ML20237B613
| ML20237B613 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/26/1977 |
| From: | Jay Collins Office of Nuclear Reactor Regulation |
| To: | Desiree Davis Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8712170001 | |
| Download: ML20237B613 (25) | |
Text
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SEP 2 61977 ,
Docket Hos. 50-282/306 MEMORANDUM FOR: Don Davis, Acting Chief, Operating Reactors Branch No. 2, )
00R FROM: J. T. Collins, Chief. Effluent Treatment Systems Branch, DSE
SUBJECT:
DSE EVALUATION OF PRAIRIE ISLAND NUCLEAR GENERATING STATION, UNIT NOS.1 AND 2, WITH RESPECT TO APPENDIX ! TO 10 CFR PART 50 Enclosed is DSE's detailed evaluation of the radioactive waste treatment systems installed at Prairin Island, Unit Hos.1 and 2, with respect to the requirements of Appendix 1. The results of our evaluation are contained in the attached " Safety Evaluation and Envirennental Impact Appraisal." !
We have also attached a draft " Notice of Issuance of Amendment to Facility Operating Licenses and Negative Declaration."
Based on our evaluation, we conclude that the radioactive waste treatment
systems installed at Prairie Island are capable of maintaining releases of radioactive materials in effluents to *as low as is reasonably achievable" ~
levels in confomance with the requirements of 10 CFR Part 50.34a, and conforms to the requirements of Sections II.A II.B. II.C, and II.D of , , ,
Appendix I.
On March 29,1977 DSE transmitted to E!Aan NRC Staff Report entitled,
" Application of Cost-Benefit Analysis Requirements of Appendix I to 10 CFR Part 50 to Nuclear Power Plants Whose Applications Were Docketed Before January 2,1971.* This report provides the staff's justifica-tion for using the September 4,1975, amendment to Appendix ! rather than performing a detailed cost-benefit analysis required by Section II.D of Appendix 1. . In our transmittal memo we requested ELD review and recommen-dation as to the most expeditfoes way of incorporating the findings of this report into the licensing process. To date we have not received their comments or recommendations. Following ELD review we will provide you a paragraph to be inserted in the enclosed Safety Evaluation providing justification for using the Septanher 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.
~
. .ORTSINAL SIGFID BY 8712170001 770926 . JOHN T. COLLINS PDR ADOCK 05000282 P PDR John T. Collins, Chief Effluent Treatment Systems Branch Division of Site Safety and
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Don Davis SEP 2 61977 Enclosuret DSE Evaluation ec: H. Denton V. Stallo R. Vollmer K. Go11er D. Jaffee D. Eisenhut W. Kreger l
M. Grotenhuis J. Collins H. Hulman B. Grimes E. Markee ,
F. Congel P. Stoddart DISTRIBUTION:
Docket Files 50-282/306 NRR Reading File DSE Reading File ETSB Reading File ETSB Occket Files 50-282/306 JTCollins l
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SAFETY EVALUATION AND ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. TO FACILITY LICENSE N0. DPR-42 AND AMENDMENT NO. TO FACILITY LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING STATION, UNIT NOS.1 AND 2 DOCKET NOS. 50-282 AND 50-306 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 amendment to Appendix 1( to provide persons who have filed applications for construction permits 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 required by Section 11.0 of Appendix I, if the proposed or installed radwaste systems satisfy the guides on design ojectives 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.I }
Following ELD review of the Generic Cost / Benefit Analysis, 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
in effluents to unrestricted areas "as low as is reasonably achievable", and .l
- 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 Northern States Power Company (NSP) filed with the Commission on June 4, 1976 I4I , July 21, 1976(5) , September 29, 1976(6) , and November 19, 1976( ,
the necessary information to permit an evaluation of the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2, with respect to the requirements of Sections II. A, II.B, and II.C of Appendix I. In these submittals, NSP provided the necessary information to shes conformance with the Commission's September 4,1975 amendment to Appendix I rather than perform a detailed cost-benefit analysis required by Section 11.0 of Appendix I.
By letter dated , NSP submitted proposed changes to Appendix A Technical Specifications for Prairie Islant Nuclear Generating Plant, Unit Nos. 1 and 2. The proposed changes implement the requirements of Appendix I )
to 10 CFR Part 50 and provide reasonable assurance that releases of radioactive materials in liquid and gaseous . effluents are "as low as is reasonably achiev-able" in accordance with 10 CFR Parts 50.34a and 50.36a.
1 DISCUSSION 1 The purpose of this report is to present the results of the NRC staff's 1 detailed evaluation of the radioactive waste treatment systems installed at d
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i Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2; 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 i requirements of 10 CFR Parts 50.34a and 50.36a, 2) to meet the individual dose design objectives set forth in Sections II. A, II.8, and II.C of Appendix I to 10 CFR Part 50, and 3) to determine if the installed radwaste systems satisfy the design objectives proposed in RM 50-2 rather than an individualized cost-benefit analysis as required by Section II.D of Appendix I.
I. Safety Evaluation The NRC staff has performed an independent evaluation of the licensee's pro-posed method to meet the requirements of Appendix I to 10 CFR Part 50. The !
l staff's evaluation consisted of the following: 1) a review of the information provided by the licensee in his June 4,1976, July 21,1976, September 29, 1976, and November 19, 1976, submittals; 2) a review of the radioactive waste (radwaste) treatment and effluent control systems described in the licensee's Final Safety Analysis Report (FSAR);I I 3) the calculation of expected releases of radioactive materials in liquid and gaseous effluent (source terms) for i
the Prairie Island facilities; 4) the calculation of relative concentration (X/0) and deposition (D/Q) values for the Prairie Island site; 5) the calcula-f tion of individual doses in unrestricted areas; and 6) the comparison of the ,
1 calculated releases and doses with the proposed design objectives of RM 50-2 and the requirements of Sections II. A, II.B II.C and 11.0 of Appendix I.
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.4-The radwaste treatment and effluent contro1' systems installed at Prairie Island Nuclear Generating Plant.have been previously described in Chapter 11 of the sta'ff's Safety Evaluation Report (SER) dated April 19, 1968,( }'and in Section 111.0 of the Final Environmental Statement (FES) dated May 1972 Unit Nos.1 and'.2 share two comon gaseous radwaste decay tank systems and share common liquid.radwaste systems for tritiated wastes, nontritiated wastes, j and laundry wastes.
Based on more recent operating data at other operating nuclear power reactors, which are applicable to Prairie Island Nuclear Generating Plant, and on changes in the staff's calculation models, new liquid and gaseous source terms have i been generated to determine conformance with the requirements of Appendix 1.
The new source terms, shown in Tables 1 and 2, were calculated using the model and parameters described in NUREG-0017.III) In making these determinations the staff considered waste flow rates, concentrations of radioactive materials in the primary system, and equipment decontamination factors consistent with
.those expected over the 30 year operating life of the plant for normal operation including anticipated operational occurrences. The principal parameters and plant conditions used in calculating the new liouid and gaseous source terms are given in Table 3.
The staff also reviewed the operating experience accumulated at Prairie Island Nuclear Generating Plant, 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
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- the licensee's Semi-Annual Operating Reports covering the period of 1974 through 1976. A summary of these releases is given in Table 4. The actual release data' provided by the licensee in his reports are the releases for .
Unit Nos.1 and 2 combined, as indicated in Table 4, whereas the calculated releases shown in Tables 1 and 2 are on a per reactor basis.
Prairie Island, Unit No.1, reached initial criticality on December 1,1973, and commercial operation on December 1.4, 1973. Unit 2 reached initial criticality on December 17,.1974, and commercial operation on December 21, 1974. Since the staff does not consider data from the first year of opera-tion to be representative of the long-term operating life of the plant, only !
i effluent release data from January 1975 through December 1976 were used in i
comparing actual releases from the Prairie Island Plant with calculated releases. !
Actual liquid radwaste releases averaged 0.12 Ci/yr/ reactor, not including tritium, which is lower than the staff's calculated release of 0.22 Ci/yr/ reactor.
The difference is not considered to be significant. Tritium in the combined gaseous and liquid actual releases from the plant averaged 670 Ci/yr/ reactor compared to the staff's combined calculated releases of 680 Ci/yr/ reactor.
The staff calculated that 250 Ci/yr/ reactor would be released in liquid effluents while 430 Ci/yr/ reactor would be released to the atmosphere; however, j the licensee's reported actual releases averaged 660 Ci/yr/ reactor in liquid effluents and only 10 Ci/yr/ reactor in gaseous effluents. Based on a review of the licensee's semi-annual release reports, as referenced in Table 4, the staff concluded that the difference in the distribution of tritium between the calculated and actual releases in liquid and gaseous effluents can be
attributed to the staff's assumption of a higher rate.for recycle of processed liquids than was employed by the licensee during the reporting periods used in this comparison. ,
Actual -releases of noble gases in gaseous effluents averaged 980 Ci/yr/ reactor, compared to the staff's calculated release of 6,500 Ci/yr/ reactor. The lower .
level of noble gas releases is attributed to better fuel performance than was assumed in the staff's calculations. Actual releases of iodine-131 in gaseous effluents were also lower than the staff's calculated values and this, too, is attributed to better fuel performance than was assumed by the staff.
. Based on the above evaluation of operating data, the staff believes that the calculational model reasonably characterizes the actual releases of radioactive materials in liquid and gaseous effluents from Prairie Island, Unit Nos.1 and 2. Therefore, the calculated releases given in Tables 1 and 2 were used in the staff's dose assessment discussed below.
The staff made reasonable estimates of average atmospheric dispersion con-ditions for the Prairie Island Nuclear Generating plant using an atmospheric dispersion model for long-term releases" . This model is based on the
" Straight-Line Trajectory Model" described in Regulatory Guide 1.111" }.
The staff assumed that all plant gaseous effluents were ground-level releases.
The staff evaluated non-continuous and intermittent gaseous releases separately from continuous releases. Based on the criteria outlined in Regulatory Guide 4 1.111, the calculations include an estimate of maximum increase in calculated i
.= ,
- p relative concentration deposition due to the spatial and temporal variation i
of the airflow not considered in the straight-line trajectory model. In the 1
' evaluation, the staff used meteorological date collected onsite, between June 1,1971 and May 31, 1972.
Table 5 presents the calculated values of relative concentration (X/0) and relative deposition (D/0) for specific points of interest ~.
The staff's dose assessment considered the following three effluent cate-gories: 1) pathways associated with radioactive materials released in liquid effluents to the Mississippi River, 2) pathways associated with noble gases released to the atmosphere; and 3) pathways associated with radio-iodines, particulate, carbon-14, and tritium released to the atmosphere.
The mathematical models used by'the staff to perform the dose calculations to the maximum exposed individual are described in Regulatory Guide 1.109.(I4)
The dose evaluation of pathways associated with the release of radioactive materials in liquid effluents was based on the maximum exposed individual.
For the total body dose, the staff considered the maximum expos,ed individual to be an adult whose diet included the consumption of fish (21 kg/yr) har-vested in the immediate vicinity of the discharge from Prairie Island, Unit Nos.1 and 2, into the Mississippi River and use of the shoreline for recrea-tional purposes (12 hr/yr). The maximum individual was also assumed to ingest 7301/yr of water from the vicinity of the discharge.
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The dose evaluation of noble gases' released to the atmosphere included a l calculation of beta and gamma air doses at the site boundary sector having the highest dose and total body and skin doses at the site boundary sector .
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having the highest dose.. The maximum doses at the site boundary were found
.at 0.31 miles NNE relative to the Prairie Island Plant.
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 a child whose diet included the consumption of 26 Kg/yr of fresh leafy vegetables, 41 Kg/yr of meat and poultry, and living at a residence 0.5 miles SSE of -the Prairie Island Plant.
Using the dose assessment parameters noted above and the calculated releases of radioactive materials in liquid effluents given in Table 1, the staff calculated the annual dose or dose commitmen't to the total body or to any organ of an individual, in an unrestricted area, to be less than 3 mrem /
q reactor and 10 mrem / reactor, respectively, in conformance with Section II. A {
i of Appendix 1. !
Using the dose assessment parameters noted above, the calculated releases of radioactive materials in gaseous effluents given in Table 2, and the I f
appropriate relative concentration (X/Q) value given in Table 5, the staff I calculated the annual gamma and beta air doses at or beyond the site boundary l to be less than 10 mrad / reactor and 20 mrad / reactor, respectively, in con- l 1
formance with Section II.B of Appendix I. ]
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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/0) 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 1.
The summary of calculated doses given in Table 6 are different from and replace those given in Table V-2 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 on a per reactor basis. As shown in Table 1 the calculated release of radioactive material in liquid effluents is less than 5 Ci/yr/ reactor, excluding tritium and dissolved noble gases. As given in Table 2, the calculated quantity of iodine-131 released in gaseous effluents is less than 1 Ci/yr/ reactor. The calculated doses combined for Unit Nos.1 and 2 are less than the dose design objectives set forth in RM 50-2 and, therefore, satisfy the requirements of Section 11.0 of Appendix I. For a comparison with the site dose design objectives given in Table 3, Column 2, the calculated values given in Table 3, Column 3 should be multiplied by two (number of units on the site).
i CONCLUSION l Based on the foregoing evaluation, the staff concludes that the radwaste treatment systems installed at Prairie Island Nuclear Generating Plant, Unit I
Nos.1 and 2, 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 requirements of 10 CFR Part 50.34a, and therefore, are Jacceptable.
The staff has performed an independent evaluation of the radwaste systems installed at Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2. 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 individual doses will not exceed 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 RM 50-2 and therefore, satisfies the requirements of Section 1I.0 of Appendix I to 10 CFR Part 50.
The staff concludes, based on the considerations discussed above, 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 by operation-in the proposed manner, ard (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.
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l h II. Environmental Impact Appraisal The licensee is presently licensed to possess and operate the Prairie Island:
Nuclear Generating Plant, Unit Nos. I and 2, located in the State of Minnesota, in Goodhue County, at power levels up to 1650 megawatts thermal (MWt) for each unit. The proposed changes to the liquid and gaseous release limits will not result in an increase or decrease in the power level of the Units.
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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 l (FES) for Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2, Docket Nos. 50-282 and 50-306.
The revised liquid and gaseous effluent limits will not significantly change the total quantities or types of radioactivity discharged to the environment from Prairie Island Nuclear Generating Plant.
The revised Technical Specifications implement the ' requirements of Appendix I to 10 CFR Part 50 and provide reasonable assurance that releases 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 objectives in a quarter, the licensee must: (1) identify the causes, (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.
l Conclusion and Basis for Negative Declaration
{
l On the basis of the foregoing evaluation, it is concluded that there would l
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 l
that a negative declaration to this effect is appropriate.
Dated:
l,
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, 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. Letter, L. O. Mayer, Manager of Nuclear Support Services, Northern States Power Company, to Mr. D. L. Ziemann, Chief, Operating Reactors Branch #2, Division of Operating Reactors, U.S. Nuclear Regulatory Commission,
" Appendix I Analysis," June 4,1976.
- 5. Letter, L. O. Mayer, Manager of Nuclear Support Services, Northern States Power Company, to Mr. D. L. Ziemann, Chief, Operating Reactors Branch #2, Division of Operating Reactors, U.S. Nuclear Regulatory Commission,
" Appendix I Filing, Supplement 1," July 21,1976.
- 6. Letter, L. O. Mayer, Manager of Nuclear Support Services, Northern States Power Company, to Mr. D. L. Ziemann, Chief, Operating Reactors Branch #2, Division of Operating Reactors, U.S. Nuclear Regulatory Commission,
" Appendix I Filing, Supplement 2," September 29, 1976. ;
- 7. Letter, L. O. Mayer, Manager of Nuclear Support Services, Northern States Power Company, to Mr. D. L. Ziemann, Chief, Operating Reactors Branch #2, Division of Operating Reactors, U.S. Nuclear Regulatory Commission,
" Appendix I Filing, Supplement 3," November 19, 1976.
- 8. Northern States Power Company, " Final Safety Analysis Report, Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2," Docket Nos. 50-282 and 50-306, Minneapolis, Minnesota, January 28,1971 (amendments 1 - 25).
April 1973.
- 9. " Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Commission, in the Matter of Northern States Power Company, Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2, Goodhue County, Minnesota," Docket Nos. 50-282 and 50-306, Washington, D.C., September 28, 1972.
2-
)
- 11. NUREG-0017, " Calculation of Releases of Radioactive Materals in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code),"
April 1976.
- 12. Sagendorf, J.F. and Goll, J.T.,1976: X0000Q, <rogram for the Meteorological Evaluation of Routine Effluent Releases at Nuclee.r Power 1 Stations,(DRAFT). U.S. Nuclear Regulatory Commission, Office of J
Nuclear Reactor Regulation, Washington, D.C.
- 13. 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," March 1976. I
- 14. Staff of the U.S. Nuclear Regulatory Ccamission, Regulatory Guide i 1.109, " Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Implementing Appendix I," March 1976.
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TABLE 1 CALCULATED RELEASES OF RADI0 ACTIVE KATERIALS IN l
LIQUID EFFLUENTS FROM l PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 RELEASE, Ci/Yr/ Unit Nuclide Ci/Yr/ Unit Nuclide Ci/Yr/ Unit Activation Products Cr-51 5(-5) I-131 6.7 -
Mn-54 1(-3) Te-132 5.2 Fe-5E 5(-5) I-132 6.5 -
Fe-59 3 -5) I-133 6 -2 Co-58 4.5 -3) I-134 2 -5 Co-60 S.8 - Cs-134 1.4 -2 Zr-95 1.4 - I-135 1.6 -2 Hb-95 2- Cs-136 4.3 -
Np-239 2 -5) Cs-137 2.5 -
Ba-137m 8.1 -
Ce-144 5.2(-3) -
Fission Products !
All others 6(-5)
S r-89 1(-5) Total, Mo-99 2(-3) except Tritium 2.2(-1 )
- Tc-99m 4.3(-3)
Ru-103 1.4 -4) Tritium 250 Ci/yr Ru-106 2.4 -3) -
Ag-110m 4.4 -4)
Te-127 3 -5)
Te-129m 4 -5)
Te-129 3 -5)
I-130 2.7 -4) '
Te-131m 4-5)
- _: o
TABLE 2 CALCULATED RELEASES OF RADI0 ACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT N05, 1 AND 2 (Ci/yr/ reactor)
Building Ventilation D y Ej or Nuclides Reactor Auxiliary Turbine Tanks Exhaust Total Kr-83m 3 a a a a 3 Kr-85m 22 3 a a 2 27 Kr-85 23 2 a 160 'l 180 Kr-87 6 2 a a a 8 Kr-88 31 5 a a 3 39 Kr-89 a a a a a a Xe-131m 29 2 a a 1 32 Xe-133m 70 5 a a 3 78 Xe-133 5400 380 a a 240 6000 Xe-135m a a a a a a Xe-135 90 9 a a 5 100 Xe-137 a a a a a a.
Xe-138 1 1 a a a 2 Total Noble Gases 6500 I-1 31 7.6(-3)b 4.l(-3) 2.2(-3) a 2.5(-2) 3.9(-2) 1-133 9.9(-3) 6.1(-3). 2.6(-3) a 3.8(-2) 5.7(-2)
Total Iodines 9.6(-2)
Mn-54 2.2(-4) 1.8(-4) c 4.5(-5) c 4.4(-4)
Fe-59 7.4(-5) 6(-5) c 1.5(-5) c 1.5(-4)
Co-58 7.4(-4) 6(-4) c 1.5(-4) c 1.5(-3)
Co-60 3.4(-4) 2.7(-4) c 7(-5) c 6.8(-4)
Sr-89 1.7(-5) 1.3(-5) c 3.3(-6) c 3.3(-5) ,
Sr-90 3(-6) 2.4(-6) c 6(-7 c 6(-6)
Cs-134 2.2(-4) 1.8(-4) c 4.5(-5 c 4.4(-4) :
Cs-137 3.8(-4) 3(-4) c 7.5(-5 c 7.5(-4)
Total Particulate 4(-3)
C-14 1 a a 7 a 8 H-3 - - - - -
430 Ar-41 25 a a a a 25 a = Less than 1 C1/yr for noble gases and carbon-14, less than 10~4 Ci/yr for iodine.
b = Exponential notation; 7.6(-3) = 7.6 x 10 -3 c = Less than 1% of total for this nuclide.
TABLE 3 PRINCIPAL PARAMETERS AND CONDITIONS USED IN CALCULATING RELEASES OF RADI0 ACTIVE MATERIAL IN LIQUID AND GASE0US EFFLUENTS FROM PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 a
Reactor Power Level (MWt) 1720 Plant Capacity Factor 0.80 Failed Fuel 0.12%
Primary System Mass of Coolant (1bs) 2.3 x 10 5 Letdown Rate (gpm) 40 Shim Bleed Rate (gpm) 1.2 Leakage to Secondary System (lbs/ day) 100 c
Leakage Leakage toto Auxiliary Containment Building (lbs/ day)
Buildings 160 Frequency of Degassing for Cold Shutdowns (per year) 2 Secondary System Steam Flow Rate (lbs/hr) 7.1 x 10 63 Mass of Steam / Steam Generator (1bs) 5 x 10 Mass of Liquid / Steam Generator (1bs) 8.3 x 10 35 Secondary Coolant Mass (1bs) _
8.2 x 10 3 RateofSteamLeakagetoTurbineBldg(1bs/hr) 1.7 x 10 3 Blowdown Flow Rate (1bs/hr) 30 x 10 Containment Building Volume (ft3) 1.2 x 10 6 Annual Frequency of Containment Purges (shutdown) 4 Containment Low Volume Purge (cfm) 4,000 Iodine Partition Factors (gas / liquid)
Leakage to Auxiliary Building 0.0075 SteamGenerator(volatilespecies) 1.0 Steam Generator (nonvolatile species) 0.01 Main Condenser Air Ejector (volatile species) 0.15 Decontamination Factors (liquid wastes)
Shim Bleed & Eq. Drain Tritiated Wastes Aerated Wastes Blowdown I 1 x 10 4 1 x 10 4 1 x 10 65 1 x 10 2I 5 5 Cs, Rb 1 x 10 1 x 10 2 x 10 1 x 10 Others 1 x 10 5 1 x 10 5 1 x 10 7 1 x 10 2 Evaporatog All Nuclides Except Iodine Iodine Aerated Waste System 4 3 Evaporator DF 10 10 Shim Bleed & Equip. 2 3
Drain Evaporator DF 10 10 Demineralizers Anions Cs, Rb Others Primary Coolant Letdown Demineralizers, DF 10 2 10 2
Radwaste Demineralizers, DF 10 (10)d 2(10) 102 (10)
Boron Recycle System Feed Demineralizers, DF 10 2 10 Evaporator Condensate Polishing Demineralizers, DF 10 10 10
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TABLE 3 (continued) 1 l
Decontamination Factors (gaseous wastes)
Containment Purge Charcoal Adsorber DF (Iodine removal) 10 Gaseous Systems HEPA filter DF (particulate removal) 100 l
I a
Prairie Island, Unit Nos.1 and 2 are currently licensed at 1650 MWt; I and the licensee made his Appendix I evaluation at a power level of )
1650 MWt. The staff has made its evaluation at the stretch power level of 1720 MWt on the assumption that the licensee will request permission ,
to operate at the higher power level at some future date. '!
b This value is constant and corresponds to 0.12% of the operating power fission product source term as given in NUREG-0017, April 1976.
cl %/ day of the primary coolant noble gas inventory and 0.001%/ day of the primary coolant iodine inventory.
I d
For demineralizers in series, or if a demineralized is used in series with an evaporator, the DF in parentheses should be used for the second demineralized or for a demineralized downstream of an evaporator.
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TABLE 4
SUMMARY
OF LIQUID AND GASE0US RADI0 ACTIVE EFFLUENTS I FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 (Releases for 1974-1976, Unit Nos. 1 and 2 Combined) I i
1975(I) 1976(2) l (Ci/yr) (Ci/yr)
Liauid Releases Total Activity (except Tritium) 0.45 0.012 Tritium 760 1900 Gaseous Releases Noble Gases 2200 1700 Iodine-131 0.018 0.011 Tritium 10 33 Particulate 0.0036 0.00022 (I)From " Prairie Island Nuclear Generating Plant, Semi-Annual Operating Report No. 4," January 1, 1975 to June 30, 1975, and " Prairie Island Nuclear Generating Plant, Semi-Annual Operating Report No. 5," July 1, 1975, to December 31, 1975.
(2)From " Prairie Island Nuclear Generating Plant, Effluent and Waste Disposal Semi-Annual Report," January 1,1976 through June 30, 1976, and " Prairie Island Nuclear Generating Plant, Effluent and Waste Disposal Semi-Annual Report," July 1,1976 through December 31, 1976.
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- UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-282 AND 50-306 NORTHERN STATES POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSES AND NEGATIVE DECLARATION The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment Nos. and to Facility Operating License Nos. DPR-42 and DPR-60, respectively, issued to Northern States Power Company, for revised Technical Specifications for operation of the Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, located in Goodhue County, Minnesota. The amendments are effective as of the date of issuance.
These amendments to the Technical Specifications will (1) implement the requirements of Appendix I to 10 CFR Part 50, (2) establish new l
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 applications for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amendmended (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 setforth in the license amendments. Prior public notice of these amendments was not required since the amendments do not involve a significant hazards consideration. -
The Commission has prepared an environmental impact appraisal for the revised Technical Specifications 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 beyond that which has already been predicted and described in the Commission's Final Environmental Statement for the facility dated April 1973.
For further details with respect to thi.s action, see (1) the application for amendments dated , (2) Amendment Nos, and to -- - - -
License Nos.' DPR-42 and DPR-60, and (3) the Commission's related Safety Evaluation and Environmental Impact Appraisal. 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 Environmental Library of Minnesota, Minneapolis, Minnesota, 55414. A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, Attention: Director, Division of Operating Reactors.
Dated at Bethesda, Maryland this day of FOR THE NUCLEAR REGULATORY COMMISSION
{
Don Davis, Acting Chief Operating Reactors Branch #2 l Division of Operating Reactors I
l . . . _ _ _ . _ _ _ _ _ . .