FRIMARY SECONDARY CASEOUS RELEASE RATE (Ci/yr)
CODIANT COO! ANT CAS STRIPPING BUILDING VENTILATION BIDhTOWN AIR EJECTOR ISOTOPE (WCi/g) (bC1/q) SHUTDOWN CONTINUOUS REACTOR AUXILIARY TURBINE VENT OFFCAS EXHAUST TOTAL Kr-83m 2.265x10" 6.255x10~ 0.0a 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-85m 1.184x10
~I 3.337x10' O.0 0 0 0 1 0.0 0.0 3.0x10 0.0 . 0.0 2.0x10 5.0x10
-8 1 2 0 0 2 Kr-85 1.051x10"I 2.944x10 5.1x10 5.7x10 7.4x10 2.0x10 0.0 0.0 1.0x10 7.0x10 8.474x10~ 1.726x10 -8 0 0 Kr-87 0.0 0.0 0.0 1.0x10 0.0 0.0 0.0 1.0x10 2.158x10"I 5.928x10~ 0 0 0 Kr-88 0.0 0.0 0.0 5.0x10 0.0 0.0 3.0x10 8.0x10
-3 Kr-89 5.399x10 1.512x10 O.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0 1 0 0 Xe-131m 1.035x10~ 2.917x10" 4.0x10 1.5x10 1.7x10 2.0x10 0.0 0.0 1.0x10 3.9x10 2.293x10
~I 6.461x10~ 0 0 0 N Xe-13 3m 0.0 0.0 - 7.0x10 5.0x10 0.0 0.0 3.0x10 1.5x10 I 1 1 3 f Xe-133 1.804x10 5.010x10~ 2.4x10 4.7x10 1.3x10 3.8x10 0.0 0.0 2.4x10 2.0x10 W Xe-135m 1.404x10~ 3.887x10~ 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 M 0 0 0 Xe-135 3.755x10"I 1.041x10~ 0.0 0.0 2.0x10 8.0x10 0.0 0.0 5.0x10 1.5x10 I
O Xe-137 9.719x10
-3 2.700x10
-9 0.0 0.0 O
0.0 0.0 0.0 0.0 0.0 0.0 t1 Xe-138 0 0 W 4.751x10~ 1.296x10 0.0 0.0 0.0 1.Ox10 0.0 0.0 0.0 '1.0x10 TOTAL NOBLE GASES 2.3x10 1-131 2.795x10~
~
4.215x10 ' O.0 -2 ~3 ~3 0.0 1.7x10~ 4.4x10 2.3x10 0.0 2.8x10 5.1x10~
~I ~3 I-133 3.986x10 3.831x10~ 0.0 0.0 7.7x10 6.3x10~ 2.1x10~ 0.0 4.0x10 7.0x10~
TRITIUM GASEOUS RELEASE 1000 Ci/yr C.s C
t*
MZ m x
O :y H 3: p.
cM o.,
ao Z p.
Hd rr
- The figure 0.0 appearing in the table indicates that the release is less than 1.0 Ci/yr for noble gas, 0.0001 Ci/yr for I. y
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TAPLE 3.5-4 (Cont'd)
AIRBORNE PARTICULATE RELEASE RATE (Ci/yr)
WASTE GAS BUILDING UENTILATION NUCLIDE SYSTEM REACTOR AUXILIARY TOTAL
-6 -4 ~3 Mn-54 4.5x10" 6.1x10 1.8x10 4.7x10
-6 -5 -3 Fe-59 1.5x10~ 2.1x10 6.0x10 1.6x10
-2 1.6x10~
Co-58 1.5x10 2.1x10~ 6.0x10~
-3 -6 ~4 Co-60 7.0x10 9.5x10 2.7x10 7.3x10'3 Sr-89 3.3x10
~4 4.7x10~ 1.1x10" 3.4x10 ~4
-6 -8 -6 Sr-90 6.0x10 8.4x10 2.4x10 6.2x10~ 62
-3 6.1x10 -6
-3 -4 Cs-134 4.5x10 1.8x10 f.7x10 y
-5 ~4 w Cs-137 7.5x10~ 1.1x10 3.0x10 7.8x10~
- M (n O e
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l Note: In addition to these releases, 25 Ci/yr of argon-41 are released from the containment and 8 Ci/yr of carbon-14 are released from the waste gas processing system.
TABLE 5.2-2 ANNUAL AVERAGE SITE BOUNDARY DOSES PLUME IMMERSION GROUND DEPOSI""!ON INHALATION DOWNWI*:D DISTANCE BETA-AIR GAMMA .".I R SKIN WHOLE BODY SKIN WHOLE BODY DIRECTION THYROID (meters) (mrad /yr) (mrad /yr) (mrem /yr) (mrem /yr) (Prem/yr) (mrem /yr) (mrem /y r)
N 1875 1.54x10 -2 1.05x10 -2 1.33x10 -2 5.80x10 ~3 1.75x10 -2 1.50x10~ 1.22x10 -2 8
NNE 1829 1.03x10" 7.55x10-3 9.29x10-3 4.26x10-3 1.52x10~ 1.30x10 -2 8.30x10 ~3 NE 1585 8 1.06x10' 7.30x10-3 9.22x10~3 4.25x10-3 1.39x10~ 1.20x10 -2 8.83x10~
ENE 1235 8 1.27x10" 8.92x10~3 1.12x10' 5.17x10-3 1.46x10 -2 1.24x10 -2 1.06x10 -2 8
E 1228 1.64x10~ 1.07x10~ 1.38x10~ 6.32x10~3 1.65x10~ 1.41x10 ~2 1.37x10 ~2 8
ESr.; 991 1.94x10 -2 1.36x10 -2 1.70x10 -2 7.83x10 -3 2.17x10 -2 1.85x10~ 1.62x10 -2 g
-2 ~3 M
SE 906* 2. 57x10 1.82x10~ 2.39x10~ 1.09x10 2.25x10~ 1.92410~ 2.44x10"
-2 -2 -2 h
. SSE
2.48x10 1.79x10 2.21x10 1.01x10~ 2.29x10~ 1.95x10~ 2.02x10 -2 3 S ,4 1.31x10 -2 1.09x10 ~2 1.28x10
-2 5.91x10~ 1.45x10 -2 1.24x10 -2 1.05x10 -2 H SSW S>a 9.62x10 -3 8.23x10 ~3 9.60x10 ~3 4.40x10~ 1.25x10~ 1.07x10 -2 7.76x10 ~3 O SW 1067 9.24x10~ 4.68x10" 9.39x10" 4.24x10~ 1.54x10 -2 1.32x10
-2 7.44x10~
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7.27x10 ~3 WSW 1212 8.64x10 -2 3.89x10~ 1.38x10~ 1.18x10 W 1189 9.64x10" 7.54x10~ 9.07x10" 4.10x10 -3 1.23x10~ 1.05x10
-2 7.71x10 ~3
~3 WNW 1227 8.54x10 3.48x10 ~3 7.96x10
-3 3.54x10 ~3 1.25x10~ 1.06x10 -2 6.85x10
~3 1128 1.02x10" 7.80x10" -3 -2 -2 NW 9.44x10" 4.15x10 1.40x10 1.26x10 8.16x10 -3
-2 -3 -2 -2 NNW 1044 2.03x10~ 1.41x10 1.77x10~ 8.0lx10 2.67x10 2.28x10 1.63x10~
C MM KZ O
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DATE 040582 PACE C8 , .
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BYRON NUCLEAR STATION LIOUID EFFLUENTS ANNUAL RELEASES TO DISCHARGE CANAL COOLANT CONCENTRATIONS--------------------------------------------------------- ADJUSTED DETERGENT TOTAL NUCLICE HALF-LIFE PRIMARY SECONDARY BORON RS MISC. WASTES SECONOARY TUR8 BLOG TOTAL LWS TOTAL WASTES (DAYS) (MICRO CI/ML)(MICRO CI/ML) (CURIES) (CURIES) (CURIES) (CURIES) (CURIES) (Ct/YR) (CI/YR) (CI/YR)
CORROSION ANO ACTIVATION PRODUCTS CR 51 2.78+01 1.90-03 2.51-07 5.3G-06 6.13-09 0.00 2.48-06 7.85-06 6.16-05 0.00 6.16-05 f?N 54 3.03+02 3.10-04 6.03-08 9.21-07 1.05-09 0.00 6.05-07 1.53-06 1.20-05 1.00-03 1.01-03 FE 55 9.50+02 1.60-03 2.12-07 4.77-06 5.44-09 0.07 2.11-06 6.89-06 5.40-05 0.00 5.40-05 FE 59 4.50*01 1.00-03 1.55-07 2.88-OG 3.29-09 0.00 1.54-06 4.42-06 3.47-05 0.00 3.47-05 CO 58 7.13+01 1.00-02 2.15-06 4.67-05 5.34-08 0.00 2.14-05 6.81-05 5.35-04 4.00-03 4.53-03 CO 60 1.92+03 2.00-03 2.73-07 5.97-06 6.81-09 0.00 2.72-06 8.69-06 6.82-05 8.70-03 8.77-03 ZR 95 6.50*01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.40-03 1.40-03 NB 95 3.50+01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 2.00-03 2.00-03 NP239 2.35+00 1.20-03 1.23-07 1.82-06 2.32-09 0.00 1.14-06 2.96-06 2.22-05 0.00: 2.32-05 FISSION PRODUCTS BR 83 1.00-01 4.80-03 1.26-07 2.11-09 2.69-09 0.00 2.24-06 2.24-06 1.76-05 0.00 1.76-05 R8 86 1.87+01 8.50-03 1.40-08 5.83-OG 1.34-08 0.00 1.38-07 5.98-06 4.69-05 0.00 4.69-05 SR 89 5.20+O1 3.50-04 6.17-08 1.01-06 1.16-09 0.00 6.13-07 1.63-06 1.28-05 0.00 1.28-05 SR 91 4.03-01 6.50-04 3.91-08 3.90-08 2.75-10 0.00 2.53-07 2.93-07 2.30-06 0.00 2.30-06 Y 91M 3.47-02 3.60-04 2.74-08 2.52-08 1.77-10 0.00 1.64-07 1.89-07 1.48-06 0.00 1.48-06 Y 91 5.88+01 G.40-05 9.24-09 2.00-07 2.27-10 0.00 9.27-08 2.93-07 2.30-06 0.00 2.30-06 7R 95 6.50*01 6.00-05 9.23-09 1.75-07 2.00-10 0.00 9.16-08 2.67-07 2.09-06 0.00 2.09-06 NB 95 3.50*01 5.00 05 9.34-09 1.50-07 1.71-10 0.00 9.30-08 2.44-07 1.91-06 0.00 1.91-06 fc0 99 2.79+00 8.40-02 1.19-05 1.42-04 1.76-07 0.00 1.11-04 2.53-04 1.98-03 0.00 1.98-03 TC 99M 2.50-01 4.80-02 2.18-05 1.35-04 1.65-07 0.00 1.58-04 2.94-04 2.31-03 0.00 2.31-03 RU103 3.96+01 4.50-05 6.21-09 1.29-07 1.48-10 0.00 6.16-08 f.91-07 1.50-06 1.40-04 1.41-04 RHtO3M 3.96-02 4.50-05 1.82-08 1.29-07 1.48-10 0.00 6.31-08 1.93-07 1.51-06 0.00 1.51-06 RU106 3.67+02 1.00-05 1.52-09 2.97-08 3.39-11 0.00 1.51-08 4.49-08 3.52-07. 2.40-03 2.40-03 AG110M 2.53*02 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 4.40-04 4.40-04 TE127M 1.OS+02 2.80-04 2.75-08 8.24-07 9.41-10 0.00 2.74-07 1.10-06 8.62-06 0.00 8.62-06 1Et27 3.92-01 8.50-04 1.31-07 8.4G-07 1.24-09 0.00 9.37-07 1.78-06 1.40-05 0.00 1.40-05 TE129M 3.40+01 1.40-03 1.87-07 3.99-06 4.56-09 0.00 1.85-06 5.85-06 4.59-05 0.00 4.59-05 TE129 4.79-02 1.60-03 5.39-07 2.56-06 2.94-09 0.C0 1.30-OG 3.86-06 3.03-05 0.00 3.03-05 1130 5.17-01 2.10-03 1.53-07 2.95-06 1.18-08 0.00 1.09-05 1.38-05 1.09-04 0.00 1.09-04 TEt31M 1.25+00 2.50-03 2.39-07 2.10-06 3.24-09 0.00 2.07-06 4.17-06 3.27-05 0.00 3.27-05 TE131 1.74-02 1.10-03 4.69-07 3.83-07 5.92-10 0.00 3.79-07 7.62-07 5.98-06 0,00 5.98 -OG
- -s. 1131 8.05+00 2.70-01 3.72-05 6.62-03 7.69-06 0.00 3.63-03 1.03-02 8.04-02 6.20-05 8.05-02 Tf132 3.25+00 2.70-02 3.07-06 4.94-05 6.02-08 0.00 2.90-05 7.85-05 6.16-04 0.00 6.16-04 1132 9.58-02 1.00-01 9.32-06 5.11-05 3.47-07 0.00 1.77-04 2.28-04 1.79-03 0.00 1.79-03 1133 8.75-01 3.80-01 3.43-05 1.85-03 3.65-06 0.00 2.80-03 4.66-03 3.66-02 0.00 3.66-02
-"a 1134 3.67-02 4.70-02 5.20-07 1.23-09 2.27-0G 0.00 4.59-07 4.62-07 3.63-06 0.00 3.G3-06 CS134 7.49+02 2.50-02 4.01-06 1.86-03 4.25-06 0.00 3.99-05 1.91-03 1.50-02 1.30-02 2.80-02 1135 2.79-01 1.90-01 9.96-06 2.10-05 5.10-07 0.00 5.33-04 5.54-04 4.35-03 0.00 4.35-03 CS136 1.30+01 1.30-02 1.78-06 8.59-04 1.98-06 0.00 1.75-05 8.79-04 G.89-03 0.00 6.89-03 CS137 1.10604 1.80-02 2.67-06 1.34-03 3.OG-06 0.00 2.GS-OS 1.37-03 1.08-02 2.40-02 3.48-02 M BA137M 1.77-03 1.00-02 7.67-OG 1.26-03 2.87-06 0.00 2.48-05 1.29-03 1.01-02 0.00 1.01-02 X 2.88-07 8.69-07 6.81-06 0.00 6.81-06 D*
04140 1.28+01 2.20-04 2.93-08 5.00-07 6.68-10 0.00 LA140 1.G7+00 1.50-04 3.55-08 5.48-07 6.04-10 2.28-10 0.00 0.00 3.47-07 9.27-08 8.96-07 2.92-07 7.03-06 2.29-06 0.00 0.00 7.01-06 2.29-06 h
p.
CE141 3.24*01 7.00-05 9.36-09 1.99-07 PR143 1.37+01 5.00-05 6.48-03 1.41-07 1.60-10 0.00 6.42-08 2.05-07 1.61-OG O.00 1.G1-06 ft CE144 2.84+02 3.30-05 6.08-09 9.80-03 1.12-10 0.00 6.05-08 1.59-07 1.24-OG 3.20-03 5.20-03 (o PR144 1.20 02 3.30-05 1.89-08 9.80-00 1.12-10 0.00 6.05-08 1.59-07 1.24-OG 2.53 06 O.00 0.0 1.24-OG 2.53-06 h
ALL OTHERS 2.03-01 7.93-Of 2.01-07 4.97-10 0.00 1.21-07 3.22-07 .
Q TOTAL m *
(rxCEP1 TRITIUM) 1.46600 1.50-04 1.43-02 2.49-05 0.00 7.00-03 2.19-02 1.72-01 6.23-02 2.34-01 SARGENTuf. UNDY
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BY;0N GALE-PWR LIQUID SOURCES FO? VAGICUS CUTOFF LIMITS-EPSIL CATE 040582 PAGE 73 *
~
BYRON NUCLEAR STATION LIQUID EFFLUENTS (CONTINUED)
ANNUAL RELEASES TO DISCHARGE CANAL COOLANT CONCENTRATIONS--------------------------------------------------------- ADJUSTED DETERGENT TOTAL NUCLIDE HALF-LIFE PRIMARY SECONDARY BORON RS MISC. WASTES SEC0fJDARY TUR8 BLDG TOTAL LWS TOTAL WASTES (OAYS) (MICRO CI/ML)(MICR0 CI/ML) (CURIES) (CURIES) (CURIES) (CURIES) (CURIES) (CI/YR) (CI/YR) (CI/YR)
CE141 3.24+01 7.00-05 9.36-09 1.99-07 2.28-10 0.00 9.27-08 2.92-07 2.29-06 0.00 2.29-06 CE143 1.37400 4.00-05 4.20-09 3.76-08 5.57-11 0.00 3.69-08 7.46-08 5.85-07 ' 00 5.85-07 PR143 1.37+01 5.00-05 6.48-09 1.41-07 1.60-10 0.00 6.42-08 2.05-07 1.61-06 v.OO 1.61-06 CE144 2.8-1+02 3.30-05 6.08-09 9.80-08 1.12-10 0.00 6.05-08 1.59-07 1.24-06 5.20-03 5.20-03 PR144 1.20-02 3.30-05 1.89-08 9.80-08 1.12-10 0.00 6.05-08 1.59-07 1.24-06 0.00 1.24-06 ALL OTHERS 2.03-01 7.76-07 1.28-09 2.46-10 0.00 8.15-10 2.33-09 1.86-08 0.0 1.86-08 TOTAL (ExCEPT TRITIUM) 1.46+00 1.50-04 1.43-02 2.49-05 0.00 7.60-03 2.19-02 1.72-01 6.23-02 2.34-01 TRITIUM RELEASE 300 CURIES PER YEAR o
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SARGENTrALUNDY .
Exhibit No. 7 BYRON-FSAR AMENDMENT 20 I MAY 1979 l TABT,E 11.2-3 O
PAT!! WAYS DOSES FROM LIQUID EFFLUENTS EXPOSURE PATHWAY ORGAN DOSE (mrem /yr/ unit) * -
Drinking Water Whole Body 9.08 x 10_yy GI-LLI 8.35 x 10 Thyroid 3.67 x 10 0 Bone 7.02 x 10 Fish Consumption
_y Whole Body 4.54 x 10 GI-LLI -2 -
Thyroid 1.92 x 10_1 Bone 1.03 x.10_1 3.48 x 10 Shoreline Recreation Skin -3 9.11 x 10 Whole Body -3 7.80 x 10
._ Swimming & Boating _
~
Skin 3.21 x 10_44 Whole Body 2.48 x 10
~
- All activities are assumed to take place in the discharge canal. !;o credit is taken for dilution of effluents in
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v the Rock River.
11.2-21 '