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UNITED STAT Es S
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NUCLEAR REGULATORY COMMISSION
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JUN 011983 y
l Docket No.: 50-289 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Operating Reactors, DL FROM:
Daniel R. Muller, Assistant Director for Radiation Protection, DSI
SUBJECT:
ENVIRONMENTAL IMPACT APPRAISAL INPUT FOR RETURN TO SERVICE OF TMI-1 STEAM GENERATORS (TACf 47484)
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{s,f" PLANT NAME: Three Mile Island Unit 1 Q
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LICENSING STAGE: OR DOCKET NUMBER: 50-289
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dlp f, RESPONSIBLE BRANCH: ORB #4 J. VanVliet, PM REQUESTED COMPLETION DATE: May 20, 1983 DESCRIPTION OF RESPONSE: EIA Input for return to service of TMI-l's OTSG REVIEW STATUS: Complete Attached is RAB input to the EIA for TMI-l return to service after the OTSG repair program. A copy of this EIA was delivered to J. VanVliet, ORB #4, DL on May 20, 1983 by M. Wangler, RIS, RAB.
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Daniel R. Muller, Assistant Director for Radiation Protection Division of Systems Integration fc.o g i
Enclosure:
I EIA Input for Return to
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Service of TMI-l dji e [ / o "g d
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g-J. VanVliet
- 0. Lynch M. Wangler C. Hinson C. Willis
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4.0 RADIOLOGICAL ASSESSMENT 4.1 Environmental-Significance of Releases The radwaste t'reatment and effluent control systems installed at Three Mile Island Nuclear Station, Unit No.1, have been previously described in Section 11.0 of the staff's Safety Evaluation Report (SER)Il' dated June 1973 and in Section 111.D of the Final Environmental Statement (FES)2/
dated December 1972.
In addition a detailed evaluation of radwaste treatment systems with respect to the requirements of Appendix I are contained in the Safety Evaluation and Environmental Impact Appraisal _/ dated December 1977'.
3 Based.on more recent operating data at other operating nuclear power reactors which are applicable to Three Mile Island Nuclear Station, Unit 1, and on changes
-in the' staff's calculation models, new liquid and gaseous source terms' have
.been generated to determine 'the radiological impact of operating Three Mile Island Nuclear Station, Unit No.1. The new source terms, shown in Tablest.1 andi.2, Mere calculated using the model and parameters described in In making these determinations, the staff considered waste flow rates, concentra-tions of radioactive materials in the primary system and equipment decontami-nation 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 liquid 3/
and gaseous source terms are given in Table (3 The staff also reviewed the operating experience accumulated at Three Mile Island Nuclear Station, Unit No.1, to correlate the calculated releases given in Tablesti andf.2 with observed releases of radioactive materials in liquid and
l gaseous effluents. Data on liquid and gaseous effluents contained in the Licensee's Semi-Annual Operating Reports Movering the period for January 1976 through-December 1978 were~used. A summary of these releases is given in Table 4.4Three Mile Island Nuclear Station, Unit No..I reached initial criticality on June 5,:1974 and comercial operation in September 1974.and has been shut down since March 1979. Since the staff does not consider data from the first year of operation nor data for the current shutdown period to be representative of the long term operating life of the plant, only the effluent release data from January 1976 through December 1978 were used in this comparison. The calculated annual releases of total mixed fission and activation products in liquid efflu-ents and total noble gases and total iodines and particulates in airborne effluents are within 25% of the geometric means of the reported annual releases from Three
'. Mile Island, Unit 1 for the period 1976 through 1978.
The differences between the reported and calculated values are not considered significant and can be attributed to better fuel performance than was assumed in the calcDlations. The staff believes that the calculational model reasonably
-characterizes the actual releases of radioactive material from this system.. The NUREG-0017 methodology for calculating releases is based,in part, on 0.12% cladding defects and 100 lb/ day primary-to-secondary leakage'. ate averaged over the operating life of the plant. When conditions differ from the above,the releases may be different from those calculated.
The staff has also reviewed a General Public Utilities (GPU) document, " An Assessment of the Three Mile Island, Unit 1, Plant Safety for Return to Service After Stern Generator Repair Top,ical Report" 6_/ dated March 30,1983.
In this document GPU discusses the potential release of radioactive materials from the radwaste system after restart.
In the discussion GPU has assumed two conditions, I
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1.) a f ailed fuel rate cf 0.03%, which is based en present primary coolant sam-
" ples, and a primary-to-secondary coolant leak rate of 24 lb/ day, which is the steam generator repair program leak-rate goal, and 2.) the same failed fuel rate and a 6 gal /hr leak rate that is an adninistrative limit. These conditions could have a significant effect on the air ejector gaseous releases but not on liquid'or other gaseous releases. Air ejector' gaseous releases calculated using NUREG-0017 values-and both sets of GPU values are compared in Table 4.5.
In addition, the staff has reviewed GPU's values and has concluded that GPU's estimates are rea-sonable for the conditions assumed.
4.2 Public Radiation Exposure This section contains conservative estimates of the impacts on the.public from the operation of Three Mile Island, Unit I after restart of the plant.
The major sources of radiation and environmental pathways were considered in preparing this section. Public radiation exposure from plant operation can.be estimated by com-paring the estimated quantities of radioactive effluents with average releases and dose estimates. from prior operation.- Estimates of source terms before shut-down and af ter restart are given in Table 4.4.
On the basis of this comparison, the staff concludes that the offsite environmental impact that may occur wilibe about the same as that which occurred prior to shutdown.
The staff has also estimated the doses to individual members of the public in the area surrounding Three Mile Island, Unit 1 based on the radioactive effluent changes estimated in Table 4.5 and on the calculational methods presented in Regulatory and 1.111./ The results of this analysis are presented in Table 4.6.
8 Guides 1.109 l 7
t The calculated doses based on 0.12% failed fuel are the NRC staff's esti-mated annual doses to an individual living near Three Mile Island, Unit 1.
The calculated doses based on GPU's estimates of primary-to-secondary leakage rat'es provide upper and lower bounds on the maximum individual doses. Although the calculated dose for 3 pathways using GPU's upper bound leak rate would exceed Appendix I design objectives, the technical specifications for Three Mile Island, Unit 1 require releases to be held within Appendix I objectives.
The staff -further feels that its projected leak rate and failed fuel rate will te more representative of actual conditions than GPU's estimates.
In summary the estimated radioactive releases resulting from the restart of Three Mile Island, Unit I will be about the same as those prior to shutdown. -The doses due to these releases will be about the same as those prior to shutdown. Therefore, the radiological impact of restarting the plant will not be signficantly different from the impact prior to shutdown and the quality of the human environment will not be additionally affectep.
4.3 Conclusion and Basis for Negative Declaration The staff'has conc.luded the following:
(1) restart of Three Mile Island, Unit I willnot result in a significant change in radi9 active effluents levels; (2) restart of Three Mile Islan'd, Unit 1 will not result in a significant change in doses to the public; (3) compliance with the present Technical Specifications will adequately control releases such that no appreciable effect on the environment will occur, i
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. REFERENCES 1.) " Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Comission in the Matter of Three Mile Island, Unit," U.S. Atomic Energy Commission, 1973.
2.) " Final Environmental Statement Related to Operation of Three Mile Island Nuclear Station, Units 1 and 2" NUREG-0552, U.S. Atomic Energy Commission, December 1972.
3.) " Safety Evaluation and Environmental Impact Appraisal," contained in memo Collins to Reid, re: OSE Evaluation of Three Mile Island Nuclear Station, Unit No.1, With Respect to Appendix I to 10 CFR Part 50, U.S. Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, December 7,1977.
4.) " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWR (PWR-GALE Code)," NUREG-0017, U.S. Nuclear Regulatory Comission, April 1976.
5.)
" Semiannual Radiological Environmental Monitoring Report " Metropolitan Edison Company, 6 reports for period January 1976 through December 1978.
6.) " Assessment of TMI-1 Plant Safety for Return to Service Aft'er Steam Generator Repair Topical Report 008, Revision 2,"
General Public Utilities, March 1983.
7.)
" Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I,"
Regulatory Guide 1.109, Revision 1, U.S. Nuclear Regulatory Commission, October 1977.
8.)
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Regulatory Guide 1.111, U.S. Nuclear Regulatory Commission, July 1977.
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CALCULATED RELEASES OF RADI0 ACTIVE MATERIALS IN.
~ LIQUID EFFLUENTS FROM THREE MILE 151AND NUCLEAR STATION,. UNIT NO.1 ~
FOR APPENDIX I EVALUATIONS Nuclide~
Ci/Yr Nuclide Ci/yr Corrosion /Actiiration Products Fission Products (continued)
Cr-51 1.9 (-4)
Te-129 8 (-5)
Mn-549 1(-3)
I-130 7(-5)
Fe _ 2 (-4)
Te-131m 4 (-5)
-Fe-59 1(-4)
I-131 5(-2)
. Co'-58 5.8(-3)
Te-132 1.1(-3)
'Co-60 8.9 (-3)
I-132 1.6 (-3)
Zr-95 1.4 (-3)
- I-133 1.7 (-2)
Nb 2(-3)
I-134 1(-5) 6(-5)
Cs-134 2.9 (-2) -
-Np-239-I-135 3.9 (-3)
Fission Products 3.3 (-3)
Cs-136 Br-83 3(-5)
- 3. 5 (-2) -
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'Rb-86' 2 (-5)
Ba-137m 9.1 (-3) 2(-5):
Sr 4 (-5)
Ba-140 1(-5)
Sr-91 1(-5) ta_14o 5~.'2 (-3)
Ho-99 3.9 (-2)
Cc-144 Tc-99m 2.4 (-2)
All Others
- 6 (-5)
Ru-103 1.4 (-4}
- Total, 2.5 (-1)
Ru-106' 2.4(-3) except H-3 5(+2).
H-3 cAg-110m
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Te-127m 3(-5)
Te-127 4 (-5)
Te-129m' 2.3 (-4) f
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CALCULA1TD RELEASES OF -RADIOACTIVE MATERIAL IN i
GASEOUS EFFLUENTS FROM 'INREE. MILE ISLAND NUCLEAR STATION, UNIT NO.1 Ci/Yr Radio d Reactor Auxiliary Turbine Air Decay Nuclide Building Building _
Building Ejector Tanks Total Kr-83m a
a a-a e
a Kr-85m 1
I a
a a
2 Kr-85 320 17
'a 4
'560 890 Kr-87 a-a a
a a
a Kr-88:
2 2-a 1
a 5
- Kr-89 a
- a a
a a
a Xe-131m-54 2
a 1
6
-63 Xe-133m 33 3
a 2
a 38 Xe-133 5500-240 a
150 25 5900 Xe-135m a
a a
a a
a Xe-135 9
4-a 3
a 16 Xe-137 a
a a
.a a
a Xe-138 a
a a
a a
a 6,900 L
TOTAL NOBLE GASES I-131-0.00039' O.0049 0.001 0.031 a
0.037
~1-133 0.00031 0.0053 0.0011 0.032 a
0.039 Mn-54 2.6(-6)b 1.8(-4) c c
4.5 (-5) 2.3(-4)
Fe-59 8.7(-7) 6(-5) e c
1.5(-5)
- 7. 6 (-5)
Co-58 8.7 (-6) 6(-4) c c
1.5(-4) 7.6 (-4)
Co-60 4 (-6) 2.7(-4) e c
7(-5) 3.4 (-4)
Sr-89 2 (-7) 1.3(-5) c-c 3.3 (-6) 1.6(-5)
Sr-90 3.5 (-8) 2.4 (-6) c
.c 6(-7) 3(-6)
Cs-134 2.6(-6) 1.8(-4) e c
4.5(-5) 2.3(-4)
Cs-137 4.4 (-6) 3(-4) e c
7.5 (-5) 3.8(-4) 2(-3)-
TOTAL PARTICULATES 510 H-3 C-14.
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c c
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- Ar-41 25 c
c c
c 25 a = -less than 1.0 Ci/yr/ reactor for noble gases and carbon-14, less than 10
- Ci yr/ reactor for iodine,
~0 b = exponential notation; 2.6(-6) = 2.6 x 10 c = less than 1% of total for this nuclide, d = radionuclides not listed are released in quantities less than those specified in notes a and c from all sources.
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PRINCIPAL PARAMETERS AND CONDITIONS USED IN
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CALCULATING RELEASES OF RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS EFFLUENTS FROM "IEREE MILE ISLAND NUCLEAR STATION, UNIT NO. I 2535 Reactor Power Level (MWt) 0.80 Plant Capacity Factor 0.12%
Failed Fuel Primary System 5
7.4 x 10 Mass of Coolant - (Ibs) 45 Letdown Rate (gpm) 480 Shim Bleed Rate (gp3) 100 Leakage to Secondary System (Ibs/ day) b Leakage to Containment Building 160 Leakage to Auxiliary Buildings (Ibs/ day)
Frequency of Degassing for Cold Shutdowns (per yr) 2 Secondary System 7
1.1 x 10 Steam Flow Rate (Ibs/hr) 3 Mass of Steam / Steam Generator (1bs) 1.4 x 104 Mass of Liquid / Steam Generator (Ibs) 2.7 x 106 2 x 10 Secondary Coolant Mass (Ibs) 3 Rate of Steam Leakage to Turbine Bldg (Ibs/hr) 1.7 x 106 3
2 x 10 Containment Building Volume (ft )
4 Annual Frequency of Containment Purges (shutdown) 20 Annual Frequency of Containment Purges (at Power)
Iodine Partition Factors (gas / liquid) 0.0075 Leakage to Auxiliary Building Steam Generator (carryover) 1.0 Main Condenser Air Ejector (volatile species) 0.1S Decontamination Factors (liquid wastes)
Shim Bleed Miscellaneous Laundry And And Eq. Drain h'aste Chain Hot Shower Drain I
1 x 10 1 x 10 1
4 5
Cs, Rb 2 x 10 1 x 10 I
5 5
Others 1 x 10 I x 10 1
All Nuclides Except Todine Iodine _
4 Miscellaneous (Dirty) Waste 10 10
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Evaporator DF 3
2 Shim Bleed 6 Equipment Drain 10 10
- Evaporator DF Anions Cs, Rb_
Other Nuclides Letdown Coolant h'aste 10 2
10 Demineralizers DF Evaporator Condensate Polishing 10 10 10 Demineralizers DF Condensate-F.eedwatei-Demineralizers 10 2
10 (PONDEX) DF i
2-
- TABLE 3 (C:ntinuad)
. Auxiliary and Fuel Handling Building Charcoal
-Adsorber DF and. Containment Recirculating 10 Cleanup. System Charcoal Adsorber DF (Iodine Removal) e HEPA Filter DF and Containment Retirculating 100 Cleanup System HEPA Filter DF O
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' Table 4.4 Sunnary.of Operating Experience for Three Mile Island.. Unit No.1 i
i Normal. 0per,ations, Ci/yr M'adured Geometric Mean Calculated
-FES Type of radioactive effluent e
e 1976-1978 NUREG-0017 Estimate i
1976 1977 1978 Gaseous Noble Gases 2760 16600 15740 9000 6900 3636 l.
Iodines and Particulates 0.0107 0.0339 0.135 0.037 0.039 0.22 i
Liquid Mixed fission and activation products 0.1.1 0.194 0.614 0.23 0.25 3.0 If J
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Table.4.5' Comparison cf Calculated Annual Gtseous R 1:ases From Three s
Mile Island Unit No. 1, Air Ejector Based on Different Fuel Cladding 0ef' cts and Primary-to-Secondary Leakages e
l Fual..Claftling Defects 0.12%
0.03%
0.03%.
i Primary-to-Secondary Leakage 100 lb/ day 1 lb/hr 6 gal /hr.
t Isotope I-131 0.031 a
0.086 I-133 0.032 0.00017 0.192 I-135 a
0.00018 0.203 Kr-85m a
a 17.
Kr-85 4.
a a
Kr-87 a.
a 59.
. Kr-88 1.
a 31 Xe-131m 1
a-a Xe-133m 2
a 6
Xe-133 15 0.
a 46 Xe-135 3
a 72 a
Less than 1.0 Ci/yr for. noble gases; less than 0.0001 Ci/yr for iodine.
- Releases are based on the primary coolant activity concentrations measured in November 1978,.which are approximately equiyalent to 0.03% fuel cladding defects.
NRC staff estimate.
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4 l-4 Table 4.6 Comparison.of Calculated Annual Maximum Individual Doses From Three'.
Mile Island, Unit 1 Based on Different Fuel Cladding Defects.and Primary-to-Secondary Leakages Appendix I Design Calculated..
Criterion Objectives Doses Fuel Cladding Defects 0.12%
.0.03 %-
0.03%
3 Primary-to-Secondary Leakage 100 lb/ day 1 lb/hr 6 gal /hr.
Liquid Effluents j'pa h a s 3 wem/yr 1.8 mrem /yr 1.8 wen /yr 1.8 mrem /yr D
10 mrem /yr 2.4 wem/yr 2.4 mrem /yr 2.4 w em/yr a
pa hwa s Noble Gas Effluents j
Gamma dose in air 10 mrad /yr 3.1 wad /yr 2.2 mrad /yr 14.1 w ad/yr
. Beta dose in air 20 r ad/yr 8.9 wad /yr 6.8 mrad /yr 18.1 wad /yr tal body of an 5 mrem /yr 1.8 r em/yr 1.8 mrem /yr 2.2 mrem /yr dvj s
nNv dua 15 w em/yr 5.5 mrem /p 4.0 mrem /p.
24 wem/p t-l Radioiodine and Particulates I
Dose to any organ from 15 mrem /yr 6.1 wem/yr 0.5 mrem /yr 16.5 w em/yr all pathways r
Intor-Sffiso Mcmcrandum Jate June 7, 1983 MSS-83-504 UC Gar
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Subject TMI-1:
System to To N. Trikouros Location Parsippany-CHB Manager Safety Analysis & Plant Control The following design information on the decay heat system is provided per your request:
- Piping:
Stainless Steel 304, SCH 40s Class N-2 Design Temperature =
300*F Design Pressure =
400 PSIG
- Pumps:
Worthington Type HN-194, single stage Centrifugal Pumps 1800 RPM, 350 HP Capacity = 3000 GPM @ 350 Ft TDH Design Temperature = 300*F Design Pressure = 470 PSIG
- Heat Exchangers:
Whitlock Shell and Tube Type 6
Q = 125 x lg BTU /HR @ 250*F Q = 30 x 10 BTU /R @ 140*F Tube side design temp.
300 F
=
Tube side design pressure = 470 PSIG
- Alarms /.
Interlocks:
Suction Pressure Alarms @ W385 PSIG Suction Temperature Alarms @ > 285'F Valves DN-Vl/V2 Interlocked (closed) @ 400 PSIG R. Spragg of Mechanical Components is checking with Worthington for information concerning the required NPSH on the pumps and the consequences of running the pumps at temperatures exceeding 300'F.
Please contact me if you require additional information.
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