ML19296B934
| ML19296B934 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/15/1980 |
| From: | JERSEY CENTRAL POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19296B927 | List: |
| References | |
| NUDOCS 8002220385 | |
| Download: ML19296B934 (12) | |
Text
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JERSEY CENTRAL POWER & LIGIT CO4PAlW OYSTER CREEK NUCLEAR GENERATIlG STATION (DOCKET NO. 50-219)
PIOVISIONAL OPERATING LICENSE DPR-16 Applicant hereoy requests the Commission to change Appendix A to the above captioned license as follows:
- 1. Sections to be changed:
Sections 3.6 and 4.6
- 2. Extent of Change:
Incorporate 10 CFR 50 Appendix I design objectives for gaseous effluent releases into the Technical Specificctions as Limiting Conditions for Operation.
- 3. Cnsnges requested:
The requested changes are on the attached revised Technical Specification pages 3. 6-2, 3. 6-2a, 3. 6-2b, 3. 6-Sa, 3.6-6, 3. 6-7, 3. 6-8, 3. 6-9, 3.6-10, and 4.6-3.
- 4. Discussion:
This Technical Specification change is the result of a meetirg between JCP&L and the NRC staff held on September 13, 1979 and reported in meeting minutes dated September 28, 1979. The charge will assure that gaseous effluent releases from the Oyster Creek station are kept as low as is reasonably achievable, as defined by 10 CPR 50 Appendix I. It is anticipated that this change is temporary, and will be deleted in tne future af ter enmplete Radiological Effluent Technical Specificatiens (RET 3) are issued for Oyster Creek and the Augmented Offgas System is in regular operation. Revised RETS for Oyster Creek are in preparation and will be subnitted as committed to in our letter lated December 3,1979.
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8002 22c: 3T5 6
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3.6-2 D. Reactor Coolant Radioactivity Tne concentration of the total iodine in the reactor coolant shall not exceed 8.0 uCi/gm. If this specification cannot be met, the reactor shall be placed in the cold shutdown condition.
E. Liquid Radioactive Waste Control Equipment installed for the treatment of liquid wastes shall be used if release of an untreated batch would result in concentrations in execss of 20% of the limits given in Section 3.6.B.(1).
F. Annual Gaseous Release Limits
- 1. Tne average release rate of noble gases frcm the site during any calendar year shall be limited by tne following equations:
for beta air dose:
(3.17 E04) x (7.4 E-08) x ?
1 Qi Ni < = 20 for gamna air dose:
(3.17 E04) x (7.4 E-08) x [ Qi Mi <= 10 Where:
? denotes summation over all isotopes detected 1
3.17 E04 = conversion factor pCi - yr/Ci-sec 7.4 E-08 = X/Q at site boundary 625 m SSE. (Ref. 13)
Qi = Average release rate of isotope i, in Ci/yr Ni = dose conversion factor for beta air dose, mrad - m(3)/ pCi-yr. from Table 3.6-1 Mi = dose conversion factor for gamma air dose, mrad - m(3)/pci-yr from Table 3.6-1
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Technical Specification Change Request No. 79
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3.6-2a
- 2. The average release rates of radiciodines and radioactive materials in particulate form released in gaseous effluents from the site during any calendar year shall be limited by the following equation:
(3.17 E-02) {: {Rii (4.8 E-08 x Qis + 2.3 E-05 x Qiv) +
(Rgi + Rvi) (5.5 E-09 x Qis + 1.0E-07 x Qiv)] (= 15 Where:
I denotes summation over all isotopes detected i
3.17 E-02 = conversion factor, uCi-yr/Ci-sec Rii = dose factor for inhalation, mrem-m(3)/Ci-yr, Table 3.6-2 Rgi = doce factor for ground plane exposure, m(2) -mrcm-sec/
uCi-yr, Table 3.6-2 Rvi = dose factor for vegetation consumption m(2) mrem-sec/
uCi-yr. Table 3.6-2 4.8 E-08 = T./O at 890 m SE for stack releases (Ref. 13) 2.3 E-05 = X/0 at 890m SE for vent releases (Ref. 13) 5.5 E-09 = D/0 at 890m SE for stack releases (Ref. 16) 1.0 E-07 = D/Q at 890m SE for vent releases (Ref. 16)
Qis = average release rate of isotope i from the stack in Ci/yr Qiv = average release rate of isotope i from the vent in Ci/yr -
Note: The Rvi for tritita should be multiplied by X/O rather than by D/O as is done for all otner nuclides.
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Technical Specification Change Request No. 79
3.6-2b Bases: Some radioactive material is released from the plant under controlled conaltions as part of tne normal operation of the facility. Otner radioactive material not normally intended for release could be inadvertently released in the event of certain accident conditions within the facility. Tnerefore, limits have been placed on the above types of radioactive materials to assure not exceeding the limits of 10 CFR 20 for the former type ard the guideline limits of 10 CFR 100 for the latter type.
Radioactive gases frcra the reactor pass through the steam lines to the turbine and then to the main condenser where they are extracted by the air ejector, passed tnrough 30-minute noldup piping ard released via the plant stack. Tne limites of release and radioactive material from the plant stack nave been calculated using meteorological data from an instrumented 400 ft tower at the plant site. Tne analysis of tnis onsite meteorological data snows that tne expected composition of radiogases af ter 30 minutes holdup in the off-gas system, a continuous release of 0.3 C1/sec would not result in a wnole body radiation dose exceeding tne 10 CFR 20 value of 0.5 rem per year. The Holland plume rise model with no correction factor was used in the calculation of tne effect of momentum ard buoyancy of a continuously emitted plume.
Independent dose calculations for several locations offsite nave been made by the ALC staff. Tne metno5 utilized onsite meteorological data developed by the licensee and utilized diffusion assumptions appropriate to tne site. 2ne metnod is described in Section 7-5.2.5 of "Meterology and Atomic Energy - 1968" equation 7.63 being used. Ine results of these calculations were equivalent to tnose generated by tne licensee provided the average gamma energy per disintegration for the assumed noble gas mixture with a 30 minute hold up is 0.7 MeV per disintegration. Based on tnese calculations, a maximum release rate limit of gross activity, except for iodines and particulates with half lives longer than eight days, in tne amount of 0.21 Y curies per secord will not result in offsite annual doses in excess of the limits specified in 10 CFR 20.
The I determination need consider only the average gamma energy per disintegration since the controlling whole body dose is due to the cloud passage over the receptor and not cloud submersion in which the beta dose could be add itive.
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Technical Specification Change Request No. 79
3.6-Sa Tne noble gas releases are controlled so that the beta air dose is less tnan or equal to 20 mrad /yr and tne gama air dose is less tnan or equal to 10 mrac/yr (in acordance with 10 CFR 50 Apperdix I) at the site boundary in the direction with the hignest X/O. A X/Q of 7.4 E-08 at 625 m in the SSE direction was chosen on the basis of the Oyster Creen AppeMix I Analysis'. The X/Q used is tnat for releases from the stack, since almost all noble gas releases are from that source. Tne equation used is B-4 from Regulatory Guide 1.109 Rev. 1 (October 1977) Appendix B. It has been used for gama as well as beta air dose from stack releases in order to simplify the calculations. Tnis simplification is conservative because equation B-4 is based on a semi-infinite cloud dose model rather than the more exact finite cloud model of equation B-1 of Regulatory Guide 1.109.
Tne releases of radioiodines and radioactive materials in particulate form are controlled so that tne thyroid dose to any real person is less than or equal to 15 mrem /yr, in accordance with the design objectives of 10 CFR 50 Appendix I.
The X/O and D/Q values used are for a distance of 890 m in the SE direction which is the location of the nighest potential thyroid dose based on the Oyster Creek AppeMix I Analysis (Ref.12) The pathways considered at this location are stored and fresh fruits and vegetables, innalation and ground plane deposition.
No milk pathway exists at tnis location. Tne meat pathway is insignificant and is not considered (Ref.12). Equations from NUREG-0133 (Ref.14) are used to calculate the dose factors, along witn dose conversion factors from Regulatory Guide 1.109 (Ref.15). Wnere no data was provided in Regulatory Guide 1.109 for thyroid dose conversion factors for certain nuclides, the total body dose conversion factors were used. All calculations are done for a child since tnis produces the highest dose. (Ref.12).
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Technical Specification Change Request No. 79
3.6-6
References:
(1) FDSAR, Volume I, Section IX-3.3.
(2) Licensing Application, Amendment 13, Meteorological Radiological Evaluation for the Oyster Creek Nuclear Power Station Site.
(3) Deleted.
(4- PDSAR, Volume I, Section VII-6.2.3.
(5) Deleted.
(6) FDSAR, Volume I, Section IX-3.1.1 (7) FDSAR, Volune I, Section II-4.3 (8) Licensing Application, Amendment 11, Question 1-4.
(9) Licensing Application, Amendment 11, Question 1-5.
(10) Deleted (11) Licensing Application, Amendment 11, Question IV-8.
(12) Evaluation of the Oyster Creek Nuclear Station to demonstrate conformance to the Design Objectives of 10 CFR 50 Appendix I, May 1976, Table 3-10 page 2 of 2.
(13) Meteorological Information and Diffusion Estimates to Conform witn Appendix I Requirements: Oyster Creex, July 1976, Table 1.3-11 B.
(14) NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, Draft of August, 1978, Pages 30-33 and 36-37.
(15) Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents For tne Purpose of Evaluating Compliance witn 10 CFR Part 50, Appendix I, Tables E-6, E-9, E-13.
(16) Ref. 13, Table 1.3-13B.
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Technical Specification Change Request No. 79
3.6-7 Table 3.6-1 DOSE FACIORS FOR EXPOSURE 'IO A SEMI-ItiFINITE CIIXJD OF 10BLE GASES Nuclide Ni* Mi*
Kr-83m 2.88E-04 1.93E-05 Kr-85m 1.97E-03 1.23E-03 Kr-85 1.95E-03 1.72E-05 Kr-87 1.03E-02 6.17E-03 Kr-88 2.93E-03 1.52E-02 Kr-89 1.06E-02 1.73E-02 Kr-90 7.83E-03 1.63E-02 Xe-131m 1.11E-03 1.56E-04 Xe-133m 1.48E-03 3.27E-04 Xe-133 1.05E-03 3.53E-04 Xe-135m 7.39E-04 3.36E-03 Xe-135 2.46E-03 1.92E-03 Xe-137 1.27E-02 1.51E-03 Xe-138 4.75E-03 9.21E-03 Ar-41 3.28E-03 9.30E-03
- mrad-m3 pCi-yr Source: Regulatory Guide 1.109, Revision 1, October 1977, Table B-1.
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Technical Specification Change Request No. 79
3.6-8 TABLE 3.6-2 THYROID DOSE FACTORS FOR INHAIATION (Rii), GROUND PIANE EXPOSURE (Bgi), AND VEGETATION CONSUMPIJON (Rvi)
NUCLIDE Rii* Fgi** Rvi**
H-3 1.1E03 0 4.0 E03 C-14 6.7E03 0 1.8 E08 Na-24 1.6E04 1.9E07 3.7 E05 P-32 9.9E04 0 1.3 E08 Cr-51 8.5E01 4.7E06 6.5 E04
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Nn-54 9.5E03 1.3E09 1.8 E08 itn-56 3.1E-01 9.0E05 4.2 Fe-55 7.8E03 0 1.3 E08 Fe-59 1.7E04 2.8E08 3.3 E08 Co-58 3.2E03 3.8E08 2.0 E08 Co-60 2.3E04 2.lE10 1.1 E09 Ni-63 2.8E04 0 1.3 E09 Ni-65 1.6E-01 3.0E05 6.6 Cu-64
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1.1 6.lE05 6.8 E03 Zn-65 7.0E04 7.5E08 1.3 E09 Zn-69 8.9E-03 0 3.2 E02 Br-83 4.7E02 4.9 E03 5.7 Br-84 5.5E02 2.0 E05 3.8 E-ll
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Br-85 2.5E01 0 2.6 E-13 Rb-86 1.lE05 3.4 E02 0 Rb-88 3.7E02 3.3 E04 3.1 E-22 Rb-89 2.9E02 1.2 E05 3.5 E-26 Sr-89 1.7E04 2.3 E04 1.1 E09 Technical Specification Change Request No. 79
3.6-9 Sr-90 6.4E06 0 3.2 Ell Sr-91 4.6 2.2 E06 2.1 E04 Sr-92 5.3E-01 7.8 E05 2.9 E01 Y-90 1.1 E02 4.5 E03 6.2 E02 Y-91m 1.8E-02 1.0 EOS 4.2 E-10 Y-91 2.4 E04 1.1 E06 5.0 E05 Y-92 5.8 E-01 1.8 E05 4.5 E-2 Y-93 5.1 1.9 E05 8.5 Zr-95 3.7E04 2.5 E03 7.7 E05 Zr-97 1.6E01 3.0 E06 4.9 E01 Nb-95 6.5E03 1.4 E08 1.1 EOS Mo-99 4.3 E01 4.0 E06 1.9 E06 Tc-99m 5.8 E-02 1.8 EOS 1.6 E02 Tc-101 1.1 E-03 2.0 E04 6.9 E-30 Ru-103 1.1 E03 1.1 E08 5.9 E06 Ru-105 5.6E-01 6.4 E05 3.3 E01 Ru-106 1.7 E04 4.2 E08 9.3 E07 Ag-110m 9.1 E03 3.5 E09 1.7 E07 Tb-125m 1.9 E03 1.6 E06 9.8 E07 Te-127m 6.1 E03 9.2 E04 3.2 E08 Te-127 2.0 E02 3.0 E03 6.9 E03 Te-129m 6.3 E03 2.0 EO7 2.8 E08
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Te-129 7.1 E-02 2.6 E04 7.7 E-04 Te-131m 9.8 E01 8.0 E06 1.1 E06 Te-131 1.7E-02 2.9 E04 1.4 E-15 Te-132 3.2E02 1.0 E08 2.5 E08 I-130 1.8 E06 5.5 E06 7.0 E07 Technical Specification Change Request No. 79
3.6-10 I-131 1.6 E07 1.7E 07 2.4 E10 I-132 1.9 E05 1.2 E06 3.4 E03 I-133 3.8 E06 2.4 E06 3.9 E08 I-134 5.1 E04 4.4 E05 2.7 E-03 I-135 7.9 EOS 2.6 E06 5.0 E05 Cs-134 2.2 EOS 6.8 E09 5.5 E09 Cs-136 1.2 E05 1.6 E02 1.6 E08 Cs-137 1.3 E05 1.0 E10 3.4 E09 Cs-138 5.6 E02 3.6 E05 5.8 E-11 Ba-139 5.4 E-02 1.1 E05 1.4 E-03 Ba-140 4.3 E03 2.1 E07 1.6 E07 Ba-141 6.4 E-03 4.1 E04 2.9 E-23 Ba-142 2.8 E-03 4.6 E04 0 La-140 7.5 E01 1.9 E07 3.8 E02 La-142 1.3 E-01 7.4 E 05 2.4 E-05 Ce-141 2.9 E03 1.4 E07 4.9 E04 Ce-143 2.9 E01 2.3 E06 1.3 E02 Ce-144 3.6 E05 6.9 E07 6.8 E06 Pr-143 9.1 E02 0 7.3 E03 Pr-144 3.0 E-03 1.8 E03 2.6 E-27 IM-147 6.8 E02 8.5 E06 4.5 E03
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W-187 4.3 2.4 E06 1.7 E04 Np-239 2.3 E01 1.7 E 06 1.3 E02
- mrem-m(3)/uci-yr.
- m(2)-mrem sec/uci-yr.
IME: Where no data was available for the thyroid dose factor in in R.G. 1.109, Rev. 1, Tables E-9 or E-13, the total body dose factor was used to calculate Rii or Rvi, es applicable.
Rvi factors for iodines were reduced by half based on the assumption that one-half the iodine released is non-elemental. (Per R.G. 1.109, Rev. 1, Page 26)
Technical Specification Change Request No. 79
, - 4.6-3 (f) If a batch is to be released on an identified radionuclide basis, the analysis shall also in-clude a gamma scan. If gamma peaks different from those determined by previous isotopic analyses are found or if the mixture concentra-tion is greater than 10% of the mixture FPC, a new isotopic analysis shall be performed and record;d.
(3) Environmental Program The environmental program described in Section B.11.6 of Amendment 65 to the Application for a Reactor Oper-ating License shall be conducted. The sampling fre-quencies specified in Table B-II-1 of Amendment 65 shall be adhered to as closely as conditions permit.
C. A sample of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to determine total radioactive iodine content.
D. Liquids contained in the waste sample tanks, floor drain sample tanks, and the waste surge shall be sampled and analy:ed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to determine the total activity in curies unless a tank has been valved out of service after determining its radioactive content.
E. The operability of all equipment installed for the treatment of liquid wastes shall be verified at least once per quarter.
F. The calculations specified in section 3.6.F shall be performed at least once per month.
Basis: The check, test, and calibration requirements are specified to detect possible equipment failure and to show that maximum permissible release rates are not exceeded. The monitors (1) operate continu-ously and by virtue of normal plant operation, the operators daily observe that the instruments are performing. Failure of an instru-ment is evident, because of upscale, downscale, or loss of voltage alarms. The monitor trip points may be readily checked by a built-in pushbutton operated c'ircuit. A portable test source may be affixed to the detector to re-establish calibration. Experience with instru-ment drift and failure modes indicates that the specified test frequency is adequate and consistent with other instrumentation.
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Continuous monitoring of the gaseous and collection of the particulate stack effluents provides the means for determining that the limits of specification 3.6.A are not exceeded and for recording the actual levels of radioactivity that are being released from the stack. The frequencies of filter and cartridge analyses and isotopic analyses are specified to assure proper identification of the isotopes being released. The sampling and analysis of each batch of the radioactive liquid effluent provide the means for determining the release rate to the discharge canal to assure the limits of Specification 3.6.B are not exceeded.
Technical Specification Change Request No. 79
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