ML17263A956
| ML17263A956 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/31/1994 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17263A955 | List: |
| References | |
| NUDOCS 9503090159 | |
| Download: ML17263A956 (70) | |
Text
SEMIANNUALRADIOACTIVEEFFLUENT RELEASE REPORT R. E. GINNA NUCLEAR PLANT ROCHESTER GAS AND ELECTRIC DOCKET NO. 50-244 JULY - DECEMBER, '1994
'9503090159 95030i PDR" ADOCK 05000244
,R PDR
TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 SUPPLEMENTAL INFORMATION 2.1 REGULATORY LIMITS 2.2 MAXIMUMPERMISSIBLE CONCENTRATIONS 2.3 RELEASE RATE LIMITS 2.4 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY 2.5 BATCH RELEASES 2.6 ABNORMALRELEASES 3.0
SUMMARY
OF GASEOUS RADIOACTIVEEFFLUENTS 4.0
SUMMARY
OF LIQUIDRADIOACTIVEEFFLUENTS 5.0 SOLID WASTE 6.0 LOWER LIMITOF DETECTION NOT MET 7.0 RADIOLOGICALIMPACT 8.0 METEOROLOGICAL DATA 9.0 LAND USE CENSUS CHANGES 10.0 ANNUALTABULATIONOF PERSONNEL EXPOSURE 11.0 LEAKTEST OF SEALED SOURCES 12.0 CHANGES TO THE OFFSITE DOSE CALCULATIONMANUAL 13.0 CHANGES TO THE PROCESS CONTROL PROGRAM 14.0 MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS
LIST OF TABLES Table 1A Gaseous Effluents - Summation of all Releases Table 1B Gaseous Effluents - Continuous and Batch Releases Table 2A Liquid Effluents - Summation of all Releases Table 2B Liquid Effluents - Continuous and Batch Releases Table 3
Solid Waste and Irradiated Fuel Shipments Table 4A Radiation Dose to Nearest Individual Receptor from Gaseous Releases Table 4B Radiation Dose to Nearest Individual Receptor from Liquid Releases Table 5A Number of Personnel and Man-Rem by Work and Job Function Table 5B Standard Report of Personnel Whole Body Exposure
~.o iNTRODUCTiON This Semiannual Radioactive Effluent Release Report is for the Rochester Gas and Electric Corporation R.E.
Ginna plant and is submitted in accordance with the requirements of Technical Specification Section 6.9.1.4.
The report covers the period from July 1, 1994 through December 31, 1994.
This report includes a summary of the quantities of radioactive gaseous and liquid effluents and solid waste released from the plant presented in the format outlined in appendix B of Regulatory Guide 1.21, Revision 1, June 1974.
All gaseous and liquid effluents discharged during this reporting period were in compliance with the limits of the R.E. Ginna Technical Specifications.
2.0 2.1 SUPPLEMENTAL INFORMATtON Re ulato Limits The Technical Specification limits applicable to release of radioactive material in liquid and gaseous effluents are:
2.1.1 Fission and Activation Gases The instantaneous dose rate, as calculated in the ODCM, due to noble gases released in gaseous effluents from the site shall be limited to a release rate which would yield < 500 mrem/yr to the total body and < 3000 mrem/yr to the skin if allowed to continue for a full year.
The air dose, as calculated in the ODCM, due to noble gases released in gaseous effluents from the site shall be limited to the following:
(i)
During any calendar quarter to < 10 mrad for gamma radiation and to < 20 mrad for beta radiation.
Radioiodine Tritium and Particulates The instantaneous dose
- rate, as calculated in the ODCM, due to radioactive materials released in gaseous effluents from the site as radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days shall be limited to a release rate which would yield < 1500 mrem/yr to any organ ifallowed to continue for a full year.
The dose to an individual, as calculated in the ODCM, from radioiodine, radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days released with gaseous effiuents from the site shall be limited to the following:
(i)
During any calendar quarter to < 7.5 mrem to any organ.
(ii)
During any calendar year to < 15 mrem to any organ.
~Uid Elll The release of radioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed the limits specified in accordance with Appendix B, Table II, Column 2 and notes thereto of 10CFR20.
For dissolved or entrained noble gases the total activity due to dissolved or entrained noble gases shall not exceed 2 E-4 uCi/ml.
The dose or dose commitment to an individual as calculated in the ODCM from radioactive materials in liquid effluents released to unrestricted areas shall be limited:
(i)
During any calendar quarter to < 1.5 mrem to the total body and to
< 5 mrem to any organ, and (ii)
During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.
Maximum Permissible Concentrations MPC 2.2.1 For gaseous effluents, maximum permissible concentrations are not directly used in release rate calculations since the applicable limits are stated in terms of dose rate at the unrestricted area boundary.
2.2.2 For liquid effluents, the maximum permissible concentration values specified in 10CFR20, Appendix B, Table II, column 2 are used to calculate release rates and permissible concentrations at the unrestricted area boundary.
A value of 2E-04 uCi/ml is used as the MPC for dissolved and entrained noble gases in liquid effluents.
2.3 Release Rate Limits The release rate limits for fission and activation gases from the R.E. Ginna plant are not based on the average energy of the radionuclide mixture in gaseous effluents; therefore, this value is not applicable.
However, the average energy of the radionuclide mixture was 0.386 Mev.
2.4 Measurements and A roximations of Total Radioactivit Gamma spectroscopy was the primary analysis method used to determine the radionuclide composition and concentration of gaseous and liquid effluents.
Composite samples were analyzed for Sr-89, Sr-90 and Fe-55 by a contract laboratory.
Tritium and alpha analysis were done using liquid scintillation and gas flow proportional counting respectively.
The total radioactivity in effluent releases was determined from the measured concentration of each radionuclide present and the total volume of effluents released.
2.5 Batch Releases 2.5.1 ticiuid 1.
2.
Number of batch releases:
Total time period for batch releases:
Maximum time period for a batch release:
Average time period for batch releases:
Minimum time period for a batch release:
Average stream flow (LPM) during periods of release effluent into a flowing stream:
1.56 E+02 6.67 E+04 min.
1.84 E+04 min.
4.28 E+02 min.
4.6 E+01 min.
1.29 E+06 LPM (3)
I 2.5.2 Gaseous 1.
2.
Number of batch releases:
Total time period for batch releases:
Maximum time period for a batch release:
Average time period for batch releases:
Minimum time period for a batch release:
1.1 E+01 2.24 E+04 min.
1.79 E+04 min.
2.04 E+02 min.
2.36 E+02 min.
2.6 Abnormal Releases There were no abnormal releases of liquid or gaseous effluents during the reporting period.
3.0
SUMMARY
OF GASEOUS RADIOACTIVEEFFLUENTS The quantities of radioactive material released in gaseous effluents are summarized in tables 1A and 1B.
All releases were considered to be elevated releases.
4.0
SUMMARY
OF LIQUID RADIOACTIVEEFFLUENTS The quantities of radioactive material released in liquid effluents are summarized in tables 2A and 2B.
5.0 SOLID WASTES The quantities of radioactive material released in shipments of solid waste transported from the site during the reporting period are summarized in Table 3. Principal nuclides were determined by gamma spectroscopy and non-gamma emitters were calculated from scaling factors determined by an independent laboratory from representative samples of that waste type.
6.0 LOWER LIMITOF DETECTION NOT MET There were no gamma emitting radionuclides that did not meet the required lower limit of detection for liquid releases.
(4)
7.0 RADIOLOGICALIMPACT An assessment of doses to the maximally exposed individual from gaseous and liquid effluents was performed for locations representing the maximum dose.
In all cases, doses were well below Technical Specification limits.
Doses were assessed upon actual meteorological conditions considering the noble gas exposure, inhalation, ground plane and ingestion pathways.
The ingestion pathways considered were the produce, vegetable, goat's milk, cow's milk and meat pathway.
The results of this assessment are presented in Tables 5A and 5B.
8.0 METEOROLOGICAL DATA The annual summary of hourly meteorological data collected during 1994 is not included with this report, but can be made available at the R.E.
Ginna Plant as allowed by Technical Specifications.
9.0 LAND USE CENSUS CHANGES There were no changes in critical receptor location for dose calculations during the reporting period.
10.0 ANNUALTABULATIONOF PERSONNEL EXPOSURE The annual tabulation of the number of station, utility and other personnel receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job function required by Technical Specification 6.9.2.2 is included as Tables 5A and 5B.
11.0 LEAKTEST OF SEALED SOURCES No sealed sources were found to be leaking when smeared by both wet and dry smears.
12.0 CHANGES TO THE OFFSITE DOSE CALCULATIONMANUAL (ODCM)
There were no changes to the ODCM during the report period.
CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)
The PCP was completely rewritten during this report period and a copy is attached as defined in Technical Specification 6.16.
MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS There were no major changes to the Radwaste Treatment Systems during the reporting period.
%e
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ROCHESTER GAS ELECTRIC CORPORATION Table 1A EFFLUENT AND WASTE DISPOSAL SEMIANNUALREPORT GASEOUS EFFLUENTS - SUMMATIONOF ALLRELEASES JULY - DECEMBER 1994 A. Fission &activation ases Unit Quarter 3rd Quarter 4th Est. Total Error,%
- 1. Total release
- 2. Avera e release rate for eriod
- 3. Percent of technical s ecificatlon limit B. Iodines CI uCi sec 9.06E+00 1.24E+01 1.14E+00 1.56E+00 1.81E44 2ASE44 6.60E+00
- 1. Total iodine-131
- 2. Avera e release rate for eriod
- 3. Percent of technical s ecificatlon limit Cl 2.95E45 uCi sec 3.71E46 8.14E43 1.18E45 1.49E46 3.27E43 1.80E+01 C. Particulates
- 1. Particulates with half-lives > Sda s
- 2. Avera e release rate for eriod
- 3. Percent of technical s ecification limit
- 4. Gross al haradioactiv'I uCI sec CI 1.38E46 8.84E47 1.74E47 1.11E47 1.31E45 8.36E46 4.00E+01 D. Tritium
- 1. Total release
- 2. Avera e release rate for erlod
- 3. Percent of technical s ecificatlon limit CI 1.71E+01 1.04E+01 uCI sec 2.16E+00 1.31E+00 2.54E44 1.54E44 3.20E+00 E. Carbon-14
- 1. Total release
- 2. Avera e release rate for erlod
- 3. Percent of technical s ecification limit Ci uCi sec 7.01E41 8.82E42 4.69E46 6.92E41 8.71 E42 4.63E46 3.00E+01 Note:
Isotope for which no value is given were not identmed in applicable releases.
ROCHESTER GAS ELECTRIC CORPORATION Table 1B EFFLUENT AND WASTE DISPOSAL SEMIANNUALREPORT GASEOUS EFFLUENTS - ELEVATEDRELEASE Nuclldes released 1.
Fission ases ar on41 k
ton%5 Unit Cl 1.50E41 1.52E41 CI 3.21E42 4.47E42 2.75E43 Continuous Mode Batch Mode Quarter Quarter Quarter Quarter 3rd 4th 3rd 4th k
tonMm CI 3.89E42 5.51E42 k
ton%7 k
ton~
xenon-131m xenon-133 xenon-133m xenon-135 xenon-135m xenon-138 others s ec Total for erlod Cl 6.71E42 9.52E42 Cl 7.98E42 1.11E41 Ci 7.43E43 Cl 4.54E+00 7.05E+00 Cl 6.54E43 1.04E42 Cl 2.70E+00 2.95E+00 Ci 8.47E41 1.09E+00 Cl 3.19E41 4.14E41 CI CI CI Cl Cl 8.76E+00 1.19E+01 1.28E43 4.58E44 2.65E41 4.58E41 7.35E44 2.53E43 8.61E43 2.79E43 3.08E41 5.11E41 2.
Iodlnes iodine-131 Ci 1.63E45 1.18E45 1.31E45 Iodine-133 iodine-135 Total for erlod Cl 2.70E45 CI CI 4.34E45 1.34E45 5.60E45 2.52E45 6.91E45 3.
Particulates strontium%9 strontium-90 cesium-134 cesium-137 Nb-95 cobalt-58 cobalt%0 Total for erlod unidentified CI CI CI CI CI'I CI CI CI 1.38E46 1.38E46 8.84E47 8.84E47 Note:
Isotope for which no value is given were not identified in applicable releases.
ROCHESTER GAS ELECTRIC CORPORATION Table 2A EFFLUENT AND WASTE DISPOSAL SEMIANNUALREPORT LIQUIDEFFLUENTS - SUMMATIONOF ALLr ELEASES JULY-DECEMBER 1994 A. Fission and activation roducts
- 1. Total release (not including tritium, ases,al ha
- 2. Average diluted concentration durin eriod
- 3. Percent of a licable limit Unit Quarter 3rd Ci 2.06E42 uCI ml 1.26E-10 9.64E43 Quarter 4th 6.39E43 3.79E-11 4.29E43 Est. Total Error, 9o 7.00E+00 B. Tritium
- 1. Total release
- 2. Average diluted concentration durin eriod
- 3. Percent of a licable limit Ci uCi ml 1.86E+01 1.14E47 3.79E43 1.88E+01 3.20E+00 1.11E47 3.71E43 C. Dissolved and entrained ases
- 1. Total release CI 2.15E43 9.56E45 4.00E+ 01
- 2. Average diluted concentration durin eriod
- 3. Percent of a Iicable limit uCi ml 1.32E-11 5.68E-13 6.58E46 2.84E47 D. Gross al ha radioactivit
- 1. Total release CI N A E. Vol. of waste released riorto dilution Liters 3.49E+07 2.12E+07 5.00E+00 F. Vol. of dilution water used durin eriod Liters 1.63E+11 1.68E+11 5.00E+00 Note:
Isotope for which no value is given were not identified in applicable releases.
ROCHESTER GAS ELECTRIC CORPORATION Table 2B EFFLUENT AND WASTE DISPOSAL SEMIANNUALREPORT LIQUIDEFFLUENTS Continuous Mode Batch Mode Nuclldes Released Unit Quarter Quarter Quarter Quarter chromium-51 man anese-54 Cl CI 3rd 4th 3rd 4th 2.24E45 iron-55 iron-59 cobalt-58 cobalt%0 zinc%5 strontium%9 strontium-90 zirconium niobium-95 mol bdenum-99 silver-110m antimon -122 antimon -124 antlmon -125 iodine-131 iodine-133 Ci 3.55E46 Ci Ci Ci Ci Ci Cl CI Ci 5.50E45 CI Ci Ci Ci Cl 4.62E45 Cl 8.87E45 6.43E45 1.07E44 2.31E45 2.29E44 2.81E44 5.42E46 3.22E45
'2.55E45 6.31E44 2.23E43 4.74E45 2.44E43 1.03E44 3.82E43 8.78E46 2.00E46 3.08E45 9.69E45 5.20E44 1.87E43 4.11E44 Iodine-135 cesium-134 cesium-136 CI Ci Ci 7.64E45, 7.1 5E45 5.53E43 2.36E43 2.55E44 1.14E43 cesium-137 bariutn lanthanum-140 cerium-141 ruthenium-106 ruthenium-103 CI 4.97E45 CI CI CI CI 2.09E45 2.53E43 1.79E43 Total for eriod above unidentified Cl 3.20E44 Ci 2.43E44 2.03E42 6.15E43 xenon-133 xenon-135 CI Ci 1.18E43 2.43E45 9.69E44 7.13E45 Note:
Isotope for which no value Is given were not identified In applicable releases.
I
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Table 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUALREPORT Solid Waste and Irradiated Fuel Shipments July - December 1994 A.
SOLID WASTE SHIPPED OFFSITE FOR BURIALOR DISPOSAL (NOT irradiated fuel)
TYPE OF WASTE Spent resins, filtersludges, evaporator bottoms, etc.
UNIT m
Ci 6 MONTH PERIOD 2.04 E+00.
9.06 E+05 EST. TOTAL ERROR 7.0 E+00 7.0 E+00 Dry compressible waste, contaminated equipment, etc.
m Ci 3.54 E+01 7.0 E+00 7.34 E-02 7.0 E+00 Irradiated, components, control rods, etc.
d.
Other (describe) m Ci m
Ci N/A N/A N/A N/A N/A N/A N/A N/A 2.
Estimate of major nuclide composition (by type of waste)
FG-55 Co-60 Ni-63 Cs-137 C-14 H-3 Cs-134 3.8 E+01 1.6 E+01 1.2 E+01 1.9 E+01 5 E+00 7 5+00 5 E+00 FG-55 Co-60 Nb-95 Cs-1 37 Zr-95 Co-58 Cr-51 C0-144 Ni-63 Cs-134 Mn-54 3.8 E+01 1.0 E+01 1.0 E+01 9.0 E+00 7 E+00 7 E+00 6 E+00 5 E+00 4 E+00 2 E+00 1 E+00 3.
Solid Waste Disposition NUMBER OF SHIPMENTS MODE OF TRANSPORTATION DESTINATION Sole use truck B. IRRADIATEDFUEL SHIPMENTS (Disposition)
Oak Ridge, TN N/A N/A None NUMBER OF SHIPMENTS MODE OF TRANSPORTATION DESTINATION
Table 4A Radiation Doses to Nearest Individual Receptor From Gaseous Releases in Rem First Quarter 1994 ADULT T.BODY THYRD SKIN T.BODY THYRD SKIN CHILD T.BODY THYRD SKIN T.BODY N
NNE NE ENE E
WNW NW NNW 3.00E-OS 3.20E-OS 3.50E48 3.70E-OS 5.90E-08 6.10E-08 6.50E48 6.60E-08 2.20E-06 2.30E-06 5.40E-06 5.40E46 5.20E-06 5.30E46 1.60E-06 1.60E-06 1.80E-06 1.80E-06 2.80E-06 2.80E-06 3.60E-06 3.60E46 2.10E-06 2.10E-06 1.80E-06 1.80E-06 2.30E-07 2.40E47 3.80E-08 3.90E-OS 3.90E-08 4.00E-08 2.70E-OS 3.00E-08 3.20E-OS 3.50E-OS 5.60E-08 5.90E-08 6.30E-08 6.60E-OS 2.50E-06 2.20E-06 6.10E-06 5.40E46 6.00E-06 5.20E-06 1.80E-06 1.60E-06 2.00E-06 1.80E46 3.30E-06 2.80E-06 4.20E-.06 3.60E46 2.40E-06 2.10E46 2.10E-06 1.80E-06 2.80E-07 2.30E-07 4.00E-OS 3.80E-08 3.90E-OS 3.90E-08 3.20E-OS 2.70E-OS 3.90E-OS 3.20E-OS 6.20E-OS 5.60E-08 6.80E-OS 6.30E-OS 2.30E-06 2.50E-06 5.40E-06 6.10E-06 5.30E-06 6.00E-06 1.60E-06 1.80E-06 1.80E46 2.00E-06 2.80E-06 3.30E-06 3.60E-06 4.20E-06 2.10E-06 2.40E-06 1.80E-06 2.10E-06 2.40E-07 2.80E-07 3.90E-OS 4.00E-OS 4.10E-OS 3.90E-OS 2.80E-08 3.10E-OS 3.50E-08 3.90E-08 5.70E-08 6.20E-08 6.60E-OS 6.90E-OS 2.20E46 2.30E-06 5.40E-06 5.50E-06 5.20E-06 5.30E-06 1.50E-06 1.60E-06 1.80E-06 1.80E-06 2.80E-06 2.80E-06 3.60E46 3.60E-06 2.10E46 2.10E-06 1.80E-06 1.80E-06 2.30E-07 2.40E-07 3.70E48 3.90E-OS 3.80E-08 4.00E-OS 2.70E-08 2.4 0E-OS 3.20E-08 3.00E-08 5.60E-08 5.20E-OS 6.30E-08 6.00E-08 2.50E-06 2.20E-06 6.10E-06 5.30E-06 6.00E-06 5.20E-06 1.80E-06 1.50E46 2.00E-06 1.70E-06 3.30E-06 2.80E-06 4.20' 3.60E-06 2.40E-06 2.10E-06 2.10E-06 1.80E46 2.80E-07 2.30E-07 4.00E-OS 3.50E-OS 3.90E-08 3.40E-OS 2.70E-08 2.70E-08 3.40E-OS 3.20E-OS 5.70E-OS 5.60E-OS 6.30E-OS 6.30E-OS 2.20E-06 2.50E-06 5.40E-06 6.10E-06 5.20E-06 6.00E-06 1.50E-06 1.80E-06 1.80E-06 2.00E-06 2.80E-06 3.30E-06 3.60E-06 4.20E-06 2.10E46 2.40E-06 1.80E46 2.10E-06 2.40E-07 2.80E47 3.70E48 4.00E-OS 3.70E-OS 3.90E-OS 5.40E-06 5.40E-06 6.10E-06 5.40E-06 5.40E-06 6.10E-06 5.40E-06 5.50E-06 6.10E-06 5.30E46 5.40E46 6.10E-06
4A
Table 4A Radiation Doses to Nearest Individual Receptor From Gaseous Releases in Rem Second Quarter 1994 T.BODY ADULT THYRD SKIN T.BODY.
THYRD SKIN T.BODY CHILD THYRD SKIN T.BODY THYRD N
NE ESE SE SSE SSW SW WSW 3.70E-07 6.00E-07 1.50E47 1.80E-07 1.20E-05 1.80E-05 1.30E46 7.00E-06 2.30E-06 1.70E-06 6.20E-06 2.60E45 1.40E-05 2.00E47 1.10E47 2.50E-07 3.70E-07 4.10E-07 6.00E-07 6.70E-07 1.50E47 1.70E-07 1.80E-07 2.00E47 1.00E-05 1.10E-05 1.40E-05 1.60E-05 1.10E-06 1.30E-06 5.50E-06 5.40E-06 2.10E-06 2.40E-06 1.60E-06 1.80E-06 5.40E-06 5.40E-06 2.20E45 2.30E-05 1.40E-05 1.60E45 1.50E47 1.50E-07 1.10E47 1.10E47 2.50E-07 2.60E-07 3.70E-07 3.70E-07 6.00E-07 6.10E-07 1.50E-07 1.50E-07 1.80E-07 1.80E-07 1.20E-05 1.10E45 1.70E-05 1.50E-05 1.30E-06 1.20E46 6.80E-06 5.50E46 2.30E-06 2.10E46 1.70E-06 1.60E-06 6.10E-06 5.50E-06 2.50E-05 2.20E-05 1.40E-05 1.40E-05 2.00E-07 1.50E-07 1.10E-07 1.10E47 2.50E-07 2.50E47 4.10E-07 3.70E-07 6.70E-07 6.00E-07 1.70E-07 1.50E-07 2.00E-07 1.70E47 1.20E-05 1.40E5 1.60E-05 1.70E45 1.30E-06 1.30E-06 5.40E-06 6.40E-06 2.40E-06 2.30E-06 1.80E-06 1.70E-06 5.40E-06 6.00E-06 2.40E-05 2.40E-05 1.60E-05 1.40E-05 1.60E-07 2.10E47 1.10E-07 1.10E7 2.60E47 2.40E47 3.70E-07 4.10E47 6.10E-07 6.70E-07 1.50E-07 1.70E-07 1.80E47 2.00E-07 1.30E45 1.30E-05 1.50E-05 1.60E-05 1.30E-06 1.40E-06 5.90E-06 5.40E-06 2.20E-06 2.50E46 1.60E-06 1.80E-06 5.80E-06 5.40E-06 2.30E-05 2.40E-05 1.40E-05 1.60E-05 1.80E-07 1.80E-07 1.10E-07 1.10E-07 2.50E-07 2.60E-07 3.60E-07 3.70E-07 5.90E-07 6.00E-07 1.50E-07 1.50E-07 1.70E-07 1.80E-07 9.80E-06 1.00E-05 1.40E-05 1.50E-05 1.20E-06 1.20E-06 5.00E-06 5.40E-06 2.10E-06 2.10E-06 1.60E-06 1.60E-06 4.90E-06 5.30E-06 2.10E-05 2.30E-05 1.40E-05 1.40E-05 1.30E-07 1.30E-07 1.00E-07 1.10E-07 2.30E-07 2.40E-07 4.10E-07 6.70E-07 1.70E-07 2.00E-07 1.10E-05 1.60E-05 1.30E-06 5.40E-06 2.40E46 1.80E-06 5.40E-06 2.40E45 1.60E-05 1.50E-07 1.10E-07 2.60E-07 2.60E-05 2.20E-05 2.30E-05 2.50E-05 2.20E45 2.40E-05 2.40E45 2.30E-05 2.40E-05 2.10E-05 2.30E-05 2.40E45
tI
Table 4A Radiation Doses to Nearest Individual Receptor From Gaseous Releases in Rem Third Quarter 1994 N
T.BODY 4.40E-08 6.30E-OS 3.90E-OS 2.30E-OS 1.60E-06 3.60E-06 1.50E-06 6.60E-07 2.60E-06 1.00E-06 8.70E-07 9.00E-07 2.40E-07 2.10E-07 4.80E-09 1.30E-08 ADULT THYRD 4.40E-OS 6.40E48 4.00E-OS 2.30E-08 1.30E-06 3.20' 1.20E-06 5.40E-07 2.40E-06 8.90E-07 7.20E47 7.20E47 2.20E-07 1.60E-07 4.90E-09 1.30E-OS SKIN T.BODY 2.40E-08 4.70E-OS 3.90E-OS 6.50E-OS 2.80E-OS 4.00E48 1.70E-OS 2.30E48 1.10E-06 1.60E-06 3.10E-06 4.50E46 1.20E-06 1.60E-06 4.30E-07 6.70E-07 1.90E-06 3.40E-06 6.80E-07 1.20E-06 5.50E47 9.20E-07 5.30E47 9.40E47 1.60E-07 2.70E47 1.20E-07 2.20E47 2.60E-09 4.90E-09 7.00E-09 1.40E-OS 4.80E-OS 2.40E-OS 6.60E-OS 3.90E-OS 4.10E-08 2.80E-OS 2.40E-08 1.70E-OS 1.30E-06 1.10E-06 4.10E-06 4.00E-06 1.40E46 1.40E-06 5.80E-07 4.60E-07 3.10E-06 2.60E-06 1.10E-06 8.90E-07 8.10E-07 6.00E-07 8.10E47 5.70E47 2.60E-07 1.90E-07 1.80E-07 1.40E-07 4.90E-09 2.60E-09 1.40E-08 7.00E-09 CHILD T.BODY THYRD 5.00E-OS 5.10E-OS 6.30E-OS 6.40E-08 4.00E-08 4.10E-08 2.30E-OS 2.40E-OS 1.70E-06 1.60E-06 7.90E-06 7.80E-06 2.20E-06 2.20E-06 7.50E-07 7.30E-07 6.10E-06 6.00E-06 2.10E-06 2.00E-06 1.10E-06 1.10E-06 1.10E-06 1.10E-06 4.00E-07 3.90E-07 2.80E-07 2.50E-07 4.50E-09 4.60E-09 1.60E-OS 1.60E48 SKIN T.BODY 2.40E-08 3.90E-OS 3.90E-OS 4.90E-OS 2.80E-OS 3.20E-OS 1.70E-OS 1.90E-OS 1.30E-06 1.10E-06 7.50E-06 5.30E-06 2.00E-06 1.40E-06 5.30E-07 4.60E-07 5.30E-06 2.40E-06 1.70E-06 2.00E-06 7.80E-07 6.60E-07 7.20E-07 7.40E-07 2.80E-07 2.30E47 1.80E-07 8.20E-08 2.60E-09 3.10E-09 7.00E-09 1.30E-08 INFANT THYRD SKIN 4.00E-08 2.40E-08 5.00E-08 3.90E-08 3.40E-OS 2.80E-OS 2.00E-08 1.70E-OS 1.20E-06 1.10E-06 5.50E-06 5.40E-06 1.50E-06 1.40E-06 5.50E-07 4.40EW7 2.60E-06 2.40E-06 2.10E46 1.90E-06 8.00E-07 5.80E-07 9.30E-07 5.90E-07 2.30E47 2.00E47 8.20E-OS 1.00E7 3.20E-09 2.60E49 1.30E48 7.00E-09 3.60E-06 3.20E-06 3.10E-06 4.50E46 4.10E-06 4.00E46 7.90E-06 7.80E-06 7.50E-06 5.30E-06 5.50E-06 5.40E-06
Table 4A Radiation Doses to Nearest Individual Receptor From Gaseous Releases in Rem Fourth Quarter 1994 T.BODY ADULT THYRD SKIN TEEN T.BODY THYRD SKIN T.BODY CHILD THYRD SKIN INFANT T.BODY THYRD N
ESE SE SSE SSW SW WSW 4.90E-OS 5.10E-08 3.80E-OS 4.80E-08 2.80E-06 2.20E46 3.20E-06 8.70E47 1.60E46 1.10E-06 1.10E-06 4.70E-07 7.90E47 6.40E-07 1.40E-08 3.20E-OS 5.10E-OS 5.20E-OS 5.30E-OS 5.10E48 3.90E-OS 3.60E48 4.90E-OS 4.60E-OS 2.80E-06 3.20E-06 2.20E-06 2.40E-06 3.20E46 3.70E-06 8.90E-07 1.00E-06 1.70E-06 1.80E46 1.10E-06 1.30E-06 1.10E-06 1.10E-06 4.80E-07 4.80E-07 8.10E-07 8.70E47 6.60E-07 7.80E-07 1.50E-OS 1.50E-OS 3.20E-OS 3.40E-OS 4.90E-OS 5.10E-OS 5.20E-OS 5.50E-OS 3.90E-OS 4.10E-OS 4.90E-OS 5.10E-08 2.80E-06 2.90E-06 2.20E-06 2.20E-06 3.20E-06 3.20E-06 8.80E-07 9.10E-07 1.70E-06 1.70E-06 1.10E-06 1.20E46 1.40E-06 1.40E-06 5.60E-07 5.70E-07 9.30E-07 9.60E-07 6.40E-07 6.70E-07 1.50E48 1.50E-08 3.20E-OS 3.30E-OS 5.20E-OS 4.80E48 5.10E-OS 5.20E-OS 3.60E-08 4.00E-08 4.60E-08 5.00E-OS 3.20E-06 2.80E-06 2.40E-06 2.20E-06 3.70E-06 3.20E-06 1.00E-06 9.10E-07 1.90E-06 1.90E-06 1.30E-06 1.20E46 1.30E-06 2.50E-06 5.70E-07 9.10E47 1.00E-06 1.50E46 7.80E-07 6.60E-07 1.50E-OS 1.40E48 3.40E-OS 3.20E-OS 5.00E-OS 5.20E-OS 5.50E48 5.10E-OS 4.10E-08 3.60E-OS 5.10E-08 4.60E-08 2.90E-06 3.30E-06 2.30E-06 2.40E-06 3.30E-06 3.70E-06 9.50E47 1.00E-06 1.90E-06 2.00E-06 1.20E46 1.30E-06 2.50E-06 2.40E-06 9.30E-07 9.00E-07 1.50E-06 1.50E-06 6.90E47 S.OOE-07 1.50E48 1.50E-OS 3.30E-OS 3.40E-OS 4.30E-OS 4.60E-08 4.50E-OS 4.80E48 3.50E-OS 3.70E48 4.60E-OS 4.80E48 2.80E-06 2.80E-06 2.10E-06 2.20E-06 3.10E-06 3.20E-06 8.50E-07 8.80E-07 1.60E46 1.70E46 1.10E46 1.10E-06 3.10E-06 3.20E46 6.60E-07 6.80E47 9.40E-07 9.60E-07 6.20E-07 6.60E-07 1.30E48 1.40E-OS 2.90E-OS 3.00E48 5.20E-08 5.10E48 3.60E48 4.60E-08 3.20E-06 2.40E-06 3.70E-06 1.00E-06 1.90E-06 1.20E-06 3.20E-06 6.90E47 1.00E-06 7.70E-07 1.50E48 3.40E-OS 3.20E-06 3.20E-06 3.70E-06 3.20E-06 3.20E-06 3.70E-06 3.20E-06 3.30E-06 3.70E-06 3.10E46 3.20E-06 3.70E-06
Table 4B Radiation Dose To Nearest Individual From Liquid Releases In REM 1994 First Quarter T.BODY BONE THYRD 5.32E-06 5.44E-06 5.34E-06 TEEN 5.35E-06 5.58E-06 5.38E-06 CHILD 5.36E-06 5.76E-06 5.40E-06 INFANT 5.29E-06 5.59E-06 5.33E-06 Second Quarter T.BODY BONE THYRD 2.54E-05 2.55E-05 2.20E-05 2.48E-05 2.86E-05 2.26E-05 2.49E-05 4.01E-05 2.45E-05 2.08E-05 2.32E-05 2.17E-05
= Third Quarter T.BODY BONE THYRD 3.31E-06 8.61E-06 3.01E-06 4.03E-06 1.27E-05 3.80E-06 6.96E-06 2.83E-05 6.86E-06 3.29E-06 1.08E-05 3.41E-06 Fourth Quarter T.BODY BONE THYRD 3.54E-06 5.25E:06 3.58E-06 3.76E-06 6.28E-06 3.81E-06 4.64E-06 1.06E-05 4.70E-06 4.40E-06 9.24E-06 4.48E-06 REM94.XLS
Table 5A ROCHES'fER GAS 8 ELECTRIC CORPORATION GINNA STATION HISSER OF PERSOHHEL AHD MAN-REH BY lORK AND JOB FUMCTIOH FOR 94/01/01. 94/12/31 ACTUAL MMOLE BODY DOSE NO.
OF PERSONNEL (>
0.100)
TO'fAL HAH~ REM lSRK PERMIT CATERGORY lQRK GROUP CONTRACT lCRKERS STAT ION EMPLOYEES UTILITY EMPLOYEES COHTRACT QNKERS STATION EHPLOYEES UTILITY EHPLOYEES REACTOR OPERATIONS B SURV HAINTENAMCE PERSONNEL OPERATING PERSOHHEL HEALTH PHY.
PERSONNEL SUPERVISORY PERSONNEL EHGIHEERIHG PERSOHHEL 57 0
42 12 2
43 27 16 11 3
35 0
0 14 3
0.123 0.000 4.968 0.205 0.056 0.624 4.232 1.781 0.058 0.048 0.217 0.000 0.000 0.848 0.000 ROUTINE MAINTENANCE NAIHI'EHAMCE PERSONNEL OPERATING PERSONNEL HEALTH PHY.
PERSOHHEL SUPERVISORY PERSONNEL ENGIHEERING PERSOHNEL 105 0
39 12 2
43 25 16 10 3
53 0
0 15 3
8.336 0.000 5.049 0.475 0.022 7.014 0.876 2.861 2.433 0.097 3.995 0.000 0.000 0.870 0.416 INSERVICE INSPECTIOH MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHY.
PERSONHEL SUPERVISORY PERSONNEL ENGINEERING PERSOHHEL 21 0
9, 10 1
14 2
7 7
2 32 0
0 10 2
0.507 0.000 0.606 0.970
- 0. 148 0.455 0.236 0.051 0.275 0.021 1.768 0.000 0.000 1.814 0.056 SPECIAL HAIN'fENANCE HAIHTENAHCE PERSONHEL OPERATING PERSONNEL HEALTH PHY.
PERSOHHEL SUPERVISORY PERSONNEL EHGIHEERIHG PERSONNEL 97 0
12 11 1
41 13 11 5
~
3 53 0
0 12 2
13.865 0.000 0.424 0.559 0.052 5.702 0.441 1.209 0.533 0.247 5.577 0.000 0.000 0.391 0.120 MASTE PROCESSIHG MAINTENANCE PERSONNEL OPERATING PERSOHHEL HEALTH PHD PERSOHHEL SUPERVISORY PERSOHHEL
,EHGIHEERIHG PERSONHEL 0.587 0.000 0.151 0.000 0.000 0.017 0 ~ 000 0.046 0.000 0.000 0.000 0.000 0.000 0.000 0.000 1EFUELING HAIHTEHANCE PERSONNEL OPERATING PERSOHHEL HEALTH PHY.
PERSOHMEL SUPERVISORY PERSOHHEL ENDINEER IHG PERSONNEL 32 0
7 2
0 22 8
1 0
1 19 0
0 5
1 8.157 0.000 0.361 0.123 0.000 1.595 1.405 0.111 0.000 0.097 1.505 0.000 0.000 0.487 0.000 MODIFICATIONS MAINTENANCE PERSOHHEL OPERATING PERSOHNEL HEALTH PHY.
PERSOHNEL SUPERVISORY PERSONNEL EHGINEERING PERSONNEL 0.060 0.000 0.026 0.047 0.000 0.226 0.020 0.004 0.018 0.020 0.102 0.000 0.000 0.004 0.017 TO'IAL, HAINTEHANCE PERSONNEL 117 OPERATING PERSONNEL 0
HEALTN PHY.
PERSONNEL 42 SUPERVISORY PERSONHEL 12 ENGINEERING PERSONNEL 2
43 27 16 11 3
53 0
0 15 3
31,636 0.000 11.585 2.379 0.278 15.631 7.208 6.063 3.318 0.529 13.164 0.000 0.000 4.414 0.609 GRAND TOTAL 95 70 45.878 32.749
- 18. 187
Table 5B St'ANDARD REPORT OF PERSONNEL WHOLE BODY EXPOSURE 1994 DOSE REM NUMBER OF PEOPLE 00.000 - 00.000 00.001
- 00.100 00.101
- 00.250 00.251
- 00.500 00.501
- 00.750 00.751
- 01.000 01.001 - 02.000 02.001
- 03.000 03.001
- 04.000 04.001
- 05.000 774 368 182 112 52 14 18 0
0 0
Total number of personnel monitored 1520 The total collective dose for 1994 is 148.0 person-rem based on the sum of all ersonnel TLD badge readings.
FIVE HIGHEST EXPOSURES FOR THE YEAR A
B C
D E
1.444 rem 1.322 rem 1.249 rem 1.224 rem 1.209 rem This report contains all personnel monitored during 1994.
PROCESS CONTROL PROQRM5 FOR R.E.
QTNNA NUCLEAR POWER PLANT Rochester Gas f EZectric Corporation Revi si on 7
I PORC:
<<C
IC iC C
TABLE OF CONTENTS I
PURPOSE Page 1
II SCOPE Page 1
III REFERENCES Page 2
IV RESPONSIBILITIES Page 2
V DEFINITIONS Page 3
VI PREREQUISITES Page 4
VII REQUIREMENTS FOR PROCESSING VENDORS Page 5
VIII WASTE MANAGEMENT PRACTICES Page 6
1.0 Waste Stream Sampling.............
Page 6
2.0 Waste Classification Page 8
3.0 Waste Packaging
~
~
Page 8
4.0 Chemical Control...............
Page 9
5.0 Volume Reduction...............
Page 9
6.0 Solidification, Stabilization, and Encapsulation Page 9
7.0 Dewatering Page 11 8.0 Radioactive Material Shipments Page 11 9.0 Quality Control for Sampling and Classification Page 11 IX INTERIM STORAGE:
Page 12 X
ATTACHMENTS Page 13
Process ControZ Program Page 1
&G3IOACTIVE WASTE PROCESS CONTROL PROGM&l PURPOSE:
To provide a description of the radioactive waste Process Control Program(PCP) at the Ginna Nuclear Power Plant (Plant).
Methods are described for processing, classification and packaging of low-level radioactive waste into a form acceptable for disposal in accordance with 10 CFR Part 61 and current disposal site criteria.
SCOPE:
Sampling, classification, processing and packaging of radioactive waste.
1.0 Waste Stream Identification Nine different waste streams have been identified for classification purposes.
Examples of waste streams are:
Dry Active Waste (DAW) paper wood plastic metal cloth rubber glass so3.1 oil and grease NOTE Contaminated oil is classified the same as DAW.
1.2 1.3 1.4 Primary Bead Resin Liquid Waste Processing(LWP)
Bead Resin LWP Activated Charcoal 1.5 LWP Filters LWP cartridge filters LWP bag filters Waste Hold Up Tank (WHUT) cartridge filters
Process Control Program Page 2
1.6 Primary Cartridge Filters Reactor Coolant (Rx) filters Boric Acid Storage Tank (BAST) filters Seal Water Return filters Seal Water Injection filters Reactor Cavity filters Ion Exchange filters Concentrates Holding Tank (CHT) filters Seal Table filters 1.7 Spent Fuel Cartridge Filters Spent Fuel Pool (SFP)
Skimmer filters Spent Fuel Pool filters Tri-Nuclear filters (3
uM)
Tri-Nuclear filters (10 uM) 1.8 1.9 Secondary Filters Sludge Lance cartridge filters High Conductivity Waste Tank (HCWT)bag filters Secondary Bead Resins AVT Bead Resins
REFERENCES:
A current set of DOT,
- NRC, EPA, New York State, Volume Reduction Facility, and Disposal Site regulations and requirements shall be maintained at the Plant and readily available for reference.
A listing of most of these references is given in Attachment l.
(Reference NRC IEB 79-19)
IV RESPONSIBILITIES:
The Health Physicist responsible for Radwaste has the overall responsibility for implementing the PCP.
Implementing responsibilities include the following:
a)
Preparing and reviewing implementing procedures.
b)
Ensuring that radioactive waste is characterized and classified in accordance with 10 CFR Part 61.55 and Part 61.56.
c)
Providing a point of contact with any volume reduction facility or disposal site.
'f 5 ~
I
~1
Process Control Program Page 3
d)
Maintaining records of on-site and off-site sample analysis of waste streams.
DEFINITIONS:
Batch:
An isolated quantity of feed waste to be processed having essentially constant physical and chemical characteristics.
Chelatin A ents:
The NRC Branch Technical Position 1983 refers to EDTA, DPTA, hydroxy-carbolic acids, citric acid, carbolic acid and glucinic acid as chelating agents'ewatered Waste:
Wet waste that has been processed by means other than solidification or encapsulation to meet the free standing liquid requirements of 10 CFR Part
- 61. 56 (a) (3) and (b) (2)
Enca sulation:
A means of providing stability for certain waste types by surrounding the waste with an appropriate encapsulation media.
Gamma-S ectral-Anal sis: Using gamma spectroscopy to identify radioisotopes.
Homo eneous:
Uniformly distributed throughout the waste material.
Most waste streams are considered homogeneous for purposes of waste classification.
Low-Level Radioactive Waste LLRW
- Waste not classified as High-Level radioactive waste, transuranic
- waste, spent nuclear fuel, or by-product material as defined in section 11e. (2) of the Atomic Energy Act (uranium or thorium tailings and waste).
Radioactive wastes that are acceptable for disposal in a land disposal facility comprised of source, special nuclear, or by-product material.
/
radioactive waste into a form that is acceptable to a disposal facility.
Scalin Factor:
A dimension less number which relates the concentration of one easily measurable nuclide (usually a
gamma emitter) to another nuclide that is usually difficult to measure (usually a beta and/or alpha emitters)
Si nificant uantit
- For the purposes of waste classification the following radionuclide values shall be considered significant and must be reported on the disposal manifest:
Process
- Contro2, Program Page 4
a)
Any real value or the LLD for H-3, C-14, Tc-99 or I-129.
b)
Greater than or equal to 1 percent of concentration limits listed in 10 CFR 61 '5 Table 1.
c)
Greater than or equal to 1 percent of the Class A
concentration limits listed in 10 CFR 61.55 Table 2
except isotopes having half-lives less than 5 years.
d) 7 uCi/cc for Isotopes with half-lives less than 5
years and radionuclides not listed in 10 CFR 61.55.
e)
Greater than or equal to 1 percent of the total activity.
f)
Greater than or equal to 1 percent of the Reportable Quantity limits listed in 49 CFR Part 172.101 Table 2
S ecial Radionuclides:
The RADMAN computer code term used to describe 10 CFR 20, Appendix F, required radionuclides.
1 integrity under the expected conditions of disposal.
Traininc[: A systematic program that ensures a person has knowledge of contaminated hazardous materials and contaminated hazardous materials regulations.
Waste Container:
A vessel of any shape,
- size, and composition used to contain the final processed waste.
Waste Form: Waste in its container, in a final form, that is acceptable for disposal at a licensed disposal facility.
Waste Stream:
A Plant specific source of waste with a distinct radionuclide content and distribution.
form tied to a specific waste.
VI PRERE UISITES:
Maintenance of Regulatory Material A current set of DOT,
'eduction Facility, and Disposal Site regulations and requirements shall be maintained at the Plant and readily available for reference.
0
Process Control Program Page 5
Representative Radionuclide Sample Data Current representative radionuclide sample data shall be on file -for each active waste stream.
When operating conditions change or processing methods require increased sampling frequency, new sample data will be generated.
Data is considered to be current if iv meets the following:
a)
NRC Class A waste streams sampled within last two years.
b)
NRC Class B or Class C waste streams and waste streams that have the potential to be NRC Class B or Class C sampled within last year.
(Reference Branch Technical Position Paper 1983) 3.
Initial and Cyclic Training A training program is developed, implemented and maintained for all plant personnel involved in processing, packaging and handling of radioactive waste (Reference NRC IEB 79-19).
Training is required for each person who:
a)
Classifies hazardous materials b)
Packages hazardous materials c)
Marks and labels packag'es containing hazardous materials d)
Prepares shipping papers for hazardous materials e)
Marks or placards transport vehicles f)
Inspects or tests packages or transport vehicles Only trained personnel will characterize or package radioactive waste or radioactive materials.
VII RE UIREMENTS FOR PROCESSING VENDORS:
2.
The Health Physicist responsible for Radwaste or a designee shall review vendor(s) topical reports and test procedures.
Plant PORC shall assure that the vendor's operations and requirements are compatible with the responsibilities and operation of the Plant.
I
Process Control Program Page 6
Vendors shall provide documentation they maintain required training programs for items specified in section VI.3.
4.
Vendors providing solidification processes shall provide documentation that tests for stability of the final product have been completed successfully and demonstrate that they meet the criteria given in Attachment 2.
VZZZ WASTE MANAGEMENT PRACTICES:
1.0 Waste Stream Sam lin The following general requirements apply to Plant waste stream sampling:
a)
Each waste stream is treated separately for classification purposes.
b)
The minimum sample frequency requirements are met.
c)
Samples are representative of the waste stream.
d)
An in-house analysis is performed for gamma emitting radionuclides for each sample sent to an independent laboratory.
e)
As waste is generated, an in-house analysis is performed for gamma emi'tting radionuclides to compare to the current database values.
The current database is usually the most recent independent laboratory results.
1.2 Changes in plant operating conditions which may affect waste streams may require increasing sampling frequencies.
Increased sampling requirements are given in implementing procedures.
1.3 Infrequent or abnormal waste types requirements are:
a)
The Health Physicist responsible for Radwaste or designee shall determine if the waste can be correlated to an existing waste stream.
b)
The Health Physicist responsible for Radwaste or designee shall ensure specific sampling and analysis requirements are met to properly classify the material.
Process Control Program Page 7
c)
The Health Physicist responsible for Radwaste or designee shall resolve any discrepancies between in-house results and the independent laboratory results for the same or replicate samples.
d)
The Health Physicist responsible for Radwaste or designee shall maintain records vf on-site and off-site sample analysis and evaluations of waste streams.
Scaling factors are established where by concentrations of radionuclides which cannot be readily measured are estimated through ratioing to radionuclides which can be readily measured a)
Scaling factor relationships are developed on a waste stream specific basis.
These relationships are periodically revised to reflect current independent laboratory data from direct measurement of samples.
b)
Hard to detect activation product radionuclides and C-14 are estimated by using scaling factors based on measured Co-60 activities.
c)
Hard to detect fission product radionuclides are estimated by using scaling factors with measured Cs-137 activities.
Hard to detect transuranic radionuclides are estimated by using scaling factors with measured Ce-144 activities.
When Ce-144 cannot be readily
- measured, transuranics are estimated by using Cs-137 activities.
e)
As a minimum, the following radioisotope activities will be determined by direct measurement or calculated from developed scaling factors.
H-3 C-14 Cr-51 Mn-54 Fe-55 Co-57 Co-58 Fe-59 Ni-59 Co-60 Ni-63 Zn-65 Sr-89 Sr-90 Nb-94 Nb-95 Zr-95 Tc-99 Ru-103 RU-106 Ag-108m Ag-110m Sb-124 Sb-125 I-129 I-131 Cs-134 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Np-237 PU-238 Pu-239 Pu-240 Pu-241 Am-241 Cm-242 Cm-243 Cm-244 H-3 activity in any waste stream should not exceed the concentration as indicated in the primary coolant
y' '*
A~
Process Control Program Page 8
1.5 The activity for each waste stream is determined by using
direct measurement of radionuclides or gross radioactivity measurement.
Current practices are:
DAW - Gross radioactivity measurement in conjunction with hand calculation or computer codes, for example:
RADNAN or TRASHP.
Filters
- Gross radioactivity measurement in conjunction with hand calculation or computer codes, for examp2.e:
FILTRK.
Soil - Direct measurement of radionuclides in conjunction with hand calculation or computer codes.
All Other Waste Streams
- Direct measurement of radionuclides in conjunction with hand calculation or computer codes.
2.0 Waste Classification Determination of the NRC waste classification is performed by comparing the measured or calculated concentrations of significant radionuclides in the final waste form to those listed in 10 CFR 61.55 and burial site criteria.
3.0 Waste Packa in 3.1 All disposal liners are manu'factured by 'and purchased from QA approved vendors.
3.2 3.3 All casks are manufactured by, leased, rented or purchased from QA approved vendor's.
All services purchased from QA approved vendors.
3.4 DOT Specification 7A, Type A Packaging for shipping and disposal are purchased to meet the requirements of 49 CFR 173.415.
3.5 Strong Tight containers for shipping and disposal are either purchased from a vendor or constructed in the Plant.
3.6 Each container of waste being shipped directly to a disposal site shall be clearly labeled to identify it as
'Class A waste, Class B waste or Class "C" waste in accordance with 10 CFR 61.55.
Process Control Program Page 9
3.7 Void spaces within the waste and between the waste and its packaging are reduced to the extent practicable to be less than or equal to 15 percent by volume or a variance is obtained from the disposal site.
4.0 Chemical Control Constituents that may cause problems with cement stabilization, filter break down, anaerobic activity, or generation of mixed waste are excluded from introduction into the waste streams.
Any new constituents that may be directly or inadvertently introduced into any waste stream will be evaluated for their effect prior to use.
5.0 Volume Reduction 5.1 Radioactive material is sent to a vendor for volume reduction processing instead of directly to a disposal site when possible.
5.2 Procedures implementing the PCP specify the requirements for waste classification before radioactive material is sent to vendor facilities.
5.3 Procedures implementing the PCP do not address the requirements for 10 CFR 61.56 Waste Characteristics for material sent to intermediate processors since the final treatment and packaging is p'erformed at the vendor facility.
5.4 The types of radioactive material sent to intermediate processors may include but is not limited to the following:
DAW Low Activity Bead Resin Low Activity Activated Charcoal Contaminated Soil Contaminated metal 6.0 Solidification Stabilization and Enca sulation 6.1 Solidification, stabilization, or encapsulation is performed in accordance with Ginna or vendor operating procedures and instructions.
The use of applicable test data (for examp2,e:
Topical Reports) may be used for Process Control Program qualification.
The test data must demonstrate that the requirements given in Attachment IV are met.
Process Contro2, Program Page ZO 6.2 Solidified Class A waste, separated from Class B and C
- waste, need only demonstrate the product is'a, free sta'nding monolith with no more than 0.5 percent of the volume as free liquid.
6.3 Solidified Class A waste not separated from Class B
and/or Class C waste, Class B and Class C waste shall meet the stability requirements of 10 CFR 61.56 (b)
En order to ensure that Class B and/or Class C waste or its container will maintain its stability, the following conditions need to be met:
a)
The waste shall be a solid form or in a container or structure that provides stability after disposal.
b)
The wastes shall not contain free standing or corrosive liquids that exceeds 1 percent by volume in containers designed to provide stability, or 0.5 percent by volume for solidified waste as measured using the method in ANS 55.1.
c)
The waste or container should be resistant to degradation caused by radiation effects.
d)
The waste or container should be resistant to biodegradation.
e)
The waste or container should remain stable under the compressive loads inherent in the disposal environment.
The waste or container should remain stable if exposed to moisture or water after disposal.
g)
The as-generated waste should be compatible with the solidification media or container.
6.4 All waste must be stabilized if it contains isotopes with greater than 5 year half-lives and the total specific activity is greater than 1 microcurie/cubic centimeter.
This includes, but is not limited to, ion exchange resins, filter media,
- sludge, liquids, biological waste and dry active waste.
Specific requests to bury DAW which exceeds the 1 microcurie/cubic centimeter, limit which has not'een stabilized may be considered by the disposal facility and NRC
~
Process Control Program Page 12.
Dewaterin 7.1 Wet radioactive material is dewatered by 'allowing water to drain from the material completely.
As water is collected it is removed by gravity drain or suction.
Material may be transferred to a vendor for further dewatering, volume reduction or incineration.
Material prepared for shipment to a burial facility will be dewatered to contain less than or equal to one percent liquid by volume.
7.2 Dewatered resins with activities which will produce greater than 1 8+08 rads total accumulated dose over 300 years will not be shipped off-site.
This is usually verified by comparing the specific activity at time of shipment to the following concentration limits:
10 curies per cubic foot or 350 micro curies per cubic centimeter 8.0 Radioactive Material Shi ments 8.1 Procedures implementing the PCP address the requirements for waste classification.
8.2 Procedures implementing the PCP address the requirements for waste characteristics.
9.0 ualit Control for Sam lin
'. and Classification 9.1 The RADMAN computer code provides a mechanism for conducting a quality control program in accordance with waste classification requirements listed in 10 CFR 61.55.
All waste stream sample data changes are written to a computer data file for comparison review and future reference.
9.2 Audit frequency and management review includes:
a)
Periodic audits which must include management reviews.
b)
Periodic management audits of the Ginna' sampling and classification program to verify the adequacy of maintenance sampling and analysis.
Process Control Program Page 22 c)
Periodic audits are performed and documented by any one of the following:
Radiation Protection Group Corporate Radiation Protection Staff Quality Assurance Group Qualified Vendors d)
The Plant audit program is detailed in QA-1802 QA Audit and Scheduling.
9.3 General requirements for vendor packages are:
a)
Perform all inspection, handling and loading operations per approved Ginna and/or vendor instructions and procedures.
9.4 b)
Store each container with its designated closure assemblies to prevent mismatching.
General requirements for inspection prior to use include:
a)
Ensuring all containers are loade'd to meet the requirements of test and evaluation, for example:
WHC-EP-0558 Test and Evaluation Document for DOT Specification 7A Type A Packaging.
b)
Visually inspecting thread and seal areas to verify they are free of foreign matter that could impair the thread or seal engagement.
c)
Visually inspecting the exterior surfaces for damage that may have occurred during transport or storage that could lessen the container integrity.
d)
Visually inspecting the lid for sufficient closure.
INTERIM STORAGE:
Waste is collected, packaged and stored at designated locations until shipment is arranged.
In the event that access to a disposal facility is denied:
Implementing procedures will describe the processing, packaging and storage of waste generated at the plant.
Implementing procedures will describe the receipt and storage of processed waste returned from a vendor following volume reduction.
Process Contro2, Program Page 13 DAW will be packaged and stored in the Upper Radwaste Storage Building (URWSB).
The waste material remaining from the DAW shipped to a vendor for processing and returned to RG&E will be stored in the URWSB.
Low activity filter media will be packaged and may be stored either in the URWSB or the HIC Open Storage Area to be constructed.
The waste material remaining from low activity filter media shipped to a vendor for processing and returned to RGRE may be stored either in the URWSB or the HIC Open Storage Area.
High activity filter media will be packaged and stored in the HIC Open Storage Area.
The waste material remaining from high activity filter media which is shipped to a vendor for processing and returned to RGEE will be stored in the HIC Open Storage Area.
ATTACHMENTS:
Figure 1, Block diagram of typical waste processing system Figure 2, Block diagram of typical dewatering system Attachment 1, References Available for Radwaste Processing and Shipments Attachment 2, Tests for stability
Process Control Program Page FlGURE 1
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v)~>
I ) M OCM~
jl(118 hp~
ilfg Qfj cC L3 CC O
I
<<C LD cd
SPENT RESIN TANK FIGURE 2 SAMPLE Q0 0
CONTAINER WHUT BURIAL OR STORAGE HIGH INTEGRITY CONTAINER DEWATER PUMP LICENSED CASK
Process Control Program Page 16 ATTACHMENT 1 Page 1 of 9
REFERENCES AVAILABLEFOR RADWASTE PROCESSING AND SHIPMENTS 1.0 1.2 1.3 1
~ 4 1.5 1.6 1.7 FEDERAL REGULATIONS Code of Federal Regulations, Title 10, Part 20 Code of Federal Regulations, Title 10, Part 61 Code of Federal Regulations, Title 10, Part 70 Code of Federal Regulations, Title 10, Part 71 Code of Federal Regulations, Title 29, Part 1900.1200 Code of Federal Regulations, Title 40, Parts 240 Through 272 Code of Federal Regulations, Title 49, Parts 100 Through 180 1.8 Code of Federal Regulations, Title 49, Parts 350 Through 399 2.0 NRC 2.1 2.2 2.3 2.5 2.6 2.7 2.8 INFORMATION NOTICES NRC Information and Enforcement Bulletin 79-19:
Packaging of Low-Level Radioactive Waste for Transportation and Burial.
Low-Level Radioactive Waste Burial Criteria.
NRC Information Notice 80-32: Clarification of Certain Requirements for Exclusive-Use Shipments of radioactive Materials.
NRC Information Notice 80-32, Rev 1: Clarification of Certain Requirements for Exclusive-Use Shipments of Radioactive Materials.
NRC Information Notice 82-47: Transportation of Type A Quantities of Non-Fissile Radioactive Material.
NRC Information Notice 83-05: Obtaining Approval for Disposing of Very-Low-Level Radioactive Waste 10 CFR Section 20.302.
NRC Information Notice 83-10: Clarification of Several Aspects Relating to Use of NRC-Certified Transport Packages.
Dewatered Spent Ion Exchange Resin Susceptibility to Exothermic Chemical Reaction.
,I ~
~ I'*
Wg
~
Process Control Program Page Z7 Attachment I Page 2 of 9
2.9 NRC Information Notice 83-33: Non-Representative Sampling of Contaminated Oil.
2.10 2.11 2.12 NRC Information Notice 84-14: Highlights of Recent Transport Regulatory Revisions by DOT and NRC.
NRC Information Notice 84-50: Clarification of Scope of Quality Assurance Programs for Transport of Packages Pursuant to 10 CFR 50 Appendix B.
NRC Information Notice 84-72: Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation.
- 2. 13 2.14 NRC Information Notice 85-92:
Surveys of Wastes Before Disposal from Nuclear Reactor Facilities.
NRC Information Notice 86-20: Low-Level Radioactive Waste Scaling Factors, 10 CFR 61.
2.15 NRC Information Notice 86-90:
Requests to Dispose of Very Low-Level Radioactive Waste Pursuant to 10 CFR 20.302.
2.16 2.17 NRC Information Notice 87-03: Segregation of Hazardous and Low-Level Radioactive Wastes.
t NRC Information Notice 87-07: Quality Control of On-Site Dewatering/Solidification Operations by Outside Contractors.
2.18 NRC Information Notice 87-31: Blocking, Bracing, and Securing of Radioactive Materials Packages in Transportation.
2.19 2.20 NRC Information Notice 88-08:
Chemical Reactions With Radioactive Waste Solidification Agents.
NRC Information Notice 88-16: Identifying Waste Generators in Shipments of Low-Level Waste to Land Disposal Facilities.
2.21 NRC Information Notice 88-62:
Recent Findings Concerning Implementation of Quality Assurance Programs by Suppliers of Transport Packages.
2.22 2.23 NRC Information Notice 89-13: Alternative Waste Management Procedures in Case of Denial of Access to Low-Level Waste Disposal Sites.
NRC Information Notice 89-27: Limitations on the Use of Waste Forms and High Integrity Containers for the Disposal of Low-Level Radioactive. waste.
l
Process Control Program Page Z8 Attachment I Page 3 of 9
2.24 NRC Information Notice 90-09:
Extended Interim Storage of Low-Level Radioactive waste by Fuel Cycle and Materials Licensees 2.25 NRC Information Notice 90-35: Transportation of Type A Quantities of Non-Fissile Radioactive Materials.
2.26 NRC Information Notice 90-50: Minimization of Methane Gas in Plant Systems and Radwaste Shipping Containers.
2.27
- NRC Information Notice 90-82: Requirements for Use of Nuclear Regulatory Commission Approved Transport Packages for Shipment of Type A Quantities of Radioactive Material.
2
~ 28 NRC Information Notice 91-03:
Management of Wastes Contaminated with Radioactive Materials
( "Red Bag" Waste and Ordinary Trash).
2.29 NRC Information Notice 91-35: Labeling Requirements for Transporting Multi-Hazard Radioactive Materials.
2.30 NRC Information Notice 91-65:
Emergency Access to Low-Level Radioactive Waste Disposal Facilities.
2.31 2.32 NRC Information Notice 92-62:
Emergency
Response
Information Requirements for'adioactive Material Shipments.
NRC Information Notice 92;72:
Employee Training and Shipper Registration Requirements for Transporting Radioactive Materials.
3.0 NRC REGULATORY GUXDES 3.1 Regulatory Guide 7.1: Administrative Guide for Packaging and Transporting Radioactive Material.
3.2 Regulatory Guide 7.2:
Packaging and Transportation of Radio-actively Contaminated Biological Materials.
3.3 Regulatory. Guide 7.4:
Leakage Test on Packages for Shipment of Radioactive Materials.
3.4 Regulatory Guide 7.5: Administrative Guide for Obtaining Exemptions From Certain NRC Requirements Over Radioactive Material Shipments.
l
~
)
Process Control Program Page 29 Attachment I Page 4 of 9
3.5 Regulatory Guide 7. 7: Administrative Guide for Verifying Compliance With Packaging Requirements for Shipments of Radioactive Materials.
3.6 3.7 Regulatory Guide 7.9, Rev 1: Standard Format and Content of Part 71 Applications for Approval of Packaging of Type B, Large Quantity and Fissile Radioactive Material.
Regulatory Guide 7 '0, Rev 1: Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material 4.0 NRC TECHNICAL POSITION 4.1 4.2 Final Waste Classification and Waste Form Technical Position Papers, May 11, 1983.
Technical Position on Waste
- Form, Rev.
1, January 1991.
5.0 ANSI STANDARDS 5.1 ANSI/ANS-40.35-1991:
Volume Reduction of Low-Level Radioactive Waste or Mixed Waste.
5.2 ANSI/ANS-55.1-1979: American National Standard for Solid Waste Processing System for Light Water Cooled Reactor Plants.
5.3 5.4 5.5 ANSI N14.9.1-1976:
Packaging for Transportation of Liquid Aqueous Radioactive Wastes From Nuclear Power Plants.
ANSI N679-1976:
Guide for Writing Operating Manuals for Radioactive Materials Packaging.
ANSI N14.7-1975:
Guide to Design and Use of Shipping Packages for Type A Quantities of Radioactive Materials.
6.0 6.1 6.2 R.E.
GINNA TECHNICAL SPECIFICATIONS Section 6.16, Process Control Program (PCP)
Section 6.17, Major Changes to Radioactive Waste Treatment Systems.
7.0 R.E.Ginna Updated Final Safety Analysis Report Section 11.4, Solid Waste Management System.
Process Control Program Page 20 Attachment I Page 5 of 9
8.0 REPORTS 8.1 8.2 8.3 8.4 8.5 8.6 EPRI, Interim On-Site Storage of Low-Level Waste, Volume 4: Waste Containers for Extended Storage, March 1992.
E
- EPRI, Radwaste Desk Reference, Volume 2: Transportation and Disposal, June 1991.
- EPRI, NP-4938, Methodology for Calculating Combustible Gas Concentration in Radwaste Containers, March 1987
- EPRI, NP-5977, Radwaste Radiolytic Gas Generation Literature Review, Spetember, 1988 NUMARC, A Technical Basis for Meeting the Waste Form Stability Requirements of 10 CFR 61.
WHC-EP-0558:
Test and Evaluation Document for DOT Specification 7A Type A Packaging 9.0 10.0 10.1 10.2 INPO GOOD PRACTICES:
Low Level Radioactive Waste Management.
DISPOSAL SITE REQUIREMENTS S20-AD-010: Barnwell Waste Management Facility Site Disposal Criteria.
Barnwell (South Carolina)
Disposal Site License.
11.0 11.1 11.2 11.3 PLANT PROCEDURES A-101 Ginna Quality Assurance Program Implementation A-805 Control of Consumable Material at Ginna Station QCIP-21 Inspection of Shipping Packages (Casks) for Radioactive Material 11.4 11.5 11.6 QCIP-21.1 Inspection of Shipping Packages (Casks) for Westinghouse Surveillance Capsule QCIP-21.4 Radlock Inspection Procedure QCIP-21.6 Inspection of Exclusive Radioactive Material Transport Vehicles
S C, l
'I
Process Control Progran Page 22, Attachment I Page 6 of 9
11.7 11.8 11.9 11.10 QCIP-700 Quality Control General Acceptance Criteria QCIP-701 Receipt Inspection Process S-4.1.27 Chem Nuclear Waste System Operation, Monitor Tank A Recirc Thru Liquid Waste Processing System.
S-4.1.30 Chem Nuclear Waste System Operation/Sampling/Isolation and Records-Chem Nuc Technician Procedure 11.11 S-4.1.31 Transfer of WHUT Thru Chem Nuclear Waste System to A Monitor Tank 11.12 S-4.4 Spent Resin Removal to Shipping Casks 11.13
- 11. 14 11.15 11.16 11.17 11.18 11.19 11.20 S-4.5 Sluicing Waste Condensate Polishing Demineralizer Spent Resin to Shipping Cask S-4.5.1 Sluicing Waste Evaporator Distillate Demineralizer Spent Resin to a Strong Tight Container II RP-RW-PROC-CN120A/B, (RD-10.14) Handling, Loading and Unloading of Chem-Nuclear 8-120 A or B(U) Transport Cask RP-RW-PROC-HIC, (RD-16.5)
Waste Solidification in Chem-Nuclear
- Systems, Inc. Polyethylene High Integrity Container RP-RW-PROC-CN14/195, (RD-10.18) Handling, Loading and Unloading of Chem-Nuclear CNSI 14-195H Transport Cask RP-RW-PROC-NP142, (RD-10.19) Handling, Loading and Unloading of NUPAC Model 10-142 Transport Cask RPA-RW-SHIP-WSTE, (RD-10.21) Preparation and Shipment of Radioactive (Waste) Material RP-RW-PROC-CNFLTR120, (RD-10.22) Transfer and Loading of Process Filters Using CNS 8-120A/B or 14-195H Transport Cask 11.21 RPA-RW-INV, (RD-10.23)
Radwaste Inventory 11.22 11.23 RP-RW-INV-SMPLG, (RD-10.24) Representative Sampling of Radioactive Material RP-RW-SHIP-VR, (RD-10.27) Preparation and Shipment of Radioactive Materials to a Volume Reduction Facility
4 g
I+il
Process Control Program Page 22 Attachment I Page 7 of 9
11.24 RPA-RW-FLTR-INV, (RD-10.28) Characterization and Inventory Tracking of Filter Assemblies Prior to Shipment
- 11. 25 RPA-RW-TRN, (RD-10.31) Training and Responsibilities of Individuals Involved in Radwaste Group Activities 11.26 RP-PCP-DEWATER, (RD-16.3)
Dewatering Wet Solid Wastes 11.27 RP-RW-PROC-GAS, (RD-16.4)
Combustible Gas Detection in or Near Radioactive Waste Liners/Containers 12.0 VENDOR PROCEDURES, LETTERS AND NOTES NOTE The operational aspects of vendor procedures are referenced or incorporated into site-specific procedures.
12.1 Pacific Nuclear Systems Inc.,
Q.A. Manual, latest revision.
12.2 12.3 12.4 12.5 Chem-Nuclear
- Systems, Inc.,
Q.A. Program Document No.
QA-AD-001, latest revision Scientific Ecology Group, Inc., Q.A. Program No.
SEG/QA-100, latest revision "RADMAN - A Computer Code",
Main Topical Report NRC acceptance Letter:
RADMAN Topical Report, July 25, 1983 12.6 12.7 WMG/P-045: Operating Procedures for RADMAN Software WMG-9006:
WMG Computer Program Dose-to-Curie Methodology Verification and Validation 12.8 12.9 WMG/P-007: Operating Procedures for FILTRK Software WMG/P-009: Operating Procedures for RAMSHP Software 12.10 WMG/P-010: Operating Procedures for TRASHP Software 12.11 WMG-QA-011:
WMG Quality Assurance Program 12.12 Chem-Nuclear
- Systems, Inc., Topical Report, Dewatering Control Process Containers, CNSI-DW-11118-01-NP-A
Process Control Program Page 23 Attachment I Page 8 of 9
Chem-Nuclear
- Systems, Inc., Topical Report, Polyethylene High Integrity Containers, CNSI-HIC-14571-01-NP TR-OP-030:
Handling Procedure for CNS~ Transport Cask Number 14-170 Series III TR-OP-022:
Handling Procedure for CNSI Transport Cask Number CNS 14-170 Series II FO-OP-023:
Bead Resin/Activated Carbon Dewatering Procedure for CNSI Liners TR-OP-008: Operation and Maintenance Manual for CNS 14-195H and Typical Trailers FO-AD-002: Operating Guidelines for Use of Polyethylene High Integrity Containers TR-MN-005: Gasket/Seal/0-Ring Replacement Repair Procedure for CNSI Cask Fleet 12.20 12.22 1'2. 23 SD-OP-048:
Process Control Program and Operating Procedure for InSitu Solidification of Suspended Objects Nuclear Packaging, Inc., Topical Report No. TP-02-NP-A, describes dewatering system OM-101: Operation and Maintenance Manual for NuPac Model 10-142 and Typical Trailers LT-29: Seal integrity Test of the NuPac 10-142 Transport Cask 12.24 12.25 12.26 12.27 12.28 H-18: Handling, Shipping, and Storage for the NuPac CL-200 Polyethylene High Integrity Container H-19 Off-Loading for NuPac Cl-200 Crossed linked Polyethylene High Integrity Container OM-16: Users Guide for the NuPac Crossed linked Polyethylene High Integrity Containers OM-43: Operating Procedure for Resin Drying (Dewatering)
System OM-46: Handling, Shipping and Storage for NuPac 14/190, and 14/210 14-Drum Casks Shielded Shipping Container Lt-04: General Procedure for Soap Bubble (Low Pressure)
Leak Test
Process Control Program Page 24 Attachment I Page 9 of 9
12.30 OM-10 Installation and Torquing of NuPac Ratchet Binders 12.31 RSM-022:
SEG Rad Services Manual for 14-170 12.32 RSM-021:
SEG Rad Services Manual for LN-142 12.33 RSM-018:
SEG Rad Services Manual for 14-215 12.34
- 12. 35 12.36
- 12. 37 RSM-006:
SEG Rad Services Manual for 3-82B Shielded Transportation Cask RSM-009:
SEG Rad Services Manual 10-142 Shielded Transportation Cask TR-OP-003:
Handling Procedure for Chem-Nuclear
- System, Inc., Transport Cask 8-120A TR-OP-035:
Handling Procedure for Chem-Nuclear
- System, Inc., Transport Cask CNS 8-120B.
International Air Transport Association, Dangerous Goods Regulations, January 1,
1993
3 Process Control Program Page 25 ATTACHMENT 2 Page 1 of 2
Tests for stability of solidified wastes to be demonstrated by vendor:
a)
Waste specimens shall be prepared on the proposed waste stream to be solidified and based on the range of waste stream chemistries expected.
b)
Solidified waste specimens should have compressive strengths of at least 50 psi when tested in accordance with ASTM C39 or ASTM D1074 when a bituminous product.
c)
Waste specimens should be exposed to a minimum of E+8 Rads in a gamma irradiator or equivalent.
The specimens shall have a minimum compressive strength of 50 psi following irradiation as tested in accordance with ASTM C39 or ASTM D1074.
d)
Specimens shall be tested for resistance to biodegradation in accordance with both ASTM G21 and ASTMG22.
Following biodegradation
- testing, specimens shall have compressive strengths greater than 50 psi as tested using ASTM39 or ASTM D1074 e)
Leach testing shall be performed for a minimum of 90 days in accordance with the procedure in ANS 16.1.
In
- addition, synthesized sea water leachant will be tested and radioactive tracers utilized in performing leach tests.
The leachability index, as calculated in accordance with ANS 16.1, shall be cfreater than 6.
Waste specimens shall have a compressive strength of 50 psi in accordance with ASTM39 or ASTM D1074, following immersion for a minimum of 90 days.
Immersion may be performed in conjunction with the leach testing.
g)
Waste specimens shall be resistant to thermal degradation.
The heating and cooling chambers used for the thermal degradation testing shall conform to the description given in ASTM B553, section 3.
Samples shall be placed in the test chamber and a series of 30 thermal cycles carried out in accordance with Section 5.4.1 through 5.4.4 of ASTM B553.
The high temperature shall be 60C and the low temperature limit-40C.
Following testing the waste specimens shall have compressive strengths greater than 50 psi as tested in accordance with ASTM39 or ASTM D1074.
h)
Waste specimens shall have less than 0.5 percent by volume as free liquids as measured using the method used in ANS 55. 1.
Free liquids shall have a pH between 4 and 11
Process Contro2.
Program Page 26 Attachment II Page 2 of 2 Zf small, simulated laboratory size s".ecimens are used for the above testing, test data from sections or cores of the anticipated full-scale products should be obt'ained to correlate the characteristics of actual size products with those of simulated laboratory size specimens.
The testing may be performed on non-radioactive specimens.
The full scale specimens shall be fabricated using actual or comparable solidification equipment.
Waste samples from full-scale specimen shall be destructively analyzed to ensure that the product produced is homogeneous to the extent that all regions in the product can expect to have compressive strengths greater than 50 psi.
Full-scale specimens may be fabricated using simulated non-radioactive
- products, but shall be fabricated using actual solidification equipment.