ML20248D974
| ML20248D974 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/27/1989 |
| From: | Morgan W COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8904120096 | |
| Download: ML20248D974 (61) | |
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f._q L 3 ./ _ 'N Commonwealth Edison j h. i 4 c,ne First N:tional Plan, Chicago, lllinois
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\V/ Address Rep!y to: Post Offic; Box 767 .
Chicago, Illinois 60690 - 0767 g ) U.S.' Nuclear Regulatory Commission Attn: Document Control Desk Hashington,-DC 20555
Subject:
Dresden Nuclear Power Station Unit 1 Response to Request for Additional
- Information: Decommissioning Plan, Technical Specifications and Emergency Plans NRC Docket No. 50-010
Reference:
(a): P.B. Erickson letter to H.E. Morgan dated July 7, 1988.
~(b): P.B. Erickson letter to H.E. Morgan dated September 27, 1988.
(c): P.B. Erickson letter to H.E. Morgan dated January 7, 1989.
Dear Sir:
References (a), (b), and (c) requested additional information concerning Dresden Unit 1 Emergency Plan, Technical Specifications and Decommissioning Plan. The following attachment provides the requested additional information, with the exception of the information pertaining to the Emergency Plan require-ments. This information will be submitted April 10, 1989, and was discussed and concurred with Mr. Pete Erickson of your staff on March 23, 1989. If you have any questions regarding this matter, please contact this office.
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Very truly yours, 4 4/V1 H. E. Morgan Nuclear Licensing Administrator i 1m Attachment 91 cc: P.B. Erickson - NRR I( Resident Inspector - Dresden ; 8904120096 090327 PDR ADOCK 05000010 > P PDC
O r p o 3 l i ATTACHMENT 1 l i l 1 _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ b
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.1 Q 1- Yaur letter dated February 25, 1977 described the storage of radioactively contaminated soil in an excavation at the Dresden Unit 1 HPCI and waste treatment facilities. Indi-cate the present status of this contaminated area, tabulate concentrations vs location and the total inventory for each radionuclides. Also, provide direct radiation measurements of soil in micro R/hr above background, as measured one meter from the surface of the contaminated soil area. Esti-mate the maximum radiation levels that would exist if the layer of . uncontaminated soil were removed. State your schedule for removal of the , contaminated soil such that levels of radioactivity would be acceptable for release to unrestricted access. Currently, we have accepted . 5 micro R-hr above natural background for gamma emitting radionu-clides as measured one meter above an unshield soil surface.
A1 300 cubic yards of contaminated soil was placed in an on-site excavation, east of the radwaste building and south of the waste treatment facility. The storage area covers an area of approximately 5,000 ft2, and the contaminated soil forms a layer approximately 1.5 ft thick. The soil is covered with a minimum of 1 ft of clean gravel to prevent dispersion of the soil. The contaminated soil contains the following calculated activity based on 1977 data, decay corrected as of 3/1/89: Total Radionuclides Inventory (C1) Concentration (DC1/ cram,1 CS-137 3.04 E-2 78 pCi/gm ; CS-134 1.7 E-4 0.44 pCi/gm CO-60 1.0 E-2 26 pC1/gm MN-54 5.2 E-8 1 0.01 pCi/gm As of 3/1/89 the maximum calculated exposure rate one' meter .; above the contaminated soil, if uncovered, would be 45 micro R/hr above background. Direct measurement of the exposure rate above the soil cannot be reliably made because other , sources of radiation that are present in the general area. However, based on the shielding provided by the one ft thick layer of gravel that covers the soil, the estimated exposure rate would be approximately 9 micro R/hr at one meter above the surface. The contaminated soil will be dealt with at the time of station dismantling. Until that time the area will remain covered with gravel, the area posted, and normal radiation protection procedures implemented.
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I Q .1 Wa undarstand that much . of tha liquid wsste produced from the 1984 I
. . chemical cleaning of the primary loop is still on-site in unprocessed ]
form. Plans for processing and disposing of this waste should be I provided, along with information on' volume, activity by nuclide, waste cicssification, and chelate content. Any necessary change to the process control program, the quality control program, -or the transportation . packaging program 'should be submitted. The plan for disposal of the chelate chemical cleaning waste should demonstrate assurance of disposal at an arid LLW disposal site prior to termination of access to such facilities. Similar information should be submitted for any other significant quantity of radioactive waste currently on-site. Costs for dealing with the significant quantitica of stored wastes should be accounted for in the overall decommissioning cost analysis. l
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A2 Processing of the NS-1 is tentatively scheduled to begin after approval is received from the NRC for the stability test report. It is planned to use the US Ecology Bitumen System for processing. The NS-1 waste will be disposed of via burial in Richland, WA. The resin contaminated with NS-1 will be disposed of via burial in Beatty, Nv. Volume of NS-1 waste: 1400 cubic feet Volume of resin : 700 cubic feet Activity by nuclide : NS-1 radionuclides content sample analysis was done on a composite sample created by mixing the individual samples taken from the three NS-1 storage tanks: Tanks 115 A,B,C It was assumed that the nuclide distribution will be the same in all three tanks because all the waste was generated during one cleaning operation. See attached sheets for nuclide concentration. Waste classification: NS-1 waste: Class C Resin : Class A Chelate Content : NS-1 waste: 37% chelate Resin : 0.9% chelate US Ecology has developed a Process Control Plan (PCP) for NS-1. This PCP has been submitted to the NRC for approval. The US Ecology PCP contains the sampling, analysis, and formulation determination by which stabilization of the NS-1 waste will be assured. No changes to the Quality Control program are anticipated. No changes to the transportation packaging program are anticipated. The only other "significant" quantity of radioactive waste currently on-site would be that associated with the unit 1 spent fuel pool. This waste will be disposed of as it is generated. 3311a l l \ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ .. _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ l
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l 03 wastes to be produced in preparation for the " dormancy" l period should be fully characterized in Section 2. I' Information should be provided on volume, activity content by nuclide, concentrations, processing plans, waste : classification, and chelate content if any. Mixed wastes should be identified if present. This should be done for each contaminated waste source listed in Section 2.1.1, "SAFSTOR" Operations, and for each waste stream during the dormancy period described in Section 2.1.2, " Dormancy". ! A3 The primary radioactive wastes that have been generated, or are projected to be generated during the preparation period for SAFSTOR, result from cleaning of the spent fuel pool (SFP) surfaces, from disposal of used filter media from the chemical cleaning system, and from cleaning miscellaneous 1 sumps and other contaminated areas within Dresden-1. Additionally it can be expected that a small quantity of dry active trash will be generated in support of these , activities. Table A3-1 summarizes the types and character- ) istics of the wastes that will be generated during the SAFSTOR preparation phase. Taese wastes will be generated on a one-time basis. None of these wastes are expected to be mixed wastes or have a chelate content. Wastes resulting from the cleaning of the SFP surfaces have already been generated. The resulting warte has been sampled and is currently undergoing analysi':. However, preliminary data is available from three samples of crud that were taken from the SFP floors prior to the cleaning. This data is given in Table A3-2. During the SAFSTOR period, radioactive wastes are projected to arise from operation of a SFP water filtration demineral-izer system which is scheduled to be installed in the near future. Table A3-3 presents the proj ected annual radio-active waste generation for operation of this system. Additionally, a small amount of dry active trash is expected to be generated in support of SFP water cleaning and from periodic plant maintenance, inspection, and surveillance activities. Water to be drained from reactor systems will be transferred to Dresden Units 2 and 3 for recycling. 1
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TABLE A3-1 PROJECTED PREPARATION FOR SAFSTOR RADIOACTIVE WASTE GENERATION Unprocessed Disposal Projected Generation Source Type Vol. (ft 3) Vol. (ft )310CFR61 Class Process Plan SFP Surface
- Sludge 240 358 B/C ** . Dewatered and packaged in hign Cleaning integrity container.
Chemical Cleaning Charcoal Filter 600 600 At Dewatered and packaged in high System Filters Medic integrity container. Misc. Cleaning of Sludge / Debris / 120 178 8t Solidified with cement. Sumps and Other Dust Areas Misc. Dry Actise Cloth, Paper, I I A Compaction Trash etc.
- Already generated
** Precise classification to be made upon receipt of analytical data, t Proj ected e Unknown small quantity
L . l V t TABLE A3-2a RADIOACTIVE SAMPLE ANALYSIS - DRESDEN 1 SFP CRUD-(a) i Sample Location: Transfer Pool j' Reference Date: 2/23/88 Measured Concentration (uC1/om) Measured Concentration (uC1/cm) Nuclide Valve % Uncer. ** Nuclide Valve % Uncer. ** Oross Alpha 6.46E-03 10 129-! <1.7E-05 Gross Beta 5.17E+01 10 131-I <7.9E-03 3-H 8.0E-04 14 134-Cs 6.20E-02 10 l 14-C 6.8E-05 19 137-Cs 2.72E+00 10 32-P (6.0E-05 140-Ba/La (4.1E-04 51-Cr (2.1E-02 141-Ce (4.3E-05 54-Nn (2.2E-03 144-Ce/Pr <1.1E-04 55-Fo 7.13E 10 234-U (2.1E-06 57-Co (8.5E-03 235-U (1.4E-06 58-Co (2.3E-03 237-Np+242Pu <2.5E-06 59-Fe (6.1E-03 238-U (7.9E-07 59-H1 2.4E-02 19 238-Pu 1.72E-03 10 60-Co 1.87E+00 10 239,240-Pu 9.93E-04 10 63-Ni 1.65E-00 10 241-Pu 5.3E-02 11 65-Zn (5.4E-03 241-Am 1.15E-03 10 89-Sr 2.6E-03 21 242-Cm 5.2E-06 13 90-Sr 1.52E-02 10 243,244-Cm 6.78E-04 10 94-Nb (2.1E-04 95-Nb (2.5E-03 95-Zr <3.9E-03 99-Tc (2.4E-05 103-Ru <2.0E-03 106-Ru/Rh (1.7E-02 10am-Ag <2.2E-03 110s-Ag <3.4E-03 124-Sb (7.4E-04 125-Sb (1.9E-02
- Analysis data obtained from Science Applications International Corp.; sample analysis reports for SAIC Sample Nos. 15498, 15691 and 15497.
** Two standard deviation counting error.
_______________________y t 1 TABLE A3-2b l RADIOACTIVE SAMPLE ANALYSIS - DRESDEN 1 SE'P CRUD (a) Sample Location: Transfer Pool-Ditch Reference Date: 4/25/88 . Measured Concentration (uCt/am) Measured Concentration (UC1/cm) Nucli de Valve % Uncer. ** Nuclide Valve 4 Uncer. ** 1 Orcss Alpha 2.10E-01_ 10 131-1 <1.5E0-01 Gross Beta 5.75E+01
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. 10 134-Cs- (9.3E-02 3-H <3.9E-03 137-Cs 1.42E+01 10 14-C <4.0E-04 140-8a/La <3.4E-02 32-P (1.5E-03 141-Ce <1.5E-03 51-Cr <7.2E-01 144-Ce/Pr (4.6E-03 54-Mn <1.2E-01 234-U 1.1E-04 41 55-Fo 2.10E+01 10 235-U <1.6E-05 57-Co <5.3E-02 237-Np+242Pu '(1.8E-04 58-Co <1.2E-01 238-U <2.9E-04 59-Fo <2.9E-01 238-Pu 3.56E-02 10 ,
59-Ni <6.3E-01 239,240-Pu 6.04E-02 10 60-Co 7.36E+01 10 241-Pu 2.1E+00 11 63-Ni- 4.27E+01 10 241-Am 5.80E-02 10 65-Zn <2.9E-01 242-Cm 2.4E-04 14 89-Sr <3.7E-03 243,244-Cm, 8.93E-03 10 90-Sr 6.77E-01 10 94-Nb <8.2E-04 95-Nb (1.2E-01 95-Zr <2.1E-01
'99-Te <5.5E-04 103-Ru (9.3E-02 106-Ru/Rh <9.6E-01 108m-Ag (9.6E-02 110m-Ag <1.9E-01 125-Sb <7.9E-01 129-! <3.6E-04 Analysis data obtained' from Science Applications Interna-tional Corp. ; sample analysis reports for SAIC Sample Nos.
15498, 15691 and 15497.
** Two standard deviation counting error.
l _ _ _ - _ - - _ _ - - - _ _ - - _ _ __ _ __-____-__ ___-_ ---_ -___-.__-__ ______ ___ _ - --_-_-_- - -___-__ --_ --- - __-__ - - _______-______ - _ a
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.. .- f TABLE A3-2c RADIOACTIVE SAMPLE ANALYSIS - DRESDEN 1 SFP CRUD (a)
. Sample Location: Storage Fuel Pool Reference Date: 2/23/88 Measured Concentration (uC1/gm) Measured Concentration (uC1/cm)
Nuclide Valve % Uncer. **- Nuclide Valve % Uncer. " Gross Alpha 1.68E-03 10 131-1 (8.8E-03 Gross Beta 3.78E+00 10 134-Cs 7.12E-02 10 3-H 1.0E-03 15 137-Cs 3.40E+00 10 14-C 163E-03 10 140-8a/La (6.0E-04 32-P <1.2E-04 141-Ce (6.9E-05 51-Cr <2.2E-02 144-Ce/Pr <1.7E-04 54-Mn <1.5E-03 234-U 1.9E-06 46 55-Fe 3.89E-01 10 235-U (6.5E-07 57-Co <1.3E-03 237-Np+242Pu <1.9E-06 58-Co (1.7E-03 238-U '1.7E-06 50 59-Fo <3.7E-03 238-Pu 4.78E-04 10 59-N1. (1.7E-02 239,240-Pu 1.90E-04 10 60-Co 5.80E+01 10 241-Pu 1.0E-02 12 63-Ni -3.45E+01 10 241-Am 3.11E-04 10 65-Zn <3.5E-03 242-Cm <5.0E-07 89-Sr. (2.2E 243,244-Cm, 1.81E-04 10 90-Sr 6.67E-02 10 94-Mb <3.8E-05 95-Nb <1.7E-03 95-Zr <2.7E-03 99-Tc <4.1E-05
'103-Ru <2.1E-03 106-Ru/Rh <1.5E-02 108m-Ag <2.5E-03 110m-Ag <2.2E-03 125-Sb (2.2E-02 129-I <3.0E-05
- Analysis data obtained from Science Applications International Corp. sample analysis reports for SAIC SamNo. 15498, 15691 and 15497.
** Two standard deviation counting error.
e e p . l l .; i l TABLE A3-3 l PROJECTED' ANNUAL SAFSTOR RADIOACTIVE WASTE GENERATION Unprocessed Disposal Proj ected Generation Source Type Vol. (ft 3) Vol. (ft 3) 10CRF61 Class. Process Plan
- 1. SFP Filtration . Filter Cartridges 240 358 8 Resins dowatered, Domineralizer Spent Rosin filters encapsu-System lated with cement.
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- 2. Misc. Dry Active Cloth, Paper, etc. *
- A Compaction
. Trash
- Unknown small quantity.
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.______.______.__._______._______________________________.__.__._____._________________________._.__________________..__________.-_.____]
.t 1
Questions 4 through 12 were addressed in the H.E. Morgan letter to the NRC dated November 21, 1988, and were answered as follows:
"The internal funding method described in Section 4.3 of the ,
Decommissioning Program Plan for the Dresden Nuclear Power Station ' Unit 1 is in compliance with a 1980 Order of the Illinois Commerce Commission and the Proposed Nuclear Regulatory Commission Regulations published in the Federal Register on February 18, 1985. However, the l internal funding method is not in compliance with the final regulations' ' of the Nuclear Regulatory Commission regarding~ General Requirements for ' Decommissioning Nuclear Facilities (10 CFR Part 50.75), effective July 27, 1988 nor with the State of Illinois Senate Bill 1615 signed into law on September 12, 1988. The Company is presently in process of formulating a nuclear decommission external funding plan for all of its nuclear generating units which will comply with the State of Illinois Law, the related forthcoming Accounting and Rate Orders of the Illinois Commerce Commission, the requirements of Section 468A of the Internal Revenue Code and the final Regulations of the Nuclear Regulatory Commission. The Company intends to submit an acceptable decommissioning and financial assurance plan to the Nuclear Regulatory Commission after the related provisions of the pending Illinois Commerce Commission Orders are known which is expected to be well before the July 26, 1990 deadline as prescribed in 10 CFR Part 50.33(k)(2) of the Final NRC Regulations." Therefore. Questions 4 through 12 will be submitted at a later date. i 0057T
4 Q 13 A local area map defining the unrestricted areas used for off-site dose calculations for gaseous and liquid effluents should be included in the TS. In addition, a site map clearly defining the restricted area pursuant to 10 CFR 20.3(a)(14) should be included. A 13 The attached Figure A13-1 " Unrestricted Area Boundry" defines the unrestricted areas used for off-site dose calculations due to gaseous and liquid effluents and defines the 10 CFR 20.3(a)(14) restricted area. This boundary, which is also the Dresden property line, encloses a
" restricted area" for which CECO maintains control for the purpose of controlling radiation exposure of personnel.
This area is patrolled by CECO security.
4 FIGURE A13-1 UNRESTRICTED AREA BOUNDARY C M LDCh CORODOR
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Q 14 a) Provide a table showing Dresden Unit 1 personnel exposure experiences for the years 1978 thru 1988 indicating the j man-rem exposures by plant cystem, regardless of how these ! exposures were obtained (e.g., during normal operations, maintenance, repair or refueling activities) and by whom (e.g., by plant operations personnel, plant maintenance personnel, contractor / vendor personnel, etc.). b) Provide a similar table for the SAFSTOR period. Provide discussion which confirms that your ALARA program will maintain the state-of-the-art for reducing personnel exposures to a minimum. A 14 The following provides a summary of Dresden Unit 1 personnel exposure history for the period covering the latter half of 1985 through the end of 1988. This time period demonstrates typical personnel exposure at the facility, as it now exists and as it will exist at the start of the SAFSTOR period. Exposure history prior to 1985 is not relevant for demonstrating present conditions or proj ecting future personnel exposure because: (1) approximately two Cobalt-60 half-lives have elapsed since 1978 and, (2) in 1984 a chemical decontamination of the NSSS was performed, which reduced the radionuclides inventory. Table A14-1 provides a summary of annual personnel exposures by plant location. This table provides, by location, the total man-mrem received, total man-hours expended, and average exposure rate for the area. Table A14-2 provides a summary of annual personnel exposures by seven work task categories. This table provides, by the ! work task, the total man-mrem received, total man-hours expended, and average exposure rate for performing the I tasks. The work tasks include decontamination, inspection, maintenance, operations, routine tasks, sampling, and radiological survey activities. Exposure data by system, as requested, cannot be provided because personnel working in a particular area may receive their exposures from multiple systems in the area. Table A14-3 provides an estimate of annual personnel exposure to be incurred during the first year of the SAFSTOR period. As the SAFSTOR period progresses, radiation levels will decrease in time due to radioactive decay. this i estimate is based on dose rate data provided in Tables A14-1 ' and 2. The ALARA program which will maintain means for reduction of pers"nnel expsorues during the SAFSTOR period is descirbed in the Question No. 23 response.
I TABLE 14-1
SUMMARY
OF ANNUAL PERSONNEL EXPOSURES BY LOCATION Average Area Total Man-Hours Exposure Rate Year location Total Man-mRom Expended mRom/ Hour 1985(a) Turbine Building 332 42 7.9 Sphere 10 48 0.6 Chem. Clean. Bldg. 4480 4429 1.0 Fuel Building 2698 487 5.5 Boiler House 195 146 1.3 Decon Building 235 84 2.8 Rad Waste Bldg. 731 103 7.1 Laundry Facility 109 37 2.9 Calibration Facility 3 7 0.4 Tank Fars 20 40 0.5 General Area (b) 5472 3771 1.5 TOTAL 14,305 9,194 1986 Turbine Building 49 6 8.2 Sphere 534 154 3.5 Chem. Clean. Bldg. 727 1013 0.7 Fuel Butiding 2787 520 5.4 Rad Waste Building 2264 349 6.5 General Area (b) 4967 3658 0.7 TOTAL 11,328 5,700 1987 Turbina building 102 99 1.0 Sphere 252 66 3.8 Chem. Clean. Bldg. 186 132 1.4 Rad. Waste Building 1505 1054 1.4 Laundry Facility 244 62 3.9 Tank Fars 112 33 3.4 General Area (b) 2650 1666 1.6 TOTAL 6651 3467 (a) Partial Year (6/85 - 12/85) i (b) Multi-area exposures i
TABLE 14-1 (Continued) q
SUMMARY
OF ANNUAL PERSONNEL EXPOSURES BY LOCATION A/erage Area Total Han-Hours . Exposure Rate Year location Total Man-mRom Expended mrem / Hour 1988- Turbine Building 550 328 1.7
$phere 3478 1182 2.9 Chem. Clean. Bldg. 328- 125 2.6 Fuel Building 5361 1524 3.5 Rad. Waste Building 652 140 4.7 Laundry Facility 195 87 2.2 Tank Fars 105 11 9.5 General Area (b) 1851 1337 1.4 TOTAL 12520 4734' (a) Partial Year (6/85 - 12/85)
(b) Multi-area exposures I l
L' TABLE 14-2 ANNUAL PERSONNEL PERSONNEL EXPOSURES BY WORK TASK' Average Area Total Man-Hours Exposure Rate Year location Total Man-mrem., Expended mRen/ Hour 1985 (a) . Decontamination 394 141 2.8 Inspections 639 169 3.8 Maintenance 8313 5249 1. 6 - Operations 410 189- 2.2 Routine 3809 3164 1.2 Sampling 158 63 2.5 Surveys 582 219 2.7 l 1986 Decontamination 206 83 2.5 Inspections 541 67 8.1 Maintenance 2774 534 5.2 Operations 1430 271 5.3 Routine 5138 4260 1.2 Sampling 178 96 1.9 Surveys 1061 -389 2.7 1987 Decontamination 5 4 1.3 Inspections 652 167 3.9 Maintenance 2130 500 4.3 Operatons 208 59 3.5 Routine 1808 1535 1.2 Sampling 91 6 15.2 Surveys 1757 1196 1.5 1988 Decontamination 425 404 1.1 Inspections 1228 359 3.4 Maintenance 6740 2265 3.0 Operations 1653 276 6.0 Routine 1989 1286 1.5 Saapling 353 100 3.5 Surveys 132 44 3.0 (a) Partial year (6/35 - 12/85)
i: TABLE 14-3 i i PROJECTED ANNUAL PERSONNEL EXPOSURES DURING SAFSTOR Average Task Exposure . Proj ected Task Rate. mrem / Hour- Man / Hours ~ Man-mrem Decontamination '2.0 0 0 Inspections 5.0 320 1600 Maintenar.co 3.5 830 2905 Operations 4.0 0 0 Routine 1.0 0 o Sampling 6.0- 40 240 surveys 2.5 320 800 TOTAL 5,545 _ _ - _ - - _ _ . _ m.________m. _ _ _ _ _ .---__-__-__m_ _ _ _ _
l Q 15 Revise Section 5.8 to show the design features of the moni-toring system for the SAFSTOR decommissioning. Include the off-site and on-site environmental monitoring stations. A 15 The Dresden Unit 1 effluent monitoring system / environmental monitoring program during the SAFSTOR period will consist of Unit 1 chimney air exhaust monitoring, chemical cleaning facility air exhaust monitoring, Unit 1 service water effluent monitoring, and the Dresden (Units 1, 2& 3) site's common environmental monitoring program. The following describes the features of the monitoring programs currently in use and which will be in use during the facility SAFSTOR period. Unit 1 Chimney Air Exhaust. Noble gases (NGs) released from the Unit 1 chimney are monitored primarily by a micropro-cessor-controlled off-line system which provides continuous monitoring of an effluent sample stream taken from a point high in the chimney where good mixing is ensured. The effluent 1as sample is drawn past several detectors viewing a sample volume. The low and medium range NG detectors view the same volume, while the high range NG detector views (externally) a section of one int /.1 stainless steel tubing as its sample volume. The low range is monitored by a beta scintillation detector (Eberline Model RDA-3A), while both the medium and high range measurements are performed by an energy-compensated GM detector (Eberline Part No. TUGM5). Background subtraction is provided for all three ranges. Sample flow is provided by a stainless steel diaphragm vacuum pump and an air flow regulator. Sample flow is set at approximately 60 liters per minute and monitored by a solid-state flow sensor. In addition to the noble gas moni-toring, the gas sample is routed through a filter paper on which particulate are depositee., then through a charcoal cartridge which traps iodines. The microprocessor-controlled gaseous effluent monitoring system (Eberline SPING-4) is located in the sphere and turbine building vent fan exhaust room. The monitoring system's control terminal, strip chart recorder and annunciating panel are located in the Unit 2 and 3 control room on Panel 923-7. Each NG channel has a strip chart recorder with a log-scale display ranging from 10 counts per i
~
o minute to 1,000,000 counts per minute. The control terminal ) l provides immediate access to current detector activity, I operating parameters, and history files for any detector. Cnanges in detector operational status are automatically printed and audibly acknowledged. The SPING-4 micropro-cessor's memory has battery backup to maintain all operating parameters and history files for any detector in the event of a system power loss. The SPING-4 monitoring system is supplied with 120-Vac. The Eberline SPING-4 effluent monitoring system's part number is SPING-4 (RM 2/3-1798). Chemical Cleaning Fagility Air Exhaust. The chemical cleaning facility air exhaust is monitored for radioactive particulate and iodines. The exhaust is monitored by continuous sampling with sample media (particulate filter and lodine cartridges) being exchanged and analyzed on a weekly basis. Unit 1 Service Water. Monitoring of the facility's service water will be provided by an off-line grab sampling system. Environmental Radiological Mon.itorina. Environmental i radiological monitoring is currently performed jointly for Dresden Units 1, 2 and 3. During the Unit 1 SAFSTOR period, the existing monitoring program will be maintained. Briefly the environmental radiological monitoring program consists of 17 airborne (particulate and lodine) sampling locations, 69 direct gamma (TLD) locations, 11 water borne sampling locations and 3 ingestion media (milk and fish) sampling locations. Table A15-1 presents a description of the expo-sure pathways and media sampled, the sampling or monitoring locations, the sampling or collection frequency and the type and frequency of analysis. The three accompanying figures (A15-1, -2 and -3) illustrate the monitoring ano sampling locations, which are keyed to the table's location codes.
t , f j TABLE A15-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM Exposure Pathway Sagle or Sampling or Type and Frequency and/or Sample Monitorina locations Collection Frequency of Analysis l 1.' Airborne a. On-site and Near Field Continuous sampler operation Radiodine Canister with particulate. filter 1-131 analysis bi-Radiciodine and D-01 On-site Station 10.6 mi WNW(1.0 km) exchange weekly and radio- weekly. Particulate 0-02 On-site Station 2 0.3 ml NE (0.5 km) iodine canister exchange biweekly Particulate Sampler Gross beta analysis following filter change. Gamma isotopic anal-ysis on quarterly composite (by loca. tion) Samp11no Train Test and maintenance weekly
- b. Far Field D-07 Clay Products, 2.0 ml S (3.2 km) Continuous sampler operation Radiciodine Canister l D-08 Prairie Parks, 4.0 ml SW (6.4 km) with particulate filter 1-131 when analyses 0-09 Coal City, 7.5 ml S (12.0 km) exchange weekly and are made D-10 Goose Lake radiolodine canister exchange Village, 3.8 al SSW (6.1 km) biweekly Particulate Sampler D-11 Morris, 8.0 ml WsW (12.9 km) Gross beta when 0-12 Lisbon, 10.0 ml HW (16.1 km) analyses are made D-13 Minocka, 4.5 ml N (7.2 km) 0-14 Channahon, 3.5 al HE (5.6 km) Samp11no Train 0-15 Jollet,12.5 m1 ENE (20.1 km) Test and maintenance D-16 Elwood, 8.0 ml E (12.9 km) weekly 0-17 Wilmington, 8.0 m) SE (12.9 km)
- 2. Direct Radiation a. At Air Samp11na Sites Quarterly Gamma dose quarterly Same locations as fixed air sampling locations in Item 1.
- b. Inner Rina 0-101-1, 1.0 ml N (1.6 km)
D-101-2, 1.0 ml N (1.6 km) D-102-1, 1.3 al NNE (2.1 km) 0-102-2, 1,3 al NNE (2.1 km) ~ ' D-106-1, 0.9 al ESE (1.4 km) D-106-2, 0.9 al ESE (1.4 km) I
..__1__.________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ ._. _____..__I
1 TABLE A15-1.(Continued) { Exposure Pathway Sample or Sampling or Type and Frequency and/or Sample Monitoring locations Collection Frequency of Analysis
- 2. Direct Radiation b. Inner Rino Cont'd.
o n t ' d .
D-109-1, 0.8 ml 5 (1.3 km) 0-109-2, 0.8 ml 5 (1.3 km) ; D-110-1, 0.6 m1 SSW (1.0 km) -{ D-110-2, 0.6 al SSW (1.0 km) D-112a-1, 0.8 ml WSW (1.3 km) 0-112a-2, 0.8 ml WSW (1.3 km) 0-112b-1, 0.9 ml WSW (1.4 km) 0-112b-2, 0.9 ml WSW (1.4 km) 0-113-1, 0.9 ml W (1.4 km) D-113-2, 0.9 ml W (1.4 km) 0-115-1, 0.8 ml NW (1.3 km) 0-115-2, 0.8 ml NW (1.3 km) 0-116-1, 1.0 ml NNW (1.6 km) 0-116-2, 1.0 ml NNW (1.6 km)
- c. Outer Ring D-201-1, 4.5 ml N (7.2 km) 0-201-2, 4.5 ml N (7.2 km) 0-202-1, 5.0 ml NNE (8.0 km)
D-202-2, 5.0 ml NME (8.0 km) 0-203-1, 4.5 ml NE (7.2 km) D-203-2, 4.5 ml NE (7.2 km) 0-204-1 5.0 ml ENE (8.0 km) 0-2004-2, 5.0 ml ENE (8.0 km) 0-205-1, 4.2 al E (6.8 km) D-205-2, 4.2 ml E (6.8 km) D-206-1, 3.5 ml ESE (5.6 km) D- A6-2, 3.5 al ESE (5.6 km) n .7-1, 4.5 m1 SE (7.2 km) i 1
-407-2, 4.5 ml SE (7.2 km)
D-208-1, 5.0 al SSE (8.0 km) 0-208-2, 5.0 ml SSE (8.0 km) 0-209-1, 5.0 ml 5 (8.0 km) 0-209-2, 5.0 ml S (8.0 km) D-210-1, 4.8 al SSW (7.7 km) 0-210-2, 4.8 al SSW (7.7 km) 0-211-1, 5.0 ml SW (8.0 km) D-211-2, 5.0 ml SW (8.0 km) D-212-1, 4.8 m1 WSW (7.7 km) 0-212-2, 4.8 ml WSW (7.7 km) D-213-1, 4.5 ml W (7.2 km) D-213-2, 4.5 ml W (7.2 km) 0-214-1, 4.5 ml WNW (7.2 km) _ _ ___._._____.__-_-__.-_____2 __
1
+ ,. i TABLE A15-1 (Continued)
Exposure Pathway Sample or Sampling or Type and Frequency and/or Sample Monitorino locations Col'ection Frequency of Analysis
.2. Direct Radiation b. Outer Rino Cont'd.
(Continued) ] D-214-2, 4.5 ml WNW (7.2 km) D-215-1, 5.1 ml NW (8.2 km) , 0-215-2, 5.1 ml NW (8.2 km) D-216-1, 4.8 ml NHW (7.7 km) -] 1 0-216-2, 4.8 ml NNW (7.7 km)
- 3. Waterborne
- a. Surface D-21, Illinois River at EJ&E RR Bridge, Weekly Gamma isotopic on 1 al W (1.6 km) monthly composite D-22, Illinois River at Morris Water Works, 8 ml W (12.9 km)
- b. Laks Wat g D-28, Dresden Pool at Illinois River Weekly Gross isotopic and 0.5 ml W (0.8 km) gross beta analysis D-34A, Cooling Lake, Dresden Road (,rossing weekly 2.6 ml S (4.2 km)
D-348, Cooling Lake, County Line Crossing, 3.0 ml SSE (4.8 km)
- c. Ground /Well D-23, Thorsen Well, 2.0 ml N (3.2 km) Quarterly Gamma beta, gross Water D-35, Dresden Lock & Dam, 0.5 ml W (0.8 km) alpha and trittur.
qt.trterly
- d. Coo 11no Water D-18, Inlet, Unit 1, at Station Weekly Gross beta analysis Sampl e D-19, Discharge, Unit 1, at Station weekly D-20, Discharge, Unit 2.3, at Station
- 3. Shoreline D-27, Dresden Lock & Dam, 0.5 ml W Annually Gamma isotopic Sediments (0.8 km) annually
- 4. Incestion
- a. Milk D-25 Vince 81ros Far,11.5 ml SSW Weekly: May to 1-131 analysis on each (18.5 km) October sample D-26, Trotter's Dairy, 6.1 ml S Monthly: November (9.8 km) to April
- b. Fish D-28, Dresden Pool of Illinois River Semi-annually Gama Isotopic analysis (0.5 mi W (0.8 km) on edible portions of each sample
I . l FIGURE A15-1 FIXED AIR SAMPLING AND TLD SITES AtQ GUTER RING TLD LOCATIONS l / =
=
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os W w?
$ l 'O 12 sie i 2i6: 2 b8 /
2,.., misi
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- a
. ris.: 2 ,,,,.i N '
l 80 2i31 8, l.. e e 205 2 1 11 s
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ris.: 2,,. , ._10 l, 7 1
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DAESOEN STATION 1......... - . . _ . _ - UNITS 1. 2 & 3 FIXED AIR SAMPLING AND TLD SITES AtiC OUTER RING TLD LOCATIONS
I
. 1 I
FIGURE A15-2 INNER RING TLD LOCATICNS A.ND NEAR STATICN WATER 3 AMPLE :,c;A;;;n3 i j l i lj l ! 0-
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. gi g, c
V a
/
; $ CECO Row s f. =n;2 2 115 1 115 2 M21 ig,,, g
'C
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! Cemetery
**d Towpath Rosi
, Og28 Dresden Island Loca t and Dam
! enot* *yet ry'l 'TD.27,7 & 35 D.20 * .
3 E .' Dl". ' Dresden Nuclear f* el
$ll;s Power Station C'e%
'u ususi 2 e,
s .. o.18 e ,,,+ s*/ t % E$ Is n2 2 2 l hDOG Q ">' u e2s' " -l, , , , , i
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-2 Late so.o o JoH*t
+ prairie
*4 State 9 g i:s.i 8
T, iir. p, vacnt Crwo
'"" 'D.33 f General Electric Co. 'es 2 d c aa 33 d i o tj2 'we 0 1 2 0 ometers l ORESDEN ST ATION U NITS 1. 2 & 3 l !NNER RING TLD LOCATIONS l AND NEAR STAT!CN WATER SMPLE LOCATIONS I
i l L _
o FIGURE A15-3 MILK SAMPLE LOCATIONS AND MORE DISTANT WATER SAMPLE LCCAT!ONS
' h
/g s D
~
e, 4: 4
$ %s s f/ s#
b c xxv@i..o f s j ll 5 p' O O i i.so y, g f D.34
- D.23
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7_ D.22 A&B D 31
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,. *D.26 j l-
- m. t u /
e , , i i
- D.25 Coal City '
- 9
- Olyo r o
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,o s io ua..
? '? DRESDEN ST ATION
'?.""***'
. m o - - ,
i U NIT S 1, 2 & 3 MILK SAMPLE LOCATIONS AND HORE DISTANT WATER
, . SAMFLE LOCATIONS l
l { l L--_______________________---_-___. _ _ - . _ .I
Q 16 In accordance with ' Section 1'2. 5, NUREG-0900, confirm that.- your. health ~ physics procedures for performing bicassgy during the SAFSTOR period will conform to ' the recommenda-tions of Regularory Guide 8.26, " Application of Bioassay for Fission and' Activation Products," or submit' equivalent bioassay criteria. A 16 Bioassay for Dresden Units 1, -2 and 3 is provided by a-single common program for the Dresden' site. That is, personnel that may work at Unit i during the SAFSTOR period will be subject to the same bioassay program that Units 2 and 3 will use. This site bioassay program is currently in-use and is acceptable to the NRC. l _______.____._..__-_._____.._-_.____m _ _ _ . _ . - _ _ _ _ _ _ _ _______ _
- * ' Q 17 Are you contemplating 'spsnt - fue1 ~ pin consolidation onsite . prior to shipment'to the Federal High Level Waste Repositorv? . If so, provi.de estimates of the; expected occupational. radiation c.cposures ' involved, of ' the. volume of low ~ 1evel waste generated, and . the contribution ~ of this activity to gaseous and liquid radioactivity ' releases to the environment.
i
.. . I A 17 Since there are no plans to increase the storage. capacity of the. !
existing Unit 1 fuel pool, onsite consolidation of Dresden Unit .1 ! spent fuel is.not planned. However, we understand that the Department I of' Energy. plans to consolidate the fuel either at a Monitored Retrievable Storage facility or at the federal geologic repository. 3311a
Q 18 Describe the treatment and monitoring to be provided for effluents from the refueling building ventilation exhaust, liquid waste storage tankage vent exhaust, hot lab vent exhaust, hot machine shop vent exhaust, radwaste treatment building ventilation exhaust, the high-level solid radio-active waste storage vaults, the low-level waste storage building, the low-level waste handling building, and any. other pathway for the release of radioactive materials. A 18 Ventilation exhaust from the turbine building, reactor sphere, off-gas filter building, blow down flash tank, fuel building and the controlled areas of the access and ware-house buildings (i.e., hot lab, hot machine shop and laundry) are treated by filtration through pre-roughing filters and high efficiency particulate absolute filters prior to discharge. Filtered exhaust air from these areas are routed to the 300 ft high Unit 1 chimney for discharge. Discharges from the unit one chimney are monitored. for radioactivity as was previously described by the Question 15 response. The liquid waste storage tank is vented to the atmosphere. Effluent from this vent is filtered through glass-wool HEPA filters. This effluent is not monitored for radioactivity. Ventilation exhaust from the radioactive waste treatment building is treated by filtration through roughing and HEPA filters. The air is discharged from a 10 ft high exhaust stack located en the building roof. The exhaust is not monitored for radioactivity. The ventilation exhaust from the low-level waste handling building (dry waste storage building) is treated by filtration and discharged to the atmosphere by roof top ventilation units. This effluent is also not monitored. The dry waste storage pit, which consists of the high and low level radioactive waste storage vaults, are not ventilated structures, as such there is no exhaust to be treated or monitored. The chemical cleaning facility ventilation exhaust is i treated by filtration with roughing and HEPA filters. The filtered exhaust is monitored as was described by the Question 15 response. L____._.________._.__ _ _ _ _
Q 19 Provide an estimate of the number of employees required for the SAFSTOR decommissioning time perioo. Provide a breakdown by permanent and contractor workers of the employees required. If this number is not expected to be relatively stable, give estimate of changes anticipated. A 19 Dresden Station has approximately 740 Commonwealth Edison employees on-site with approximately 250 contractor employees on-site on average. This number of contractors varies with refueling outage ! schedules. Workers required for SAFSTOR will be appropriated from this existing work force. As noted in Table 14-3, above, Dresden projects approximately 1500 man hours / year for SAFSTOR. ALso, Dresden expects these numbers to remain f airly stable over the SAFSTOR period. t l l l 3311a I l
Q 20 Provide a description of fission and corrosion product sources in the spent fuel pool (SFP) water from: movement of fuel from the core into the pool, and (a) (b) defective fuel stored in the pcol. Include a listing of the radionuclides and their concentrations (expressed in uC1/mL) expected during decommissioning. The radionuclides of interest should. include 58 o, C 60Co, 134Cs. A 20 All spent fuel at the Dresden Unit i facility is located in the spent fuel pool. Therefore, there will be no movement of spent fuel from the core to the pool. The spent fuel pool water is sampled on a quarterly basis. The latest sample analysis results (2/6/89) indicate that Cobalt-60, Cesium-137 and Cesium-134 are present. Cobalt-58 was not detected. are as follows: The concentration of these radionuclides Cobalt-60 8.8 E-5 micro C1/ml Cesium-134 1.5 E-3 micro C1/ml Cesium-137 9.4 E-2 micro C1/ml The activity in the SFP water comes from two likely sources. The Cobalt-60 in the SFP water arises from the surface of items in the SFP upon which activated corrosion products have been deposited. (Deposition occured while these items were in the NSSS, the primary items being the spent fuel assemblies.) These deposits came off in time and were suspended or dissolved into the SFP water or deposited onto SFP surfaces. The second source, the Cesium fission products, originated from leaking spent fuel. The fission products in the pool water may come from fuel that is currently leaking or had leaked in the past and has now contaminated surfaces in the same manner as Co-60. l l O
Q 21 Provide a description of radioactive materials that may become airborne as a result of failed fuel and evaporation (e.g., 85Kr, and 3H, respectively). The radionuclides i description should include calculated or measured concentra- ' tions expected during decommissioning. A 21 During the Dresden-1 SAFSTOR period airborne radioactive materials (H-3 and Kr-85) are not expected to be present in significant or detectable concentrations. The SFP water currently (2/25/89) has a Tritium concentration of 8.07 E-4 microcuries per ml. This concentration is less than 1% of the 10CFR20, Appendix B, Table 1 concentration for Tritium in water. To illustrate-the low magnitude of expected airborne Tritium concentration in the fuel building during the SAFSTOR period, a worst case estimate was calculated. If the fuel building ventilation was stopped during the summer and the ambient air temperature in fuel building rose to 90*F, and the air in the fuel building attained 100% relative humidity, and all moisture in the air being water from the SFP, the airborne concentration of Tritium would be 2.8 E-8 Ci/ml. This worst case concentration represents only 0.56% of the 10CFR20, Appendix B Table 1 concentration for Tritium in air. During the SAFSTOR period, when the ventilation system is functioning and tr._ ambient temperature in the fuel building is between 60-70'F, the Tritium concentration would be a fraction of the worst case concentration. Airborne Krypton-85, like the Tritium, is not expected in the fuel building during SAFSTOR. This conclusion is based upon current Unit 1 chimney monitoring for Nobel gases, where no Nobel gases are detected. (Fuel building exhaust air is discharged through the Unit 1 chimney.) This conclusion is additionally based on past monitoring in the SFP area where again Nobel gases were not detected. Given the age of the fuel, it is likely that any significant I leakage of Krypton-85 from defective fuel has already occurred. l 1 i j I 1
Q 22 Provida a - description of tha dosa rate at the surface of the pool-water from the fuel assemblies, control' rods, burnable . poison rods or . any miscellaneous materials that may be stored in the pool. A 22 A radiation dose rate survey was conducted on March 9 and 17,1989 - to determine dose rates above the spent fuel pool surfoce. The average dose rate at 1 f t . above the water surf aces in the range of 30. - 50 mrem /hr, exclusive of hot spots. A hot spot of 280 mR/hr . was found near the drain to the radwaste collector. Figure A22-1, "SFSP Surface Dose Rate Survey Results ," presents the results of the March 17, 1989 survey. 3311a
i FIGURE A22-1 SFP SURFACE DOSE RATE SURVEY RESUITS D-1. FUEL POOL Grids are 5 ft x 5 ft A B C D E F G H I
*2
*1 l 1 45 40 40 35 55 42 40 43 60 N 2 38 40. 40 38 40 46 40 39 39 3 38 40 40 35 40 40 40 39 37 4 40 - 40 40 35 45 40 40 40 35 40 40 38 39 5
6 40 36 39 40
*1 68 mR/hr above a channel / temperature guage
*2 280 mR/hr near overflow holes to Radwaste collector due to debris Note: all readings are taken approximately 1 ft above the surface of the water in mR/hr.
1 1
4 Q 23 Discuss the manner in which occupational exposure will be kept ALARA during the decommissioning. Include the need for and the manner in which cleaning of the crud on the SFP walls will be performed to reduce exposure rates in the SFP area. A 23 The ALARA program policy for Dresden Unit One, as well as Units 2/3, is described in Commonwealth Edison Company's document " Policy and Procedures for Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable (ALARA)," November 1, 1981. ALARA plans specific to Unit 1 decommissioning include reduction of radioactive sources and reduction of manpower usage in controlled areas. The spent fuel pool walls have been hydrolazed and vacuumed to reduce radioactive crud and thereby reduce area exposure rates. Currently, a filtration-demineralization system is being installed which will further remove and/or prevent ; buildup of radioactive crud in the SFP. Commonwealth Edison is also currently investigating the feasibility of decon-taminating the walls of the reactor sphere and other areas as appropriate to reduce radiation and radioactive materials. During the SAFSTOR period, personnel access to Unit 1 will be restricted to those activities required to maintain the unit in a safe condition or to maintain those systems that will be required for the ultimate dismantling. These include quarterly radiation surveys, inspections, or maintenance activities.
'. . Q 24 Tha radionuclida inv:ntory data providsd are insuf ficient to datermine waste processing requirements, waste volumes, waste classification, or identify special wastes such as oils or chelates. As requested !
earlier in connection with Section 2, a complete characterization of waste generated, processed, and disposed during the SAFSTOR phase of . I decommissioning is needed. A 24 NRC Question 24 requests the same information as Question 3. Dresden believes the response to Question 3 addresses this question. . l 1 l i l l 3311a l l
Q 25 In Section 7.1 it is stated that waste volume estimates assume a 40 to 60 percent void fraction in the waste containers. To maximize stability in a low-level warte disposal unit it is necessary to minimize voids in waste packages to the extent practicable. It is unclear whether the assumed voids are based on incomplete container filling or on the odd-shaped nature of rubble and scrap wastes. In any case efforts should be made to minimize voids so that long-term stability can be achieved at disposal sites, a A 25 The statement in Section 7.1 of the Decommissioning Plan which pertains to the assumed void fraction of 40 to 60 percent is used for the purpose of estimating the cost and disposal volume of concrete rubble and pipe. When material such as concrete rubble is put into containers, a void fraction is present because of the irregularity of the rubble pieces. The same would be true for pipe which has an internal void volume, and when atacked with other pipe will produce an external void by virtue of a pipe's cylindrical geometry. In the worst case, it could mean that a container could have as much as a 40 to 60 percent void. However, this void cannot cause subsidence at the disposal facility. The containers would be completely filled with concrete rubble or pipe, which by virture of the nature of the material, will produce an interstitial void fraction even with the containers filled to volumetric capacity. When material such as concrete rubble or pipe is packaged into a , i container, the material is in essence interlocked in place and becomes noncompactable. Whenever possible existing void fraction is utilized to package material such as concrete scabbling dust or paper / plastic wastes which will fill the interstitial void space. When feasible small bore pipe can be vested into larger bore pipe to increase packaging efficiency. In the case of containers which are loaded with concrete rubble or pipe and the containers weight capacity is reached befcre its volumetric capacity is filled, light weight ccmpacted dry active waste would be used to fill the remaining container void. In the case of dry active wastes, such as paper, cloth, or plastic, the material would be compacted to the maximum extent possible. I 1
l l Q 26 Note ~1 on P. 53 states that most wastes from the dormancy period can be' disposed at Dresden Units 2 or 3. We assure this means that wastes generated would be processed and shipped from Units 2 or 3 rather than disposed at Units 2 or
- 3. These wastes should be identified and accounted for as Unit 1 wastes.
A-26 Radioactive wastes to be generated'by the Dresden 1 facility-as a result of preparation for SAFSTOR or that will result during the SAFSTOR period will be disposed of at an off-site l licensed disposal facility. Processing and shipping of ' wastes from all the Dresden units 'is provided by facilities l that are common for Units 1, 2 and 3 (i.e., by Unit 2 and 3 ' facilities). With the exception of dry active wastes, all radioactive wastes ~that originate from Unit 1 are, and will be, accounted for as Unit 1 wastes.
)
l . . - l Q 27 In Section 7.4 nonradioactive wastes are briefly discussed. Criteria that will be used for separating nonradioactive from radioactive wastes should be specified. Also, the administrative programs and equipment to be used to ensure ' these wastes are properly separated should be described. l A 27'Non-radioactive wastes will be segregated from radioactive l wastes and disposed of by unrestricted release to appropri-ate non-radioactive disposal facilities, such as landfills or scrap recovery facilities. Procedure for surveying material to be released off-site is given in Dresden's
" Radiation Protection Procedure", DRP .14 80 - 3 Rev. 2, July 1987. This procedure was prepared in reference to NRC I.E.
Bulletins 81-07 and 85-92. The procedure calls for the surfaces of items in' question to be scanned with a GM survey meter or equivalent type of instrument having a minimum sensitivity of 5,000 dpm per 100 cm2 at a maximum sweep velocity of 5 cm per.second. Additionally, all items are to be surveyed with a gamma scintillation detector. Criteria . for unrestricted release of material will be Regulatory ! Guide 1.86 criteria, in conjunction with the 5 micro R/hr above background at one meter (as described in NRC Question 1.) However, as a conservative measure when deemed practical, an item found with any detectable contamination or elevated gamma levels will be considered as a radioactive waste. Large bulky material, such as concrete rubble or scrap steel or other material not likely to come into contact with the general public, are examples of items not deemed practical. Items which are of a size or construction to make potentially contaminated surfaces inaccessible for measurement will be assumed to be radioactive wastes, as is recommended by NRC Regulatory Guide 1.86. Based on SFP crud sample data (provided with Question 3 response), routine alpha contamination measurements are not deemed necessary. The SFP crud represents the worst case contamination in regard to ratio of alpha contamination to beta contamination. The everage and maximum ratics of alpha activity to beta activity, bared on sample analysis results with response to Question 3, are 1.4-3 and 3.65-3. provided respectively. A shielded thin window GM survey detector has a typical sensitivity of about 1,500 dpm/100cm2 That beta contamina-tion level would correspond to a maximum alpha contamination level of 6 dpm/100 cm2, which is only 6Ss of the Regulatory Guide 1.86 limit for worst case alpha emitters. --r,.-- -- - - - . _ - - - - . - - - - - - _ _ _ _ _ _ - - - . _ _ _ _ - . - __. - _ - - - - _
Dresden Station will provide a formal submittal of the changes to the Technical Specifications upon acceptance of the proposed changes contained in this submittal. I 1 l L i l j j l
., l .' TECHNICAL SPECIFICATIONS Q1 Offsite dosimeters should have their location documented on a' map as part of the TS.
Offsite water sampling s h id be specified (location and radionuclides). The sampling should be in a downstream location with respect to surface water and groundwater. ) J Onsite sampling should be described as to radionuclides sampled and the location. 1 l A1 This information is contained in the Off-Site Dose Calculation Manual (0DCM) which is referenced in the proposed Technical' Specifications for Unit 1. Figure 8.4.1 of the Off-Site Dose Calculation Manual (0DCM), Fixed Air Sampling and TLD Sites and Outer Ring TLD Locations, contains the map specified. Table 8.4.1 of the Off-Site Dose Calculation Manual (ODCM), Environmental Radiation Monitoring Program, provides the offsite water sampling locations and type of analysis. Table 8.4.1 of the Off-Site Dose Calculation Manual (0DCM), Environmental Radiation Monitoring Program, provides the onsite sampling; locations and type of analysis. A c<:py of these tables and figure may be found in the response to Decommissioning Plan Comment #15 (Attached). l \. l 3311a
i. TS.(Section-5.0) should include training on 10CFR Parts 19, 20, 61 and Q 2'- 71 as a minimum. A retraining frequency should be specified. A2 Training on 10CFR Parts 19,20,61. and 71 will . be given to appropriate plant personnel and will be controlled by an administrative procedure.- The f requency of ret' raining will be as specified in' Section 5.1.F of the proposed Technical Specifications (TS) where it states.that: Retraining shall be conducted at intervals not exceeding two years. l 3311a I 1.. . - _ _ _ _ - _ _
} l - l Q3 The procedure list (Section 5.0) should include calibration ) instruments, effluent releases, transportation and emergency plan implementing procedures. A3 This comment will be incorporated into the proposed Unit 1 Technical Specifications. I 3311a
s-i. Q4 In Saction '5.thn
.$. licensoe - has ~ daleted tha requiram:nt to retain records of' radioactive shipments. .We' consider these records should be retained for-'the' currently required period of five years.
A4 Deleting this requirement was an oversight on Dresden's part. The-requirement will be reinserted. 3311a i
.' Q5 In a letter, datsd Dactmbar 10, 1987, the licanssa respond 3d ' to an N
- allegation concerning potsntial degradation of tha fuel racks and concrete structure as a result of stagnant conditions in the spent fuel pool for over two years. This led to the growth of micro-organisni which could compromise the safe storage of the Dresden Unit 1 fue due to microbial influenced corrosion. The Decommissioning Plan and Technical Specifications should address the corrective action taken- to assure safe spent fuel storage.
Specifically, the Technical Specifications . should address source and quality of makeup water and emergency makeup water and operability of spent fuel pool cleanup system including quality of spent fuel pool water. A5 Makeup and emergency water for the Unit i spent fuel pool originates from the Demineralized Water Storage Tanks (T-105A & B). T-105A supplies contaminated demineralized water. T-105B supplies clean demineralized water. The quality of the contaminated demineralized water is normally:
- Turbidity: 10 or less Nephelometric Turbidity Units (NTU) or 1 part per million (ppm) or less filterable solids.
- Conductivity: less than 1.0 micromho per centimeter (umho/cm)
- Gross Activity: 8.0 E-4 microcuries/ milliliter (uCi/ml) or less
- Silica: less than or equal to 100 ppb
- Total Organic Carbons (TOC): less than or equal to 0.40 ppm
- pH - 5.6 to 8.6
- Chloride - less than or equal to 20 ppb.
- Sulfate - less than or equal to 20 ppb.
Note these values are currently provided in Dresden Operating Procedure (DOP) 2000-30, Water Going To Unit 2/3 Condensate Storage. The quality of the clean demineralized water is normally:
- Conductivity: less than or equal to 1.0 micromho per centimeter (umho/cm)
- Gross Activity: less than or equal to 100 picoCuries per Liter (pCi/L)
- Silica less than or equal to 20 ppb
- Total Organic Carbons (TOC): less than or equal to 400 ppb o pH: 6.0 to 8.0
- Chloride - less than or equal to 20 ppb.
- Sulfate - less than or equal to 20 ppb.
Note that these values are currently provided in Dresden Chemistry Procedure (DCP) 1250-T11, Storage Tank Reporting Conditions. l 3311a
- c. .c h *;. ,
A-5 (Cont'd) The . following fuel pool water quality. limits will be included in the Technical Specifications: ,
-Chlorides 1 0.5 ppm l
Conductivity 1.10.0 mmhos/cn @ 25 C ! 5.3 1.pH I 8.6-If these limit.s are exceeded, the parameter (s) will be restored within 30 days or a report shall be made to the NRC describing. the actions that will be taken to restore the water quality to the above limits. The spent fuel pool cleanup system shall be able to maintain this quality. 1 3311a i
b Q6 Defino GSEP in paragraph A of prga B.3/4.2-2. A6 This comment will be incorporated. GSEP stands for Generating Station Emergency Plan. l i l 3311a l
4 1 i I
\
J i ATTACHMENT 2 I i t I l i j l I 1 1 l l
*g.
i Q-1 Page 1-1, Section 1.0. Provide the definition of the Offsite Dose Calculation Manual (00CM) as referenced in T.S. 5.2.A.10. A-1 The following comment concerning the ODCM will be incorporated into the proposed Technical Specifications: The ODCM contains the methodology and parameters used in the calculation of offsite doses due to radioactive gaseuus and liquid effluents, and in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints, i 0057T/6 l 1 l
e Q-2 Page 3/4.0-3, Section 4.0.:'.1 The environmental monitoring requirements reference Units 2 and 3 licence:. The Unit i environmental monitoring requirements should be spelled out or an appropriate reference to the ODCM given. A-2 This comment will be incorporated into the prc90 sed Technical Specifications by referencing the ODCM. 1
. Q3 Pzga 3/4.0-3, Item 4.0.E.b. Tha word "may" should ba ch ngsd to "shall". A3 This comment will be incorporated into the proposed Technical Specifications. 1 3311a J
' Q '.4 Pegn; 3 / 4 .' l-1, Saction 3.1. 'All suxiliary electrical systcm I requirements have batn dsleted excspt for the D.C. Supply System.
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Justify the deletion of requirements for auxiliary electrical systems. A4 In order to justify the deletion of requirements for auxiliary - electrical systems, it is . first necessary to understand ' the reason these requirements were initially placed in the Technical Specifications. The bases for the current Unit 1 Technical Specifications provides the following reasons for requiring auxiliary electrical systems: (1) operate the auxiliaries during plant operation (2) operate facilities to cool and lubricate the plant during shutdown (3)' operate the engineered safeguards following an accident Addressing these points individually, deletion of requirements for auxiliary electrical systems may be justified. (1) Since the plant will never be operated again, there is no need to assure electrical power to the auxiliaries during plant operation. (2) The plant is now shutdown and cool with no fuel in the core. No facilities are required now to keep the plant cool during continued shutdown. Therefore, electrical power is not needed to keep the plant cool. The plant is waiting to be dismantled; so, lubrication of the plant during shutdown is not necessary. Hence, electrical power is not required to lubricate the plant during shutdown. (3) The license now prohibits operating the Unit i reactor. Engineered safeguards are only required to minimize the consequences of an accident that occurs during operation. Therefore, electrical power is not required to operate the j engineered safeguards. To summarize: the auxiliary electrical systems are not required for a unit waiting to be dismantled and may therefore be deleted from the Technical Specifications for Unit 1. Dresden is planning to delete the D.C. Supply System for.the Unit 1 Technical Specifications as well. This deletion may be justified by considering the Technical Specification bases for requiring D-C supply. Section 3.9 of the current Unit 1 Technical Specifications provides the following reasons for requiring the D.C. Supply: i 3311a
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,. A 4 (Cont'd)
The D-C' supply is required for control'and motive power for switchgear and engineered safety features. .. As discussed above, the engineered safety ' features are not required for a plant that is prohibited f rom -' operating. . Control and motive power for the switchgear is necessary, but not safety related.
.Therefore, inclusion of the D-C supply in the Technical Specifications is not required.
However, Dresden does agree -that electric. power is required to maintain the spent fuel pool. The spent fuel pool water quality-will be controlled by a new Technical Specification as described in Comment 5 from Attachment 1. The spent fuel pool water . level is already controlled by Technical Specification. Therefore, having a Technical Specification dedicated to auxiliary electric power is not. deemed to be necessary. 3311a
1 . l l l Q-5 Page 3/4.2-1, Section 3.2. Provide Technical Specification restrictions l with respect to the handling of heavy objects over the spent fuel. Also, j provide a corresponding safety analysis. A-5 The safety analysis to be provide in response to the Reference (a) request will demonstrate that no 10 CFR 100 concerns exist. Therefore, restrictions with respect to the handling of heavy objects over the spent fuel pool will be controlled by procedure rather than by the Technical Specifications. This analysis will be provided April 10, 1989. 0057T
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. . i 3.3.A Q6 Page 3/4.3-1, Section and Section 4.3.A. These TS shotid specify Unit I requirements rather than referring to Units 2 anc 3 TS. The Unit i diesel fire pump should be a required part of the Unit i fire protection system.
A6 Section 3.3 will be deleted from the Technical Specifications and incorporated in the Dresden Administrative Technical Requirements (DATR). This change will keep the Unit 1 Technical Specifications consistent with the Unit 2/3 Technical Specifications provided that approval is received for changing Units 2 and 3's Technical Specifications. If the proposed change to the Unit 2/3 Technical Specifications is not approved, then this comment will be incorporated into the proposed Unit 1 Technical Specifications. It should be noted that this change is being made in response to Generic Letter 88-12, Removal of Fire Protection Requirements from Technical Specifications. I i l l 3311a 1 1 C___________.___. _
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*- Q7 Page-5-1, Szction 5.1.C. This TS rafors to Unit 2 and Unit 3 TS for shift manning. These Unit 2 and 3 TS do.not reflect the permanent shutdown status of Unit 1. Shift manning for Unit 1 should be specified in the Unit 1 TS independently.
I l A7 A change to the Unit 1 Shift manning table requires a change to the Units 2/3 Technical . Specifications to assure consistency among all three units at Dresden. At the present time, Dresden is planning- to incorporate the Dresden 2/3 shif t manning table into the proposed Unit 1 Technical Specifications. This shif t manning table will be changed on all three units following a required review. i l 4 3311a
j -, 6 Q 8: Page 5-1, Section 5.1.C. The requir'ement for a five man fire brigade has been deleted from the Unit 1 TS. Provide TS requirements on fire brigade size and justify the change from the present TS. A8 The requirement for a five man fire brigade will be incorporated into the Dresden Administrative Technical Requirements. This change will keep the Unit 1 Tech'nical Specifications consistent with the Unit 2/3 Technical Specifications provided that approval is received for changing Units 2 and 3's Technical Specifications. If the proposed change to the Unit 2/3 Technical Specifications is not approved, then this comment will be incorporated into the proposed Unit 1 Technical Specifications. It should be noted that this change is being made in response to Generic Letter 88-12, Removal of Fire Protection Requirements from Technical Specifications and that Dresden' has no intention of reducing the five man fire brigade requirement. 3311a
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- l Q9 Paga 5-3, , Figure 5.1-1. Corporata and s ta, tion organization chart 'i should be' updated to be consistent with present station and corporate organizations. The Decommissioning Program Management Chain should be included in the organization chart.
I A9 Corporate and station organization _ charts .will be updated to be consistent with present station and corporate organizations. The corporate, station, and decommissioning management organization chains will be incorporated into the Management ' Plan for Nuclear Operations, Section 3 -Organizational Authority, Activity; Section 6 Interdepartmental Relationships and will be deleted from the Unit 1-proposed Technical Specifications. This change will keep the Unit 1 Technical Specifications consistent with the -Unit 2/3 Technical Specifications provided that approval is received for changing Units 2 and 3's Technical Specifications. If the proposed change to the Unit 2/3 Technical Specifications is not approved, then this comment will be incorporated into the proposed Unit 1 Technical Specifications. i I 3311a a I L_ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ 3
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- s. Q 10 Pags 5-5. Section ' 5.1.G.1.b, Ittms (4) and (7). TS should spacify tha' f requency for audits of quality assurance and for audits - of l of f-site reviews.
A 10 The ' frequency of audits was inadvertently left out of the proposed
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changes. The 2 year frequency will be reinstated into the proposed Technical Specifications. l l i l 3311a e__-_-_- _ _ . __- - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - _ _ -
v, Q 11 Paga 5-9, Section 5.1.G.2.a. A TS should be addad to requira a review of all proposed changes to the Decommissioning Plan. A 11 This comment will be incorporated into the proposed Technical Specifications as Section 5.1.G.2.a.13. 3311a t_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ -. . - - . - _ - . - . - - - - . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ __-._____ ____________ _ __
Q 12 Pega 5-13, Ssetion 5.4.A.1. R3quircments for maintaining rGeords of normal plant operations should be deleted. j 4 A 12 This comment will be incorporated into the proposed Technical Specifications. I i 1 l 1 l l 3311a 1
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