ML20237B561
| ML20237B561 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/19/1977 |
| From: | Jay Collins Office of Nuclear Reactor Regulation |
| To: | Lear G Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8712160351 | |
| Download: ML20237B561 (23) | |
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, f n 9 !c y DEC 191977 M Docket Ho. 50-285 MEMORANDUM FOR: G. Lear, Chief, Operating Reactors Branch No. 3. DOR FROM:
John T. Collins, Chief. Effluent Treatment Systems Branch, OSE DSE EVALUATION OF FORT CALH00H STATION, UNIT NO.1. MITH
SUBJECT:
RESPECT TO APPENDIX I TO 10 CFR PART 50 Enclosed is DSE's detaiJ ed evaluation of the radioactive waste treat: ent systems installed at Fort Calhoun Station, Unit Mo.1, Wth respect to the recuirenents of Appendix I. The results of our evaluation are contained in the attached " Safety Evaluation and Enviremental Impact Apprai sal ." We have also attached a draft " Notice of Issuance of Amend-ment to Facility Operating Licenses and Negative Declaration."
9ased on our evaluation, we conclude that the radioactive waste treatment systems installed at Fort Calhoun - 1 are capable of maintaining releases, of radioactive materials in effluents to "as low as is reasonably achievable'"-
1cvels in confomance with the requirements of 10 CFR Part 50.34a, and conforms to the requirements of Sections II. A, II.3, II.C, and II.D of Appendix I.
On March 29, 1977, DSE transmitted to ELD an HRC Staff Reoort entitled,
" Application of Cost-Benefit Analysis Requirements of Appendix ! to 10 CFR Part 50 to Nuclear Power Plants Whose Applications Were Docketed Before January 2,1971." This report provides the staff's justification for using the September 4,1975 amencment to Appendix I, rather than than performing a detailed cost-benefit analysis recuired by Section II.D of Appendix I. On August 17, 1977, we received ELD cements on this report and we are currently preparing a HUREG report Wich will document our findings. When this report is completed, we will forward to you a paragraph to be inserted on pace 1 of the enclosed Safety Evaluation, providino justifi-cation for the use of the September 4 option to the cost-benefit analysis.
When the model effluent radiological Technical Specifications, currently under development, have been approved, they will be forwarded to you for transmittal to the licensee.
ORN mg DISTRIBUTION: Docket File 50-285 m L co m DSE Reading File ETSB Reading File J hn T. Collins, Chief ETSB Docket File Effluent Treatment Systems Branch
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G. Lear DEC 101977 i
Enclosure:
DSE Evaluation cc: H. Denton V. Stello R. Vollmer K. Go11er D. Jaffee D. Elliot D. Eisenhut W. Kreger H. Hulman B. Grimes E. Markee F. Congel W. Burke R. Bellamy T. Huetter aul
SAFETY EVALUATION AND ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. TO FACILITY LICENSE NO. OPR-40 0F%HA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285 INTRODUCTION On May 5,1975, the Nuclear Regulatory Commission announced its decision in the rulemaking proceeding concerning the numerical guides for design objectives and limiting conditions for operation to meet the criterion "as low as is reasonably achievable" for radioactive materials in light-water-cooled nuclear power reactor effluents. This decision is set forth in Appendix I to 10 CFR Part 50.II) On September 4,1975, the Commission adopted an amemdment to Appendix I to provide persons who have filed applications for construction pennits for light-water-cooled nuclear power reactors which were docketed on or after January 2,1971, and prior to June 4,1976, the option of dispensing with the cost-benefit analysis reauired by Section 11.0 of Appendix I, if the proposed or installed radwaste systems satisfy the guides on design objectives for light-water-cooled nuclear power reactors proposed by the Regulatory Staff in the rulemaking proceeding on Appendix I (Docket RM 50-2), dated February 20, 1974.(3)
A paragraph will be added which will provide justification for using the September 4,1975 amendment to Appendix I for application for construction permits filed prior to January 2,1971.
Section V.B of Appendix I to 10 CFR Part 50 requires the holder of a license authorizing operation of a reactor for which application was filed prior to January 2,1971, to file with the Commission by June 4, 1976; 1) information necessary to evaluate the means employed for keeping levels of radioactivity
_2_
in effluents to unrestricted areas "as low as is reasonably achievable", and
- 2) plans for proposed Technical Specifications developed for the purpose of keeping releases of radioactive materials to unrestricted areas during normal operation, including anticipated operational occurrences "as low as is reasonably achievable."
In conformance with the requirements of Section V.B of Appendix I, the Omaha Public Power District (OPPD) filed with the Commission on June 4,1976,(
the necessary information to permit an evaluation of the Fort Calhoun Station, Unit No.1, with respect to the requirements of Sections II. A, II.8, and II.C ,
of Appendix 1. In this submittal, OPPD provided the necessary information to show conformance with the Commission's September 4,1975 amendment to Appendix I rather than perform a detailed cost-benefit analysis reauired by Section II.D of Appendix I.
By letter dated , OPPD submitted proposed changes to Appendix A Technical Specifications for Fort Calhoun Station, Unit No.1. The proposed changes implement the requirements of Appendix I to 10 CFR Part 50 and provide reasonable assurance that releases of radioactive materials in liauid and gaseous effluents are "as low as is reasonably achievable" in accordance with 10 CFR Parts 50.34a and 50.36a.
DISCUSSION The purpose of this report is to present the results of the NRC staff's l
detailed evaluation of the radioactive waste treatment systems installed at
, (
)
the Fort Calhoun, Unit No.1, Station; 1) to reduce and maintain releases of radioactive materials in liquid and gaseous effluents to "as low as is reasonably achievable" levels in accordance with the requirements of 10 CFR
}
Parts 50.34a and 50.36a, 2) to meet the individual dose design objectives set forth in Sections II. A,11.8, and II.C of Appendix I to 10 CFR Part 50, and 3) to detemine if the installed radwaste systems satisfy the desion objectives proposed in RM 50-2 rather than an individualized cost-benefit b analysis as reauired by Section II.D of Appendix I.
I. Safety Evaluation The NRC staff has perfomed an independent evaluation of the licensee's pro-posed method to meet the requirements of Appendix I to 10 CFR Part 50. The staff's evaluation consisted of the followina: 1) a review of the infomation provided by the licensee in his June 4,1976, submittal; 2) a review of the radioactive waste (radwaste) treatment and effluent control systens described in the licensee's Final Safety Analysis Report (FSAR); 5) 3) the calculation of expected releases of radioactive materials in linuid and caseous effluent (source tems) for the Fort Calhoun facility; 4) the calculation of relative concentration (X/0) and deposition (D/0) values for the Fort Calhoun site reoion; 5) the calculation of individual doses in unrestricted areas; and 6) the comparison of the calculated releases and doses with the p' oposed r desian i
objectives of RM 50-2 and the requirements of Sections II. A, II.8, II.C and 11.0 of Appendix I.
/
) The radwaste treatment and effluent control systems installed at the Fort i
Calhoun Station have been previously evaluated in Section 3.1.7 of the staff's Safety Evaluhtion Report (SER), dated August 09, 1973,(6) and h' ave been described and evaluated in Section III.D.2 of the Final Environmental Statenent (FES) dated August 1972. Since the FES was issued, the licensee has proposed a modification to the steam generator blowdown system to include two 75-gpm mixed bed ion exchangers in series. The modifications noted above s
3 were considered in the staff's evaluation.
Based on more recent operating data at other operating nuclear power reactors, which are applicable to the Fort Calhoun Station, and on changes in the staff's calculation models, new liquid and gaseous source terms have been generated to detennine conformance with the requirements of Appendix I. The new source ,
terms, shown in Tables 1 and 2, were calculated using the model and parameters described in NUREG-0017.(8) In making these determinations, the staff con-sidered waste flow rates, concentrations of radioactive materials in the primary and secondary system and equipment decontamination factors consistent with those expected over the 30 year operating life of the plant for normal opera-tion including anticipated operational occurrences. The principal parameters and plant conditions used in calculating the new liquid anc gaseous source tenns are given in Table 3.
The staff also reviewed the operating experience accumulated at the Fort Calhoun Station in order to correlate the calculated releases given in Tables 1 and 2 ,
/
( *
- with observed ' releases of radioactive materials in liquid and gaseous effluents.
. Data on liquid and gaseous effluents are contained in the licensee's Semi-Annual. 0perating Reports covering the period for January 1975 through June 1977. These observed releases are summarized in Table 4.
Fort Calhoun Station, Unit No.1, reached initial criticality on August 6,1973, and commercial cperation on June 20, 1974 The staff does not consider tae releases .for the period from August 1973 to December 1974 as being representative of current operating conditions at the Fort Calhoun Station. Only those releases reported for the period January 1975 through June 1977 were considered for comparison with calculated. releases.
The actual average liquid release of 0.40 Ci/yr was somewhat lower than the calculated release of 1.0 Ci/yr. The actual releases of noble gases in gaseous effluents averaged 2000 Ci/yr, compared to the calculated release of 5,800 Ci/yr.
In both cases, the lower release rate is due to better fuel performance than the staff assumed in its evaluation.
The actual releases of iodine-131 in gaseous effluents averaged 0.014 Ci/yr, j or about 20% of the staff's calculated release of 0.067 Ci/yr. As in the l
previous cases, the lower release rate is due to better fuel performance than the staff assumed in its evaluation.
- The calculated . releases given in Tables,1 and 2 were used in the staff's dose assessment discussed below.
. .- := . .= .=- - - - - - - - - - - -- - - - -- - -
The' staff. has made reasonable estimates of averace atmosphere dispersion con-ditions for the Fort Calhoun Site using an atmospheric dispersion model appropriate ' for long-term. releases. The model used by the ' staff is based upon the " Straight-L'ine Trajectory tiodel" described in Reaulatory Guide 1.111.(10)
'This evaluation is different from and replaces the evaluation aiven in Section V.D of the FES.I7I Using the guidance given in Reaulatory Guide 1.111, the
. staff considered that gaseous effluents were a mixture of elevated and around-level releases. Intermittent gaseous releases were evaluated separately from continuous releases. The calculations also include an estimate of maximum increase in-calculated relative concentration and deposition due to the spatial and temporal variation' of the. airflow not considered in the straight-line trajectory model . The contributions of the variations are discussed in Regul atory. Guide 1.111.
The staff's dose assessment considered the followinn three effluent catenories:
- 1) pathways' associatedLwith radioactive naterials released in liouid effluents ,
i to the Missouri River, 2) pathways associated with noble cases released to the atmosphere; and 3) pathways associated with radioindines, particulate, 1 carbon-14, and . tritium released to the atmosphere. The mathematical models used by the staff to perform the dose calculations to the maximum exposed individ-ual are described in Regulatory Guide 1.109."I 4
-The dose evaluation of pathways associated with the release of radioactive materials in liouid effluents was based -on the maximum exposed individual.
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. . 1 I
l For the total body dose, the staff considered the maximum exposed individual to be an adult whose diet included the consumption of fish (21 kg/yn) har-vested in the immediate vicinity of the discharge from Fort Calhoun Station into the Missouri River and use of the shoreline for recreational purposes (10 hr/yr) . )
The dose evaluation of noble gases released to the atmosphere included a calculation of beta and gamma air doses at the site boundary and total body and skin doses at the residence havina the highest dose. The maxi-mum air doses at the site boundary were found at 0.66 miles WNW relative to the Fort Calhoun Station. The location of maximum total body and skin doses were determined to be at the same location.
The dose evaluation of pathways- associated with radiciodine, particulate, carbon-14, and tritium released to the atmosphere was also based on the maximum exposed individual . For this evaluation, the staff considered the maximum exposed individual to be an infant whose diet included the consumption of 3301/yr of milk from animals grazing 6 months of the year at 0.75 miles SSW of Fort Calhoun.
Using the dose assessment parameters noted above and the calculated releases of radioactive materials in liauid effluents given in Table 1., the staff calculated the annual dose or dose commitment to the total body or to any organ of an individual, in an unrestricted area, to be less than 3 mrem / reactor and 10 mrem / reactor, respectively, in conformance with Section II. A of Appendix I.
______-___a
a__ ., _
. s.
1 1Using the dose assessment parameters noted' above, the calculated releases of radioactive materials in gaseous effluents given in Table 2, and the' appro-priate relative concentration (X/0) value given in Table 5, the staff calculated the annual gamma and beta air doses at or beyond. the site boundary to be less than 10 mrad / reactor and 20 mrad / reactor, respectively, in' con-formance with Section II.B of Appendix I.
Using the dose assessment parameters noted above, the calculated releases of radioiodine, carbon-14, tritium, and particulate given in Table 2, and the appropriate relative concentration (X/Q) and deposition (D/Q) values given in Table 5, the staff calculated the annual dose or dose commitment to any organ
' of the maximum exposed individual to be less than 15 mrem / reactor in conformance with Section II.C of Appendix I.
The sunmary of calculated doses given in Table 6 are different from and replace those given in Table 5.4 of the FES.
Rather than performing an individualized cost-benefit analysis required by Section II.D of Appendix I, the licensee elected to show conformance with the numerical design objectives specified in the September 4,1975 amendment to Appendix I (RM 50-2). The dose design objectives contained in RM 50-2 are on a site basis rather than a per reactor basis while the curie releases are oni a per reactor basis. As shown in Table 1 the calculated release of ;
i radioactive material in liauid effluents is less than 5 Ci/yr/ reactor, ex- l cluding tritium and dissolved noble gases. As given in Table 2, the calculated k.i - i i . . . _ . . . . . . . _ .. . . . .. . _ _ .
-- ~ ~ - - -
1
. quantity of iodine-131 released in gaseous effluents is less than 1 C1/yr/
r eactor. The calculated doses for the Fort Calhoun Station are less than the dose design objectives set forth in Rti 50-2, therefore, satisfy the require-ments of Section II.D of Appendix I.
CONCLUSION Based on the foregoing evaluation, the staff concludes that the radwaste treatment systems installed at the Fort Calhoun Station, Unit No.1, are capable of reducing releases of radioactive materials in liquid and gaseous effluents to "as low as is reasonably achievable" levels in accordance with the require-ments of 10 CFR Part 50.34a, and therefore, are acceptable.
The staff has performed an independent evaluation of the radwaste systems installed at Fort Calhoun. This evaluation has shown that the installed systems are capable of maintaining releases of radioactive materials in liquid and gaseous effluents during normal operation including anticipated operational occurrences such that the calculated individual doses are less than the numerical dose design objectives of Section II. A. II.B and II.C of Appendix I to 10 CFR Part 50. In addition, the staff's evaluation has shown that the radwaste systems satisfy the design objectives set forth in Rti 50-2 and, therefore, satisfy the requirements of Section II.D of Appendix I to 10 CFR Part 50.
The staff concludes, based on the considerations discussed aboye, that:
(1) because the revised Technical Specifications do not involve a significant increase in the probability of consequences of accidents previously considered and does not involve a significant hazard consideration, (2) there is reason-able assurance that the health and safety of the public will not be endangered I
by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and; the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of.the public.
II. Environmental Impact Appraisal The licensee is presently licensed to possess and operate the Fort Calhoun Station, Unit No.1, located in the State of Nebraska, in Washington County, at power levels up to 1420 megawatts thermal (MWt). The proposed changes to the liquid and gaseous release limits will not result in an increase or decrease in the power level of the reactor. Since neither power level nor .
fuel burnup is affected by the action, it does not affect the benefits of electric power production considered for the captioned facility in The Commission's Final Environmental Statement (FES) for Fort Calhoun Station, Unit No.1, Docket No. 50-285.
The revised liquid and gaseous effluent limits will not significantly change the total quantities or types of radioactivity discharged to the environment from Fort Calhoun.
The revised Technical Specifications implement the requirements of Appendix I j to 10 CFR Part 50 and provide reasonable assurance that releas,es of radio-active materials in liquid and gaseous effluents will be "as low as is reasonably achievable." If the plant exceeds one-half the design objcctives in a quarter, the licensee must: (1) identify the cases, (2) initiate a program to reduce the releases; and (3) report these actions to the NRC. The revised !
Technical Specifications specify that the annual average release be maintained at less than twice the design objective quantities set forth in Sections II. A, II.B, and II.C of Appendix 1.
Conclusion and Basis for Necative Declaration On the bas.is of the foregoing evaluation, it is concluded that there would be no significant environmental impact attributable to the proposed action.
Having made this conclusion, the Commission has further concluded that no environmental impact statement for the proposed action need be prepared and that a negative declaration to this effect is appropriate.
Dated:
l O
REFERENCES
.1. Title 10, CFR Part 50, Appendix I. Federal Register, V. 40, p.19442, May 5, 1975.
- 2. Title 10, CFR Part 50,- Amendment to Paragraph II.D of Appendix I, Federal Register, V. ~40 p. 40816, September 4,1975, and revised as of January 1,1976.
- 3. U.S. Atomic Energy Commission Concluding Statement of Position of the Regulatory Staff (and its Attachment) - Public Rulemaking_ Hearing on:
Numerical Guides for' Design Objectives and Limiting Conditions for .
Operation to Meet the Criteria "As low As is Reasonably Achievable" l for Radioactive Material in Light-Water-Cooled Nuclear Power Reactors,
- Docket No. RM 50-2, Washington, D.C., February 20, 1974.
4.. " Evaluation of Fort Calhoun Station, Unit No.1, in Accordance with 10 CFR Part 50, Appendix I," Omaha Public Power District, June 1976.
- 5. Omaha Public Power District, Final Safety Analysis Report for Fort Calhoun Station, Unit. No.1, November 28, 1969.
- 6. Staff of the U.S. Atomic Energy Commission, " Safety Evaluation of the Omaha Public- Power District, Fort Calhoun Station, Unit No.1,"
Docket No. 50-285,' Washington, D.C., August 9,1973.
- 7. Staff of the U.S. Atomic Energy Commission, " Final Environmental Statement Related to the Operation of Fort Calhoun Station, Unit No.1," Omaha Public Power District, Docket No. 50-285, Washington, D. C. , August 1972.
8.- NUREG-0017, " Calculation of Releases of Radioactive Materials In Gaseous'and Liquid Effluents From Pressurized Water Reactors (PWR-GALE Code)," April 1976.
- 9. Sagendorf, J.F. and Goll, J.T. ,1976: X00D00, Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, (DRAFT). U.S. Nuclear Regulatory Commission, Office of '
Nuclear Reactor Regulation, Washington, D.C.
- 10. Staff of the U.S. Nuclear Regulatory Commission, Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, (Rev.1)," l July 1977.
- 11. Staff of the U.S. Nuclear Regulato.ry Commission, Regulatory Guide 1.109,
" Calculation of Annual Average Doses to Man from Routine Releases of ;
Reactor Effluents for the Purpose of Implementing Appendix I, (Rev.1)," l October 1977. -
?
i
- b. ,
f, Table 1 a
y CALCULATED RELEASES OF RADI0 ACTIVE MATERIALS IN LIQUID EFFLUENTS FROM FT. CALHOUN, UNIT 1 (Ci/yr)
Nuclide Ci/yr Nuclide C[/yr Corrosion 6 Activation Fission Products Products (continued)
Cr-51 6.2(-47 Te-129 3.1 (-4)
FM-54 1.1 (- 3) I-130 6.4(-4)
Fe-55 6.0(-4) Te-131m 1.7(-4)
Fe-59 3. 5 (-4) Te-131 3.0(-5)
Co-58 9 . 7 (- 3) I-131 6. 5 (- 1)
Co-60 9. 5 (- 3) Te-132 3. 9 (-3)
Zr-95 1. 4 (- 3) I-132 1.1 (- 2)
Nb-95 2. 0 (- 3) I-133 1.9(-1)
Np-239 1.4 ( ,4) I-134 8 . 0 (- 5)
Fission Products Cs-134 2. 5 (-2)
Br-83 1.5(-4) I-135 3.0(-2)
Rb-86 3.0(-5) Cs-136 4.7(-3)
Sr-89 1. 2 (-4) Ba-137a 8.0(-3)
Sr-91 2.0 (-5) Ba-140 ,6.0(-5) i Y-91 2. 0 (-5) La-140 6.0(-5)
Zr-95 2.0(-5) Ce-141 2. 0 (-3)
Ko-95 2. 0 (-5) Pr-143 1. 0 (-5)
Mo-99 1.1(-2) Ce-144 5. 2 (-3)
Tc-99m 1.1 (-2) Pr-144 1. 0 (- 5 )
Ru-103 1. 6 (-4) All Others 3. 0 (-5)
Rh-103c 2 . 0 (-5) Total .
Except Tritium 1.0 Ru-106 2. 4 (-3)
Ag-110m 4.4(-5) Tritium Release 290 Te-127m 1. 0 (-4)
Te-127 1. 2 (-4) ,
Te-129m 4.7(-4)
"Exponentia] notation; 6.2(-4) = 6.2 x 10~ ,
- g. .
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Table 2 f
W- CALCULATED RELEASES OF RADIOACTIVE h%TERIALS IN GASEOUS EFFLUESTS FROM
[ FT. CALHOUN, UNIT 1 RELEASE (Ci) PER YEAR Condenser Waste Gas Air Processing Building Ventilation Total Auxiliary Turbine Ejector Nuclides System Reactor a a a a I
a a Kr-83m 1 4 2 a a 1 Kr-85m a 160 a
f- 130 21 1 j Kr-85 a 1 1 a a a i Kr-87 2 7 q 4 a a 1 i Kr-88 a a a a l a a j Kr-89 a 1 73 21 2 Xe-131m 49 a 3 31 24 4 Xe-133m a a 190 5500 3000 310 Xe-133 2000 a a a a a Xe-135m a a 4 19 8 7 Xe-135 a a a a a a Xe-137 a !
a a ' 9 a a 3800 Xe-138 a 3.1(-3) 2.4 (-2) 6.7(-2)
TOTAL NOBLE GASES a 5.6(-4) 3.9 (-2)
I-131 3.5(-2) 9.5(-2) a 7.7(-4) 5.6(-2) 3.3(-3)
I-133 c c 3.4(-4) 7.0(-5) c 2.7(-4)
Co-60 c 7.S( 4) c 6 . 0 (-4) c,,
Co-58 1. 5 (-4) c 7. 5 (-5) 6.0 (- 5) c
- 1. 5 (-5) c Fe-59 c c 2.3(-4)
- 4. 5 (-5) c 1.8(-4)
Mn-54 c 3. 8 (-4) 3.0(-4) e 7 . 5 (-5) c Cs-137 c 2.3(-4) 1.8(-4) c
- 4. 5 (-5) c Cs-134 c 3. 0 (-6) 2.4(-6) e 6.0(-7) c Sr-90 c 1.6 -5)
- 1. 3 (-3) e Sr-89 3. 3 (- 6) c 2.0 -3)
TOTAL PARTICULATE 1 8
a = less than 1.0 Ci/yr for noble gases, less than 10' Ci/yr for iodine, b = exponential notation; 7.0(-5) = 7.0 x 10-5 c = less than 1% of total d = For the purpose of the dose calculation, half of the tritium was assumed to be released from the continuous venting and half in 4-24 hr. purges.
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l
$ Table 3 T
M 7 Principal Parameters Used In Estimating Releases of Radioactive Material In Effluents From Ft. Calhoun, Unit 1 j
l l
1 1500 l Reactor power. level (MWt) 1 0.80 l Plant capacity factor 0.12%"
Operating power fission product source term Primary system 5 2.5 x 10 Mass of coolant (1bs) 36 Letdown rate to makeup and purification (gpm) 1.1 Shim bleed rate (gpm) 100 Leakage rate to secondary system (1bs/ day) 160 Leakage rate to annulus building (1bs/ day) 2 Frequency of degassing for cold shutdowns (per year)
Secondary system 6 6.2 x 10 Steam flow rate (1bs/hr) 3 5.0 x 10 Mass of steam in e~ach generator (1bs) 4 7.8 x 10 Mass of liquid in each generator (1bs) 5 3.8 x 10 Mass of secondary coolant (1bs) -
1,700 Rate of steam leakage to turbine building (1bs/hr) 4 2 x 10 Steam generator blowdown rate (1bs/hr) 6 1.1 x 10 Containment building volume (ft )
37 Frequency of contair. ent purges (per year) .l 5
Turbine building leak rate (gpm)
I l
Iodine partition factors I 0.01 Steam generator internal partition 0.0075 Primary coolant leak to auxiliary building 0.15 Condenser / air ejector (volatile species)
Decontamination factor for ventilation system 10 Charcoal adsorbers Containment building internal recirculation system 2.2 x 10 Flow rate (cfm) 16 Operating period / purge (hrs) )
70%
Mixing efficiency
l
.s
- , Tabic 3 (continued) 1 h
Decontamination factors (liquid wastes) i Boron Recovery Clean Waste Dirty Waste Blowdown i 3
I 10 10 10 10 4 4 2 Cs, Rb 10 10 10 10 _,
4 4 4 3 Others 10 10 10 10 All Nuclides Except Iodine Iodine Boron recovery system 3 2 evaporator DF 10 10 Waste processing system 4 3 evaporator DF 10 10 Anions Cs, Rb Other Nuclides Evaporator condensate ishing demine'ralizer po}OH)
(11 DF 10 10 10 owdown I 2 2 '
Steam generator demineralized (Hb}OH ) DF10 (10)b
~
10(10) 10 (10) a This value is constant and corresponds to 0.12% of the operating power fission product source term as given in NUREG-0017, April 1976.
b Number in parentheses are for second demineralized in series.
J
N Table 4 Summary of Operating Experience for Fort Calhoun Station, Unit No. 1 1976 b , c yg77d, e Average Liquid Effluent Release Data 1975*
Total fission 6 activation products (Ci/yr) 3.1(-1)f 5. 5 (-1) 2.9(-1) 4 . 0 (-1)
-. Total tritium (Ci/yr) 120 120 190 130 Gaseous Effluent Release Data Total noble gases (Ci/yr 430 2200 4800 2000 Total iodine-131 (Ci/yr) 6.6(-3) 2.5(-2) 4.4 (-3) 1.4(-2)
Total tritium (Ci/yr). 2.4 2.5 2.2 2.4 Total particulate (Ci/yr) 3. 3 (-4) 1.6(-3) 3. 6 (-5) 7.8(-4)
"" Evaluation of Fort Calhoun Station, Unit No. 1, in Accordance with 10 CFR Part 50, Apperdix I," Omaha Public Power District, June 1976.
" Fort Calhoun Station, Unit No. 1, Semi-Annual Operating Report for Technical Specification, Section 5.9.4 and Appendix B," Omaha Public Power District, January 1 thru June 30, 1976.
l c" Fort Calhoun Station, Unit No. 1, Semi-Annual Operating Report for Technical Specification, Section 5.9.4 and Appendix B," Omaha Public Power District, July 1 thru December 31, 1976, d
The figures for 1977, based on the first six months of operation, have been doubled to give a yearly activity for purposes of comparison.
'" Fort Calhoun Station, Unit No.1, Semi-Annual Operating Report for Technical Specification, Section 5.9.4 and Appendix B," Omaha Public Power Distriet, January 1 through June 30, 1977.
~I Exponential notation; 3.1(-1) =.3.1 x 10 4
l
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Table 5
.,. Calhoun, Unit No. 1 Relative "4 :entration(X/Q) and Deposition (D/Q)
'_ 4 For Dose Calculations RELATIVE LOCATION SOURCE X/Q (sec/m ) DEPOSITION (m-)
Unit 1 Unit 1 A 6.1E-06 3.2E-08 Nearest ** Site 1.8E-07 Land Foundary B 3.5E-05 C 5.0E-05 2.6E-07 (0.66 mi. WNW) 3.3E-07 D 6.4E-05 E 6.8E-06 3.2E-08 A 3.4E-06 4.5E-08 Nearest Residence 1.8E-07 and Garden B 1.4E-05 C 1.9E-05 2.5E-07 (0.73 mi.S) 3.0E-07 D 2.3E-05 l E 3.7E-06 4.5E-08 1.3E-06 1.6E-08 Nearest Milk 6 A 1.1E-07 Meat Animal B 8.7b-06 -
C 1.3E-05 1.6E-07 '
(0.75 mi. SSW) 2.1E-07 D 1.7E-05 E 1.4E-06 1.6E-08
- " Nearest" refers to the type of location where the highest radiation dose is expected to occur from all appropriate pathwhys.
Source A is reactor continuous release Source B is plant intermittent release (15/yr for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)
Source C is plant intermittent release (24/yr for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
Source D is plant intermittent release (13/yr for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
Source E is turbine building continuous release 4
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- 4 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-285 l OMAHA PUBLIC POWER DISTRICT NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSES AND NEGATIVE DECLARATION The U.S. Nuclear Regulatory Commission (the Commission) has issued Anendment No. to Facility Operating License No. DPR-40, issued to Dnaha Public Power District, for revised Technical Specifications for operation of the Fort Calhoun Station, Unit No.1, located near Omaha, Washington County, Nebraska. The amendments are effective as of the date of issuance.
i These amendments to the Technical Specifications will (1) imple-ment the requirements of Appendix I to 10 CFR Part 50, (2) establish new limiting conditions for operation (LCO) for the quarterly and annual average release rates, and (3) revise environmental monitoring programs to assure conformance with Commission regulations.
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. Prior public notice of these amendments was not required since the amendments do j not involve a significant hizards considerations.
1
_ _ - - - _ - - - _ _ _ l
The Commission has prepared an: environmental impact appraisal for the revised Technical Specificatiores and has concluded that an environ-
- mental impact statement. for- the particular action is not. warranted because there will be no significant effect on the quality of the human environment J beyond that whic'n has already' been predicted and' described in the Commission's Final Environmental ' Statement' for the facility dated August 1972.
For further details with respect to this action, see (1) the application j for amendment dated , (2) Amendment No. to License No. DPR-40, and-(3) the Commission's related Safety Evaluation and Environmental Impact' .
Apprai sal . All of these items are available for public inspection at the Commission's Public Document Room,1717 H Street, N. W.,. Washington, D. C.,
and at the Blair Public Library,1665 Lincoln Street, Blair, Nebraska. A copy.of items (2) and (3) may be obtained upon reouest addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, Attention:
Director, Division of Operating Reactors.
?
f Dated at Bethesda, Maryland this day of FOR THE NUCLEAR REGULATORY COMMISSION l
George Lear, Chief !
Operating Reactors Branch #3 Division of Operating Reactors I