ML18005B022
| ML18005B022 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/30/1989 |
| From: | Pollard A CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18005B021 | List: |
| References | |
| NUDOCS 8909050096 | |
| Download: ML18005B022 (89) | |
Text
Carolina Power a Light Shearon Harris Nuclear Power Plant License No. NPF-063 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January 1, 1989 to June 30, 1989 Prepared by:
Q Prospect Specialist Radiation Control Reviewed by:
snag Environmental a Radiation Control Approved by:
P ant neral Nan er csP S909050096 890S29 PDR ADOCK 05000400 PDC
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Table of Contents Introduction Discussion Appendix 1.
Supplemental Information Page No.
Appendix 2.
Effluent and Waste Disposal Report 1.
Lower Limits of Detectability (LLD's) 2.
Effluents Released 3.
Solid Waste Disposal 2/1 2/3 2/11 Appendix 3.
Changes to Process Control Program 3/1 Appendix 4.
Changes to Offsite Dose Calculation Manual (ODCM)
Appendix 5.
Changes to the Environmental Monitoring Program 1.
Environmental Monitoring Program 2.
Land Use Census 5/1 5/2 Appendix 1.
2.
3.
4.
6.
Additional Technical Specification Responsibilities Inoperability of Liquid Effluent Monitors 6/1 Inoperability of Gaseous Effluent Monitors 6/3 Unprotected Outdoor Tanks Exceeding Limits 6/5 Gas Storage Tanks Exceeding Limits 6/6 Appendix 7.
Major Modifications to Radwaste System
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Introduction This Semiannual Radioactive Effluent Release Report is in accordance with Technical Specification 6.9.1.4 to the Shearon Harris Nuclear Power Project (SHNPP) Operating License No. NPF-63. It provides effluent monitoring information obtained in fulfillment of the plant's Radiological Effluent Technical Specifications (RETS).
The Shearon Harris Nuclear Power Plant achieved initial criticality on January 3, 1987.
This report covers the period from January 1, 1989 to June 30, 1989, during which the plant was in cycle 2 operation with a capacity factor of 94.09%.
Discussion Appendices 1 and 2:
The information on gaseous and liquid effluents and solid waste is given in accordance with Regulatory Guide 1.21 (Rev.
- 1) Appendix B format. As required by Technical Specification 6.9.1.4, the solid waste table has been supplemented to include 10CFR61 class, type of container and solidification agent or absorbent.
Except for continuous noble gas releases, liquid and gaseous average concentrations (uCi/cc) and total curies released are for only those nuclides that were positively identified. If no activity for a nuclide is reported for a quarter, the Lower Limit of Detection (LLD) tables show a typical sensitivity level for detection of the nuclide.
Continuous noble gas effluent activities were based on hourly average stack monitor readings (in uCi/cc) and stack flow rate estimates based on design fan flow rates.
No specific noble gas nuclides were identified in any of the stack gas grab samples taken for characterizing continuous gaseous releases.
Therefore, the total noble gas activities are based on stack monitor readings and apportioned as per the GALE code (NUREG 0017) nuclide assumptions as given in the ODCM.
Nuclides reported in gaseous batch releases are determined from the isotopic analysis for each batch release.
One unplanned gaseous release occurred during this Report period.
The release occurred during a nonroutine sampling of the Volume Control Tank gas space and was caused by a leaking sample apparatus.
The release was monitored by Plant Vent Stack 1 radiation monitors and was estimated to be approximately 3.3 Ci of Noble Gas. Off-site doses were evaluated with resp ct to 10CFR20 (dose rate) and 10CFR50 (total dose) limits at the site boundary.
The highest dose was calculated to be approximately 0.08% of its respective limit.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Discussion (continued)
Appendices 1 and 2 (continued)
No activity above background was detected in any continuous liquid release pathway.
Therefore, the reported activities are the summation of nuclides in batch releases only.
A total of 59.3 m
of solid waste, containing 3.702 Ci of radioa~tivity, 3
was shipped for burial during the Report period, compared with95.0 m and 8.200 Ci shipped during the previous Report period.
Appendix 3:
No changes to the Process Control Program (PCP) were made during this Report period.
Appendix 4:
Revision 2 of the ODCM, issued in January
- 1988, was substantially revised during the Report period.
As a result, this Appendix provides Revision 3 in its entirety rather than just a package of the affected pages.
The changes
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reflected in Revision 3 are intended to provide improvements in the accuracy and reliability of the dose calculations and setpoint determinations.
Appendix 5:
Minor changes were made to the Environmental Monitoring Program during this reporting period.
Based on more extensive meteorological data, an air sampler was relocated to a sector with a higher D/Q, and a new TLD location was added to provide coverage for the inner ring in the WSW sector.
The annual Land Use Census was performed during this Report Period.
As a result of the census, Table 3.2-2 of the ODCM was revised.
ODCM Table 3.2-2 provides the distances to the nearest special locations, i.e.,
residences, milk
- animals, gardens, and meat animals.
To determine whether any of the new special locations yielded calculated doses greater than the locations previously used, the 1989 land-use data was coupled with the SHNPP 1988 meteorology, and dose calculations were performed by the GASPAR program using the GALE source terms provided in Table 3.2-1 of the ODCM.
No significant differences in estimated doses occurred as a result of the 1989 changes in the locations of the nearest
- resident, garden, or meat animal.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Discussion (continued)
Appendix 6:
All effluent monitor inoperabilities greater than 30 days are given along with a brief explanation.
During these periods, compensatory sampling and flow rate estimations consistent with Technical Specification requirements have provided accountability and control of effluents.
Several changes made in Revision 3 of the ODCM regarding monitor setpoint methodologies will allow several of the liquid monitors that have been inoperable to be returned to service during the next reporting period.
No unprotected outdoor tank or gas storage tank exceeded Tech Spec limits during this Report period.
Appendix 7:
The Liquid Radwaste System has undergone major modifications during this Report period.
The system has been changed to include a vendor supplied modular fluidized transfer demineralization system as an alternate means of processing liquid radioactive waste.
Existing systems are in stand-by and have been placed in wet lay-up.
The changes are explained in detail in this Appendix.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 1: Supplemental Information 1.
Regulatory Limits A.
Fission and activation gases (1)
Calendar Quarter a.
5 mrad gamma b.
10 mrad beta (2) Calendar Year a.
10 mrad gamma b.
20 mrad beta B.
I-131, I-133, I-135, H-3 and particulates with half-lives greater than eight days (1)
Calendar Quarter a.
7.5 mrem to any organ (2)
Calendar Year a.
15 mrem to any organ C.
Liquid effluents (1)
Calendar Quarter a.
1.5 mrem to total body b.
5 mrem to any organ (2)
Calendar Year a.
3 mrem to total body b.
10 mrem to any organ
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 1 (Continued):
Supplemental Information 2.
Maximum permissible concentrations and dose rates which determine maximum instantaneous release rates.
A.
Fission and activation gases (1) 500 mrem/year to total body (2) 3000 mreay'year to the skin B.
I-131, I-133, I-135, H-3 and particulates with half-lives greater than eight days.
1500 mreny'year to any organ C.
Liquid effluents The concentration of radioactive material released in liquid effluents to unrestricted areas after dilution shall be limited to the concentration specified in 10CFR20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases.
For dissolved and entrained noble gases, the MPC shall be equal to 2.0E-4 uCif'ml.
3.
Measurements and Approximations of Total Radioactivity A.
Fission and activation gases Measurements by continuous monitors of activity concentrations times total stack flow, and analysis by gamma spectroscopy and liquid scintillation counting for specific radionuclides in representative grab samples.
B.
Iodines Continuous charcoal cartridge sampling and analysis by gamma spectroscopy for specific radionuclides times total stack flow.
C.
Particulates Continuous particulate sampling and analysis by gamma spectroscopy, alpha counting and radiochemical analysis for specific radionuclides times total stack flow.
D.
Liquid Effluents Pre-release representative sampling and analysis by gamma spectroscopy and liquid scintillation counting for specific radionuclides times total release volume.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 1 (Continued):
Supplemental Information 4.
Batch Releases A.
Liquid (1)
Number of batch releases:
(2)
Total time period for batch releases:
(3)
Maximum time for a batch release:
(4)
Average time for a batch release:
(5)
Minimum time for a batch release:
8.50 E+Ol 4.59 E+04 min.
7.62 E+02 min.
5.40 E+02 min.
1.00 E+00 min.
(6)
Average stream flow during periods of release:
7.11 E+03 gpm B.
Gaseous (1)
Number of batch releases:
1 (2)
Total time period for batch releases:
(3)
Maximum time for a batch release:
(4)
Average time for a batch release:
(5)
Minimum time for a batch release:
0.00 E+00 0.00 E+00 min.
0.00 E+00 min.
0.00 E+00 min.
0.00 E+00 min.
5.
Abnormal Releases A.
B.
Liquid No abnormal liquid releases were made in the period.
Gaseous On March 11, 1989 a unplanned release was made from the Plant Vent Stack 1. This resulted in an estimated 3.2 Ci of Noble Gases being released.
The isotope quantities are included in the Quarter 1 Batch Mode accountability table for fission and activation gases.
(Appendix 2, Enclosure 2, Table 1C).
1/3
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2: Effluent and Waste Disposal Report Enclosure 1
LOWER LIMITS OF DETECTION (LLD)
- 1. LLD's for Gaseous Effluents NUCLIDE H - 3 Ar-41 Cr-51 Mn-54 Co-58 Fe-59 Co-60 Zn-65 Kr-85 Kr-85m Kr-87 Kr-88 Rb-88 Sr-89 Sr-90 Nb-95 Mo-99 RQ-103 I -131 Xe-131m I -133 Xe-133 Xe-133m Cs-134 I -135 Xe-135 Xe-135m Cs-137 Xe-138 Ba-140 La-140 Ce-141 Ce-144 Gross Alpha LLD (uCi/cc) 5.37 E-09 4.42 E-08 1.64 E-13 1.79 E-14 3.48 E-14 4.40 E-14 2.51 E-14 4.71 E-14 8.56 E-06 2.17 E-08 3.82 E-08 2.63 E-08 1.88 E-08 7.91 E-16 4.99 E-16 1.74 E-14 1.89 E-13 2.54 E-14 4.54 E-14 7.78 E-07 3.03 E-13 4.48 E-08 1.60 E-07 1.83 E-14 9.64 E-10 1.93 F 08 4.88 E-07 2.31 E-14 9.68 E-07 4.42 E-14 4.27 E-14 2.63 E-14 9.65 E-14 4.28 E-15
Semiannual Radioactive January 1,
1989 Effluent Release Report to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 1
LONER LIMITS OF DETECTION (LLD) 2.
LLD's for Liquid Effluents NUCLIDE H-3 Be-7 Na-24 Ar-41 Cr-51 Mn-54 Fe-55 Co-57 Co-58 Fe-59 Co-60 Zn-65 Kr-85m Sr-89 Sr-90 Zr-95 Nb-95 Mo-99 Tc-99m Rh-105 RU-103 Ru-105 Sn-113 Sb-122 Sb-124 Sb-125 I -131 Xe-131m I -'133 Xe-133 Xe-133m Xe-135 Cs-134 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Hf-181 W -187 Gross Alpha LLD(uCif'ml) gg~
3.17 E-06 3.83 E-07 7.00 E-08 6.61 E-08 3.66 E-07 3.05 E-08 1.34 E-07 3.01 E-08 3.04 E-08 6.87 E-08 4.19 F 08 7.90 E-08 3.78 E-08 8.84 E-09
, 5.72 E-09 5.12 E-08 2.83 E-08 3.73 E-07 3.07 E-08 1.88 E-07 2.09 E-08 1.43 E-07 3.17 E-08 1.62 E-08 1.62 E-08 1.21 E-07 2.89 E-08 1.51 E-06 2.41 E-08 9.45 E-08 2.28 E-07 4.03 E-08 2.35 E-08 2.93 E-08 1.80 E-07 5.77 E-08 5.86 E-08 2.37 E-07 3.07 E-08 9.12 E-08 6.18 E-08
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 1A :
GASEOUS EFFLUENTS SUPINATION OF ALL RELEASES Units Quarter 1
Quarter 2
Est. Total Error Fission 6 Activation Gases A.
Total Release B.
Average Release Rate for Period C.
Percent of Technical Specification Limit Ci uCi/sec 4.93 E+02 2.99 E+02 4.50 E+01 6.34 E+01 3.80 E+01 1.40 E+00 1.24 E+00 2.
A.
Total Iodines B.
Average Release Rate for Period C.
Percent of Technical Specification Limit Ci uCi/sec 0.00 E+00 0.00 E+00 2.00 E+01 0.00 E+00 0.00 E+00 0.00 E+00 0.00 E+00
- 3. Particulates A.
Particulates with Tl/2> 8 days B.
Average Release Rate for Period C.
Percent of Technical Specification Limit Ci 0.00 E+00 1.81 E-07 2.00 E+01 uCi/sec 0.00 E+00 2.33 E-08 0.00 E+00 1.10 E-03 D.
Gross Alpha Radioactivity Ci 0.00 E+00 0.00 E+00
- 4. Tritium A.
Total Release B.
Average Release Rate for Period C.
Percent of Technical Specification Limit Ci 0 F 00 E+00 0 F 00 E+00 3.00 E+01 uCi/sec 0.00 E+00 0.00 E+00 0.00 E+00 0.00 E+00 2/3
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 1B GASEOUS EFFLUENTS ELEVATED RELEASES All releases at Shearon Harris are made as ground releases.
2/4
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 1C GASEOUS EFFLUENTS GROUND LEVEL RELEASES 1.
Fission and Activation Gases Continuous Mode
- Batch Mode**
Nuclides Units Quarter Quarter Quarter Quarter Released 1
2 1
2 H-3 Ar-41 Kr-85 Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-138 Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci
< LLD
< LLD
< LLD 1.06E+01 3.53E+00 1.76E+01
< LLD 4.23E+02 7.05E+00 2.47E+01
< LLD 3.53E+00
< LLD
< LLD
< LLD 6.46E+00 2.15E+00 1.08E+01
< LLD 2.58E+02 4.31E+00 1.51E+01
< LLD 2.15E+00
< LLD 4.4 E-03
< LLD 2.1 E-02 1.8 E-04 1.7 E-02
< LLD 2.8 E+00 5.8 E-02 4.3 E-01
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD LLD
< LLD Total Ci 4.90E+02 2.99E+02 3.33 E+00
< LLD Noble Gas quantities apportioned as per GALE code.
The quantities in Quarter 1 are from the unplanned release.
- routine batch gaseous releases were made.
No 2.
- Iodines, Nuclides Released Units Continuous Mode Quarter Quarter 1
2 Batch Mode Quarter Quarter 1
2 I-131 I-132 I-133 I-135 Ci Ci Ci Ci
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD Total Ci
< LLD
< LLD
< LLD LLD 2/5
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 1C (Continued):
GASEOUS EFFLUENTS GROUND LEVEL RELEASES 3.
Particulates Continuous Mode Batch Mode **
Nuclides Released Cr-51 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Sr-89 Sr-90 Nb-95 Mo-99 Cs-134 Cs-137 Ba/La-140 Ce-141 Ce-144 Gross Alpha Total Units Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Quarter 1
< LLD
< LLD
< LLD
< LLD LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD Quarter 2
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD 1.81 E-07
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD 1.81 E-07 Quarter 1
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD Quarter 2
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD LLD
< LLD
< LLD
< LLD
< LLD
< LLD
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 2A
- LIQUID EFFLUENTS SUMMATION OF ALL RELEASES Units 1.
Fission s Activation Products Quarter Quarter
'Est. Total 1
2 Error A.
B.
C.
Total Release (not including tritium,
- gases, or alpha)
Average Diluted Concentration during Period Percent of Technical Specification Limit Ci uCi/ml 3.58 E-02 4.67 E-02 3.50 E+01 1.30 E-08 1.09 E-08 5.77 E-02 3.31 E-02 2.
Tritium A.
B.
C.
Total Release Average Diluted Concentration during Period Percent of Technical Specification Limit Ci uCi/ml 7.24 E+01 1.50 E+02 3'0 E+01 2.63 E-05 3.49 E-05 8.75 E-01 1.16 E+00 3.
Dissolved and Entrained Gases A.
B.
C.
Total Release Average Diluted Concentration during Period Percent of Technical Specification Limit Ci 1.63 E-02 4.04 E-04 3.50 E+01 uCi/ml '.91 E-09 9.39 E-11 2.96 E-03 4.69 E-05 4.
Gross Alpha Radioactivity Total Release Ci
< LLD
< LLD 3'0 E+01 2/7
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 2A : LIQUID EFFLUENTS SUHMATION OF ALL RELEASES Units Quarter Quarter Est. Total 1
2 Error 5.
Volume of water released prior to dilution A. Batch Release B. Continuous Release liters liters 3.03 E+06 2.54 E+06 1.00 E+01 1.69 E+07 1.65 E+07 1.00 E+01 6.
Volume of dilution water liters used during period 2.76 E+09 4.30 E+09 1.00 E+01 2/8
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 2B
- LIQUID EFFLUENTS 1.
Fission and Activation Products Continuous Mode Batch Mode Nuclides Released H-3 Units Ci Quarter 1
< LLD Quarter 2
< LLD Quarter 1
Quarter 2
7.24 E+01 1.50 E+02 Na-24 Cr-51 Mn-54 Fe-55 Co-57 Co-58 Fe-S9 Co-60 Zn-65 Sr-89 Sr-90 Zr/Nb-95 Nb-97 Mo-99 Tc-99m Ru-103 Ru-105 Sn-113 Sb-124 Sb-125 I -131 I -133 Cs-134 Cs-137 Ba~-140 Ce-141 Ce-144 Hf-181 W -187 Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD 1.41 E-05 2.31 E-04 8.36 E-03 4.20 E-04 6.95 E-05 1.48 E-02 6.43 E-04 7.01 E-03
< LLD
< LLD
< LLD 3.85 E-04 1.42 E-05 4.06 E-OS 4.40 E-05
< LLD LLD 1.16 E-04 LLD 2.58 E-03 2.33 E-04 2.84 E-04
< LLD 4.68 E-06
< LLD 1.99 E-05
< LLD
< LLD 1.19 E-03 3.53 E-02 5.51 E-06 4.40 E-03 1.52 E-05 3.26 E-03
< LLD 1.55 E-06
< LLD 1.24 E-04 1.82 E-05 LLD 1.42 E-04
< LLD
< LLD 2.64 E-05
< LLD 1.46 E-03 2.86 E-04 2.25 E-04 4.86 E-05 1.51 E-04
< LLD
< LLD LLD
< LLD
< LLD Total Ci Gross Alpha Ci
< LLD
< LLD
< LLD 7.24 E+01 1.50 E+02 2/9
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 2
- Effluents Released Table 2B (Continued)
- LIQUID EFFLUENTS 2.
Dissolved and Entrained Gases Continuous Mode Batch Mode Nuclides Released Units Quarter 1
Quarter 2
Quarter 1
Quarter 2
Ar-41 Kr-85m Xe-131m Xe-133 Xe-133m Xe-135 Ci Ci Ci Ci Ci Ci
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD
< LLD 4.04 E-05 1.61 E-02 1.47 E-04 5.54 E-05
< LLD
< LLD
< LLD 3.42 E-04 LLD 6.14 E-05 Total Ci
< LLD
< LLD 1.63 E-02 4.03 E-04 2/10
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 3
- Solid Waste Disposal Table 3
- SOLID WASTE AND IRRADIATED FUEL SHIPMENTS 1.
Solid Waste Shipped for Burial or Disposal
( WASTE CLASS A
)
A.
Type of waste 6 month Units Period Est. Total Solid. Cont.
No.
Error (0)
Agent Type Form Ship a.
Spent Resin, filter m3 4.14 E+01
- sludge, evaporator bottoms, etc.
Ci 3.33 E+00 1.00 E+01 Cement STP S~
7 b.
Dry Compressible
- Waste, contaminated eguipment, etc.
m3 1.79 E+01 1.00 E+00 NA STP D
8 Ci 3.72 E-01 ***
c.
Irradiated Components, m3 No waste of this type shipped.
Control rods, etc.
Ci d.
Other (Describe) m3 No waste of this type shipped.
Ci STP Strong Tight Package n
S Solidified D Dewatered
- See page 2/12
i
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 3
- Solid Waste Disposal Table 3
- SOLID WASTE AND IRRADIATED FUEL SHIPMENTS (Continued)
B.
Estimate of Major Nuclide Composition (by type of waste)
Type of Waste a ~
b.
Nuclide H -3 C -14 Mn-54 Fe-55 Co-58 Co-60 Ni-63 Sb-125 H -3 14 ***
Cr-51 Mn-54 Fe 55 ***
Fe-59 Co-58 Co-60 63 ***
Nb-95 Zr-95 Cs-137 Percent Composition 1.36 E+01 2.60 E+00 1.37 E+01 3.09 E+01 1.59 E+Ol 1.25 E+01 1.07 E+01 1.00 E-01 4.93 E+00 5.00 E-02 4,60 E+00 1.17 E+01 3.00 E+01 2.20 E+00 3.27 E+01 6.00 E+00 2.30 E+00 3.70 E+00 1.80 E+00 2.00 E-02 Total Activity Ci 4.54 E-01 8.81 E-02 4.55 E-01 1.03 E+00 5.29 E-01 4.17 F 01 3.55 E-01 4.29 E-03 1.83 E-02 1.98 E-04 1.72 E-02 4.33 E-02 1.12 E-01 8.32 E-03 1.21 E-01 2.24 E-02 8.60 E-03 1.39 E-02 6.57 E-03 9.07 E-05 c.
No waste of this type shipped.
d.
No waste of this type shipped.
- These values are estimates only.
In May, SHNPP was informed by the vendor (SAIC) who performs our 10CFR61 sample analyses, that an error was made in the Dry Compressible Waste sample analysis.
SHNPP has received the amended analysis and has corrected all curie totals for waste shipped to Scientific Ecology Group (SEG) for processing.
SHNPP is currently awaiting the corrected totals from SEG for all SHNPP waste that has been shipped by SEG for burial.
The amended totals will be incorporated into the next Semiannual Report.
2/12
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 2 (Continued): Effluent and Waste Disposal Report Enclosure 3
- Solid Waste Disposal Table 3
- SOLID WASTE AND IRRADIATED FUEL SHIPI1ENTS (Continued)
C.
Solid Waste Disposal Number of Shipments Node of Transportation Destination 1.50 E+01 Truck Barnwell, S.C.
9 of these shipments (1 type 1.A.a and 8 type 1.A.b) were made from the Scientific Ecology Group (SEG) processing facility in Oak Ridge, Tennessee.
The other 6 shipments were made from the Harris site.
2.
Solid Waste Shipped for Burial or Disposal
( WASTE CLASS B
)
No waste of this type was shipped during this Report period.
3.
Solid Waste Shipped for Burial or Disposal
( WASTE CLASS C
)
No waste of this type was shipped during this Report period.
4.
Irradiated Fuel Shipments (Disposition)
No irradiated fuel was shipped during this Report period.
2/13
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 3
- Changes to Process Control Program (PCP)
Technical Specification 6.13 No changes were made to the PCP during this Report period.
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale TABLE OF CONTENTS i-v Title change.
New sub-section added.
Appendix letter change when Appendix C in Revision 2 was deleted.
New tables added.
SECTION
1.0 INTRODUCTION
A complete list of Technical Specifications (T.S.) which relate to the ODCM was added.
This includes T.S. which address effluent monitor alarm'trip setpoints, environmental monitoring program, land use census, interlaboratory comparison
- program, and routine reports.
1-2 A paragraph was added to specify that dose estimates to members of the public, which are included in the Semiannual Radioactive Effluent Release
- Report, are generated by using the NRC codes LADTAP II and GASPAR.
These codes use the
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Page Number Number Rationale parameters and methodology described in NUREG 0133 and Regulatory Guide 1.109.
SECTION 2.0 LIQUID EFFLUENTS
~
2-1 2-2 The introduction to the Section 2.0, Liquid Effluents, was re-written for clarity.
The relationship between liquid effluent radioactivity limits and the corresponding T.S.,
Paragraph 2.1.1 was re-written to provide the philosophy behind liquid monitor setpoints.
Revision 3 provides two setpoints; the high alarm and the alert alarm.
The high alarm is established to prevent exceeding T.S. limits for release to unrestricted areas whereas the alert is set at a fraction of the high alarm to provide a warning to the operator in advance of termination of the release.
2-3,4,5 Paragraph 2.1.1.2 introduces a different approach to determining the Minimum Acceptable Dilution Factor (D
) for batch releases.
The approach used in Revision 2 did not include the dilution required by non-gamma emitters in the calculation.
The new equation (2.1-1) calculates D
as a function of three (3) components:
- gamma, non-gamma, and tritium activities in relation to their respective HPC's.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale 10 2-5 The nominal average blowdown dilution flow rate was changed to reflect current operational data and the effect of a plant modification creating a cooling tower make-up water cross-tie to the discharge line.
2-6 Nominal values were assigned to the T factor used in Equation 2.1-5.
12 2-6 2-6,7 The concept of an Available Dilution Factor (D
- 1) is introduced.
This factor is basically the ratio of the dilution flow rate to the batch release rate.
The method for determining pre-release compliance with 10CFR20-based T.S. was simplified from Revision 2.
In Revision 3, the ratio of D 1/D is the criteria.
Since avl o
D is a function of the tank activities and represents the 0dilution required to ensure T.S. compliance, only a test of whether the estimated available dilution meets or exceeds the minimum requirement is necessary.
14 2-7,8,9 Revision 2 provided three (3) methods for determining liquid effluent monitor alarm'trip setpoints.
The first method (Equation 2.1-5) was actually a monitor response because no relationship was established between the setting and the 10CFR20-based limits.
The other two alternative setpoint methods were based on (1) the I-131 MPC, and on (2) the analyses of the batch prior to release.
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale In the case of the I-131 method (Equation 2.1-5a, Revision 2), the equation included the term MRR, the maximum effluent discharge rate calculated in Equation 2.1-3.
However, in many instances the MRR could have exceeded the actual tank discharge pump rate resulting in an incorrect setpoint.
The correct term should have been RR, the lower value between the maximum release rate and the pump capacity.
The setpoint method based on effluent analysis prior to release in Revision 2 (Equation 2.1-5c) used a term DFB, a conservative dilution factor, derived from DF
, the 0
minimum acceptable dilution factor.
- However, as noted previously, the dilution required by non-gamma emitters in the batch release had not been accounted for.
Furthermore, this method summed the uCi/ml values for all gamma
- nuclides, implying that all nuclides produce the same monitor response.
Revision 3 provides two (2) setpoint methodologies.
One is based on the known gamma activities in the batch and assumptions about the activities of the non-gamma components.
The second method is used when no detectable gamma emitters are present in the sample.
It assumes that any gamma activity is due entirely to I-131, the nuclide with the most restrictive MPC.
Revision 3 provides an improvement in the accuracy of the setpoints based on sample analysis by introducing the factor "Sens g" into equation (2.1-7).
"Sens g" is the energy-dependent gamma sensitivity of the monitor in 4/4
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation llanual (ODCIl)
Technical Specification 6.14 Change Page Number Number Rationale 2-9 cpay'uCi/ml.
- Thus, the relative nuclide composition is accounted for in determining the setpoint.
This setpoint equation (2.1-7) also uses a modified dilution factor.
Since the monitor is responsive to gamma emitters only, the dilution required by and available for the gamma emitters (and composite sample activities),
D and D
ogc influences the setpoint.
D 1
is derived by subtracting avlgc the minimum acceptable dilution for tritium, D t, from the ot'otal available dilution, D
1, while D represents the avl'gc difference between the minimum acceptable dilution factor for the release (D
) and that required by tritium (D t).
0 ot
'evision 3 provides an improvement over Revision 2 in setpoint methodology by including equations for determining alert alarm setpoints for liquid and gas monitors.
16 2-9,10 Revision 3 introduces a Check for Excessive Honitor Background for liquid monitors (Equation 2.1-9).
The purpose of this check is to ensure that monitor contamination is corrected before the ability to detect actual releases of radioactivity is masked.
17 2-10,11 The setpoint methodology based on I-131 IIPC (Equation 2.1-
- 10) uses the "Sens g" and modified dilution factor previously described in change number 14.
18 2-11 See change number 16 4/5
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale 19 2-13 Revision 2 provided two methods for establishing setpoints for continuous release monitors.
Method 1 (Equation 2.1-
- 10) was based on the MDC for an unspecified nuclide, required ambient gamma ray measurements, and did not add the background in the setpoint value.
Method 2 (Equation 2.1-11) defined background only in terms of monitor internal contamination and established the setpoint at twice background.
Revision 3 provides improved methodology for establishing the setpoints on monitors of continuous releases (Equation 2.1-15).
The basis of the method is the assumed FSAR nuclide composition of the Secondary Waste Sample Tank which has been extrapolated to apply to the Normal Service Water and all other continuous release sources.
The monitor high alarm setpoint is established at 10% of the MPC ff for the mixture.
eff 20 2-13,14 See change number 15 21 2-14 See change number 16 22 2-15 Equation numbers changed 23 2-15 Editorial change to indicate that annual summations of total body and organ doses are calculated by the ODQ1 software.
4/6
l*>
Ci
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale 24 2-17 A new value for the A factor is used as a result of the change in the nominal average blowdown flow rate (see change number 10).
25 2-17 Editorial changes to clarify the sources of the C.
and Fk values in the dose equation 2.2-1 for continuous releases.
26 2-18 Editorial changes for the sake of clarity.
27 2-19 2-20 Editorial addition to note that the dilution factor used in developing the A.t dose factors is a conservative value.
it Corrected to completely cite the T.S. requirement.
29 2-23 Table 2.1-1 changed to reflect operational conditions.
The TLsHS pump discharge limit will vary depending on whether the cross-tie to the floor drain system is open.
30 2-24 Table 2.1-2 added.
As a result of plant modifications, the dilution flow rate provided by the cooling tower blowdown and the cooling tower make-up water cross-tie may be varied.
2-24 Table 2.1-3 was added.
The signal processor time constants are needed in several equations (e.g., 2.1-9).
32 2-25 Table 2.1-4 was added.
The table presents the gamma energy-dependent sensitivities for the General Atomics (GA)
Cl
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCH)
Technical Specification 6.14 Change Page Number Number Rationale RD-53 monitors.
The weighted gamma emission rate was used to develop an effective energy for the principal nuclides in plant liquid effluents.
This value was used in conjunction with the GA calibration curve for the RD-53 to establish nuclide specific sensitivities (Sens g) in cpny'uCi/ml (see change number 14).
33 2-26,27 Corrected errors and omissions from Revision 2.
34 2-27 Added values for additional nuclides observed in liquid releases during the first two years of operation (i.e.,
Sn-113, Sb-124, Hf-181) 35 2-29 Figure 2.1-1 was updated to show the current configuration of the Liquid Waste Process Flow, particularly the use of the filter/demineralizer system to treat the floor drain and detergent drain tank effluents and the installation of by-pass lines from high and low conductivity storage tanks to the demineralizers.
36 2-30 Figure 2.1-2 was modified to add the note stating that, by procedure, radioactive liquids are not permitted to be sent to the Waste Neutralization Basin.
37 2-31 Figure 2.1-3 was modified to show the cooling tower make-up (CTMU) pumps and the CTMU bypass line.
4/8
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale 38 2-32 Figure 2.1-4 was modified to show radiation monitor REM-3528 in its actual location downstream of the sumps.
SECTION 3.0 GASEOUS EFFLUENTS 39 3-1I2I3 Revision 2 provided three methodologies for gas monitor setpoint determination.
However, the method based on a conservative nuclide mix (Equation 3.1-5) did not include a safety factor and could not handle the simultaneous batch and continuous ventilation flows which could occur in vent stacks 1 and 5.
The alternative method based on gas effluent analyses prior to release (Equation 3.1-10) assumed that the continuous ventilation flow used to dilute a batch release would be non-radioactive and also omitted a safety factor.
The third method (Equation 3.1-13) based on gas analysis and estimates of maximum acceptable flow rate for a batch release did not take into account the effect of dilution flow thereby providing a setpoint for a monitor at the tank outlet rather than the vent stack.
Revision 3 provides two (2) setpoint methods:
a general one and a conservative alternative.
The principal method (Equation 3.1-5} can be applied to continuous vent stack flows with or without concurrent batch releases and whether or not the composition is known or assumed.
The general method also includes a safety factor.
The 4/9
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale conservative alternative method assumes the gaseous effluent'is composed solely of Kr-89, the noble gas with the highest dose factor.
These changes provide more correct methodologies for setpoint determination and a higher degree of safety.
40 3-3,4 Vent stack flow rate units were changed from cfm to acfm and updated with current rating for each vent stack.
41 3-5,7 The factor T is now introduced earlier in the development of the setpoint than it was in Revision 2.
3-6 Revision 2 utilized meteorological data from 1976-1982.
Revision 3 has taken advantage of the availability of SHNPP meteorological data for 1976-1987.
Using the 12 year data base results in changes to the K/Q and D/Q values and to the sector in which they occur.
43 3-6 The T values for the vent stacks were revised based upon current flow ratings.
44 3-7,8 Maximum allowable noble gas release rates were calculated for each vent stack based on whole body and skin dose and included for reference.
45 3-8 In Equation 3.1.5, a safety factor of 2 is, incorporated into the high alarm setpoint for the gas channel (uCi/'cc).
4+0
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Hanual (ODCl<)
Technical Specification 6.14 Change Number Page Number Rationale 46 3-9 The RN-11 console monitors gaseous effluent releases in units of uCi/cc (the gas channel) and uCi/sec (the effluent channel).
Setpoints are needed for each channel.
Equation 3.1-6 provides the conversion from a gas channel setpoint to the effluent channel setpoint.
47 3-10 An alternative setpoint method is provided.
3-11 Table 3.1-1, Gaseous Source
- Terms, was modified to show the A. values in directly useful terms of uCi/cc and to add assumed nuclide releases through Waste Processing Building ventilation flow via WPB Vent Stack 5A.
49 3-15 Paragraph was re-written to indicate that compliance with T.S. 3/4 11.2.1 for iodine and particulates should be assessed for a child's thyroid dose as a result of inhalation rather than the infant exposure through the milk pathway.
50 3-16 Equation 3.2-3 was changed to eliminate all exposure pathways except inhalation.
51 3-16 Paragraph was re-written to indicate that the bases for T.S. 3/4 11.2.1 state that compliance with the 10CFR20-based annual thyroid dose rate limit should be evaluated for a child via the inhalation pathway at or beyond the site boundary.
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODQC)
Technical Specification 6.14 Change Number Page Number Rationale 52 3-18 Table 3.2-2 was updated to include the data from the 1988 land use census.
53 3-20 Table 3.2-4 was completely revised to present only the P.1 values for a child and the inhalation pathway.
In addition, values for Sn-113, Sb-124, and Hf-181 were included.
54 3-28 In Revision 3, compliance with 10CFR50 annual dose limits for radioiodines, particulates and tritium (Equation 3.3-
- 10) is evaluated for a hypothetical child at the exclusion boundary in the South sector.
This represents a change from Revision 2 which used an infant at the limiting real offsite location.
The change in receptor age and location for the test case provides a calculational model which would prevent a substantial underestimate of the actual exposure to a member of the public. It also avoids the possible need to revise the ODQ1 software each year if residential patterns shifted.
55 3-29 Describes the use of the GASPAR code to establish the critical sector and age group for evaluating 10CFR50 compliance for radioiodines, particulates and tritium.
56 3-51 Figure 3.1 has been revised to show ventilation from the condensate polisher area exhausted to Vent Stack 3A.
4/12
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale SECTION 4.0 RADIOLOGICAL ENVIRONI'KNTALMONITORING PROGRAM 57 4-2 Corrected SR number.
58 4-2 Corrected SR number and distance.
59 4-2 Corrected sector.
60 4-3 Corrected distance.
61 4-3 Air sampler location was shifted from the ENE sector to the SSW sector.
The 12 year meteorological data base (1976-1987) showed a higher D/Q in this sector.
62 4-9 TLD location added to provide coverage for the inner ring in the WSW sector.
63 4-13 Food crop sampling locations were changed.
The 12 year meteorological data base indicated higher D/Q values for these locations.
64 Note added to Table 4.1 to explain the changes in food crop sampling locations.
65 4-19 Figure 4.1-4 revised to show new sampling locations 47, 54, 55, and 56.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale 66 4-20 Legend for Figure 4.1-5 amended to show new sample locations and sample types.
SECTION 5 INTERLtQ30RATORY COMPARISON STUDIES 67 5-1 Sentence added to show the laboratory responsible for composite sample radiochemical analyses.
SECTION 6.0 TOTAL DOSE (COMPLIANCE WITH 40CFR190) 68 6-1,2 Section 6.0 has been re-written.
Revision 2 used a series of "calculational rules" to evaluate conformance with EPA annual dose limits.
These rules led to rough estimates of the dose to members of the public.
Revision 3 provides an improvement in the estimation of offsite doses by indicating that the Regulatory Guide 1.109 and NUREG 0133-based NRC codes LADTAP II and GASPAR should be used for this purpose.
69 6-2 T.S. 6.9.1.4 requires an assessment of the radiation dose to members of the public due to their activities inside the site boundary be performed in accordance with the methodology and parameters in the ODCM.
Revision 2 did not address this issue.
In Revision 3, Section 6.2 has 4/14
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Manual (ODCM)
Technical Specification 6.14 Change Page Number Number Rationale been added to to indicate that the methodology to be used for this assessment is the LADTAP II and GASPAR codes.
APPENDIX A METEOROLOGICAL DISPERSION FACTOR CALCULATIONS 70 A-1 Cites the version of the XOQDOQ computer code used to generate the meteorological tables in Appendix A.
71 A-2 Revised to note that a 12 year record (1976-1987) was used as the historical data base in Revision 3.
Reference to data collected at the 61.42 meter tower height was removed because plant releases are assumed to be ground level.
72 A-4 Tables A-1 through A-4 were revised to provide current X/Q and D/Q for special locations identified in the 1988 land use census.
These values came from Version 2.0 of XOQDOQ using a 12 year data base.
73 A-5 Tables A-5 through A-13 were revised based on 12 years of through data and Version 2.0 of XOQDOQ.
A-15 A-16 Table A-14 was revised to show the input parameters used in through Version 2.0 of XOQDOQ.
A-21 4/15
'i4
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 4 (Continued)
- Changes to the Off-site Dose Calculation Nanual (ODCM)
Technical Specification 6.14 Change Number Page Number Rationale APPENDIX B DOSE PARMETERS FOR RADIOIODINESg PARTICULATESg AND TRITIUN 75 B-1 NUREG-0133 defines P.
as "the dose parameters for 1
radionuclides other than noble gases for the inhalation, food and ground plane pathways.
The dose factors are based on the critical individual organ and the most restrictive age group (infant)".
The P. values are used in Equation 1
3.2-3 to test compliance with 10CFR20 limits.
- However, T.S.
3/'4 11.2.1 bases state that the limit applies to the child's thyroid dose rate as a result of inhalation of radioiodines, particulates and tritium.
76 B-2 Discussion of the child as the critical receptor and the general equation for calculating P.
for a child is iI presented.
77 B-8 Corrected table numbers.
78 B-11 Corrected subscript.
79 B-14 Added brackets to the equation.
80 B-15 Corrected parameter letter.
81 B-15 Corrected units.
4/16
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 5
- Changes to the Environmental Monitoring Program Enclosure 1
- Environmental Monitoring Program Technical Specifications 3.11.2.3 3.12.1 3.12.1.c Three changes were implemented in the Environmental Monitoring Program during the Report period.
1.
TLD location I56 was added to provide inner-ring coverage in the WSW sector.
This coverage had not previously been provided because inner-ring locations are normally near the site boundary.
The nearest point of access in the WSW sector is 2.8 miles from the site boundary.
- However, in 1989, it was decided to install a TLD at 2.8 miles from the site boundary to provide an approximate inner-ring complement to the TLD located at 5.6 miles in this sector.
2.
A review of SHNPP historical meteorology data for the period 1976-1987 indicated that at the site boundary, the D/Q value was higher in the SSW sector than the D/Q value in the ENE sector.
As a result, air sample location I3 at the Harris Environmental and Energy Center (ENE sector) was discontinued and the equipment shifted to create a new air sample location I47 in the SSW sector.
Furthermore, relocation of the air sampler provides air sampling at a location closer to the real critical Member of the Public of SHNPP, i.e.,
a child in the SW sector.
3.
Appropriate revisions to ODCM Table 4.1 and Figures 4.1-4 and 4.1-5 have been included in Revision 3 to the ODCM.
tE
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 5
- Changes to the Environmental Monitoring Program Enclosure 2
- Land Use Census Technical Specifications 3.12.2.a 3.12.2.b A Land Use Census was performed in May of 1989.
A number of changes have occurred in the locations of the nearest
- resident, garden, milk animal, and meat animal. Table 1 summarizes these locations.
Table 2 lists the kinds of meat animal at each meat animal location.
Table 1
Distance to the Nearest Special Locations for the Harris Nuclear Project (Miles)*
(Comparison of 1988/1989 Data)
ENE E
WNW NW NNW 1.32 1.33 1.33 1.33 1.33 1.33 1.33 1.33 1.36 1.33 1.33 1.33 1.33 1.33 1.26 1.26 2.2 2.2 1.8 1.8 2.3 2.3 3.6 2.0 1.9 1.9 2.7 2.7 4.3 4.3 4.4 4.4 3.9 3.9 2.8 2.8 4.3 4.3 2.7 2.8 2.1 2.1 1.8 1.8 1.5 1.5 Exclusion Residence Sector Boundary 1988 1989 Milk Animal 1988 1989 2.2 2.2 4.6 4.6 Garden 1988 1989 2.2 2.2 1.7 1.7 2.3 2.3 2
0 1.9 1.9 2.7 4.4 4.3 4.3 3.9 3.9 2.8 2.8 4.3 4.3 3.0 2.9 2.1 2.1 1.8 1.8 1.7 1.7 2.2 1.8 2.3 1.9 4.3 2.8 3.6 2.3 2.2 2.9 4.3 2.8 4.3 3.1 2.5 1.8 1.7 2.8 4.3 3.1 2.5 1.8 1.7 liat Animal 1988 1989
~ Distance estimates are + 0.1 miles except at the exclusion boundary.
5/2
1~
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 5
- Changes to the Environmental Ilonitoring Program Enclosure 2
- Land Use Census Technical Specifications 3.12.2.a 3.12.2.b Table 2
Neat Animal Type at Nearest Location to SHNPP by Sector Sector N
NNE NE ENE E
WNW NW NNW Distance (miles) 2.8 3.6 2.3 2.2 2.9 4.3 2.8 4.3 3.1 2.5 1.8 1.7 Heat Type(s)
Beef Chickens Beef/Chickens Chickens/Hogs Beef Chickens/Hogs Chickens Hogs Rabbits Chickens Beef Beef/Chickens/Hogs Owner
- Lawrence, H.
- Kudyba, P
James Rest Home
- Harris, H.
- Ragan, E.
- Taylor, M.
Pollard, J.
Smith P.
Allen, S.
Williams, H.
Fish, J.
- Godwin, W.
5/3
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 6
- Additional Technical Specification Responsibilities Enclosure 1
- Inoperability of Liquid Effluent Monitors Technical Specification 3.3.3.10, Action b Monitors Out-of-Service ) 30 Days During the Report Period Effluent Monitor Days Inop.
Reason REM-01MD-3528 139 Turbine Building Drains Electronic noise problems during pump starts.
Modification has been initiated to correct problem.
REM-01MD-3530 Tank Area Drains 129 Due to a design deficiency, the high alarm signal trips the process flow pump but does not isolate the release.
Modification has been initiated to correct problem.
REM-1WL 3540 Treated Laundry and Hot Shower Tank 33, Leak in monitor piping and performance of primary calibration.
REM-21WL-3541 Waste Monitor Tank 34 Modification to Cooling Tower Blowdown System completed and ODCM Rev.
2 changes allowed a more flexible alarm setpoint.
REM-1WS-3542 Secondary Waste Sample Tank 181 Monitor is operating;
- however, an incorrect ODCM calculation for the continuous release setpoint results in the monitor going into alarm.
The implementation of ODCM Rev.
3 will resolve this.
FT-1968 AGB Cooling Tower Make-up Bypass Line Flow Rate Monitors 181 Flow monitors A & B are not within the required tolerances, resulting in inaccurate flow measurements.
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 6 (Continued): Additional Technical Specification Responsibilities Enclosure 2
- Inoperability of Gaseous Effluent Monitors Technical Specification 3.3.3.11, Action a Monitors Out-of-Service
> 30 Days During the Report Period Effluent Monitor Days Inop.
Reason PNL-21AV-3509-SA PNL-21AV-3509-1SA Plant Vent Stack Flow Rate Monitor 181 Problems with calibration of flow measurement system resulting in discrepancies between actual and expected flow rates.
Modification initiated to correct.
RM-01TV-3536-1 TB Vent Stack 3A Flow Rate Monitor 181 Moisture interferences with the flow measurement system resulting in discrepancies between actual and expected flow rates.
Modification initiated to correct.
REM-01WL 3546 WPB Vent Stack 5
Noble Gas Monitor 35 Firmware failure in monitor resulted in check-source failures.
New firmware installed.
PNr 1wv-3546 a
PNL-1WV-3546-1 WPB Vent Stack 5
Flow Rate Monitor 181 Problems with calibration of flow measurement system results in discrepancies between actual and expected flow rates.
Modification initiated to correct.
PNL-1WV-3547 6 PNL-1WV-3547-1 WPB Vent Stack 5A Flow Rate Monitor 181 Problems with calibration of flow measurement system resulting in discrepancies between actual and expected flow rates.
Modification initiated to correct.
- ~ % i1-(
n
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 6 (Continued): Additional Technical Specification Responsibilities Enclosure 2
- Inoperability of Gaseous Effluent Monitors Technical Specification 3.3.3.11, Action a Monitors Out-of-Service ) 30 Days During the Report Period Effluent
'onitor Days Inop.
Reason OAI-21WG-1101 Waste Gas Compressor Discharge Oxygen Monitor 180 Monitor does not agree with analyzed samples.
HAIC-21WG-1118B Waste Gas Recombiner nBn Outlet Gas Hydrogen Monitor 181 Modification to improve reliability of monitor initiated.
OARC-1119 B
Waste Gas Recombiner "B" Outlet Gas Oxygen Monitor 181 Monitor filled with water. Monitor was cleaned but since it was not needed at the time, it was not returned to service.
6//3
.I,T 1 I
Semiannual Radioactive Effluent Release Report January 1',
1989 to June 30, 1989 Appendix 6 (Continued): Additional Technical Specification Responsibilities Enclosure 3
- Unprotected Outdoor Tanks Exceeding Limits Technical Specification 3.11.1.4, Action a No unprotected outdoor tank exceeded the Technical Specification limit during this Report period.
6/4
'iP'~
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 6 (Continued): Additional Technical Specification Responsibilities Enclosure 4
- Gas Storage Tanks Exceeding Limits Technical Specification 3.11.2.6, Action a No gas storage tank exceed the Technical Specification limit during this Report period.
'f if
Semiannual Radioactive Effluent Release Report January 1,
1989 to June 30, 1989 Appendix 7
- Major Modifications to Radwaste System Technical Specification 6.15.1 RAONt'ATE DEMINERALIZATIONPROCESS SYSTEM In January 1989 the Liquid Radwaste System was modified to process liquid radwaste through a vendor demineralization system.
The system installed was Chem-Nuclear's state-of-the-art Modular Fluidized Transfer Demineralization System.
SUMMARY
OF 10 CFR 50.59 REVIEW PCR-3547 provides the necessary document changes and evaluations to permit the installation and operation of the vendor demineralizer system which can process liquid waste streams for release to the environment.
This system is being utilized to reduce solid waste volumes and cost.
All Equipment utilized meets.
Branch Technical Position ESTB 11-1, Rev.
1 and is located in the Waste Processing Building.
The Liquid Waste Processing System (LWPS) is a FSAR Chapter 15 initiating system.
Demineralization is already employed in the LWPS as a polisher for other treatment methods; therefore, it is not a new process.
The system is being installed as an alternate means of processing liquid wastes.
The existing LWPS is being maintained in an operable condition to meet all FSAR and Technical Specification design requirements.
If effluent monitoring/sampling indicates that 10CFR50 Appendix I limits or 10CFR20 limits may be exceeded, the existing processing system will be utilized to process the waste prior to release.
Therefore, the demineralization Skid will not affect the ODCM.
The effluent from the demineralization system will be routed through the normal release paths.
Therefore, no new release paths are created.
The vendor equipment is required to meet branch technical position ESTB 11-1, Rev.
1 which meets the regulatory and design requirements of existing LWPS equipment.
The new components are not located near any Safety-Related equipment and the floor loading has been analyzed; therefore, there is no impact on Safety-Related equipment.
The Radwaste demineralizer skid is located in the same area as existing process equipment.
The size of the vendor demineralizer vessels is bounded by existing LWPS demineralizers and any liquid released due to the failure of the new components will be contained within the Waste Processing Building and routed back to the LWPS by the existing floor drain system.
These scenarios are covered by the scope of the analysis provided in Section 15.7.2 of the FSAR.
Therefore, the addition of the vendor demineralization system will not increase 7/1
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Semiannual Radioactive Effluent Release Report January 1, 19S9 to June 30, 1989 Appendix 7 (Continued):
Major Modifications to the Radwaste System Technical Specification 6.15.1 the probability or consequences of an accident or malfunction of equipment to important to safety previously analyzed; create the possibility of an accident or malfunction of equipment important to safety not previously analyzed; or impact the margin of safety as defined in any Technical Specification bases.
REASON FOR THE CHANGE The objective of the Radwaste Demineralization System (RDS) is to reduce the overall operational cost of processing liquid radwaste and to reduce the volume of Solid Radwaste while maintaining the activity in the liquid released to the environment at a small fraction of Technical Specification limits.
If operation of the RDS proves effective and reliable, the existing systems will be mothballed/removed.
The RDS is considerably cheaper to operate and maintain.
DESCRIPTION OF RADNASTE DEMINERALIZATIONSYSTEM The Radwaste Demineralization System consists of the following modules:
(2) Vessel Modules Booster Pump Module Control Module Sample Sink Module (2) Bag Filters (2) Cartridge Filters Vessel Module Each module consists of two 30 cubic feet demineralizers designed and fabricated in accordance with Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code.
The vessels are 36" in diameter with a maximum bed size of 22.5 cubic feet and are 304 Stainless Steel with stainless steel internals.
The vessels can be operated in parallel, series or a combination of the two. The vessels are connected by the use of hoses.
Booster Pump Module The pump module consists of a centrifugal pump, relief valve, block valves and interconnecting piping.
The module is only connected when required to provide design flow rates.
Control Module The control module consists of the valves and the interconnecting piping to control the flow to the vessel modules and provide the air and flush water interfaces.
The Skid also includes a relief valve to protect the system and a flow totalizer.
The control skid is connected at all times upstream of the'vessel modules.
The Control Module is connected to the existing system by two hoses, a module influent and effluent hose.
le 1+1>>
,'v 7
Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 7 (Continued): Major Modifications to the Radwaste System Technical Specification 6.15.1 Sample Sink A stainless steel sample sink is provided to sample system influent, effluent from each bed, and the system final effluent.
These samples are used to determine the efficiency of each bed and the system and to determine when a bed is exhausted and needs to be sluiced out of the vessel.
Bag Filter Modules One bag filter module upstream of the vessel modules and one bag outlet of the system to capture any solids utilize bags with micron ratings between 5
modules permit monitoring of the inlet and filters.
is provided to collect solids filter module is provided on the leaving the system.
The filters and 50 microns.
The piping on the outlet pressures and bypassing the Cartridge Modules Two cartridge filter modules are provided to collect solids upstream, of the vessel modules.
The cartridge filters utilize six 30 inch cartridges.
Cartridges with micron ratings of 0.1 to 30 can be utilized.
The piping on these modules also provide the means to bypass the filter and monitor inlet and outlet pressures.
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Overall System The modules can be connected in a variety of ways using the hoses provided with the system.
Attachment 1 depicts a typical setup. All equipment including the interconnecting piping and hoses are designed for the maximum operating pressure of 150 psig and were hydrostatically tested to 225 psig for 30 minutes.
INTERFACE WITH LWPS E
The RDS can take its feed from the following Liquid Waste Processing subsystems:
Equipment Drains System Floor Drains System Laundry and Hot Showers System The effluent from the RDS can be sent either to the Waste Monitor Tanks or the Treated Laundry and Hot Shower Tanks for release through existing monitored release paths.
Tie-ins were also made to the sluice water system and the spent resin system to dispose of the media as it is expended.
The relief valves on the booster pump module and control module are piped to the floor drains for reprocessing.
Attachment 2 shows the main tie-ins to the LWPS.
7/3
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Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 7 (Continued): Major Modifications to the Radwaste System Technical Specification 6.15.1 EVALUATION OF SOLID, LIQUID AND GASEOUS EFFLUENT PREDICTIONS Solid Radwaste The concentrates produced by the Waste Evaporators and the Reverse Osmosis Evaporators and the resin from the existing Radwaste Demineralizers would be eliminated by operation of the Radwaste Demineralization System.
The projected
.volume of ion exchange media and filters would be 1000 cubic feet based on processing four million gallons of liquid by the RDS.
This is in lieu of 2215 cubic feet predicted in the FSAR based on 1.2 million gallons of liquid processed through the existing system.
The spent media from the RDS is dewatered and shipped to the burial site in accordance with existing regulations.
This is the same method of disposal used for other spent ion exchange media including Fuel Pool Demineralizer and CVCS Demineralizers.
Liquid Radwaste The volume of liquid radwaste processed is expected to decrease due to the shutdown of the Radwaste evaporators.
The predicted concentration of liquid radwaste is not expected to exceed the concentrations previously predicted in the FSAR.
Releases are made on a batch basis; isotopic concentrations are analyzed and dose to the public is calculated prior to any release in accordance with Technical Specifications.
As a backup the existing equipment is being maintained in an operable condition and can be utilized in case of failed fuel.
Gaseous Radwaste Operation of the RDS will not increase the release of gaseous effluent since no sources are created by operation of the system.
EXPECTED AND ACTUAL RELEASES Solid Radwaste During the period of January 1989 through June 1989 expected and actual waste generated is given below and compared with solid waste generated July 1988 through December 1988.
7/4
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Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 7 (Continued):
Major Modifications to the Radwaste System Technical Specification 6.15.1 Source Spent Resin Evaporator Bottoms Filter Particulates Form Dewatered Solidified Solidified Expected cu. ft.
500 415 80 995 Generated 162 QTR 1989 cu. ft.
368 2,074*
80 2,522*
Generated 3&4 qtr 1988 cu. ft.
1,031 1.237 0
2,268, The increase in volume of solid radwaste generated in 162 Quarter of 1989 was not due to operation of the RDS system but was due to an increase in generation of evaporator bottoms during the later part of 1988 and early 1989. This increase in generation rate was due to the addition of anti-foam solution to the RO Concentrates evaporator to reduce'd carryover.
The actual generation of solid radwaste from the RDS system (260 cu.ft. for 1,073,000 gallons processed) was just above the rate predicted of 250 cu.ft. per 1,000,000 gallons processed.
Liquid Radwaste The actual and expected volume of liquid radwaste generated during the period of January 1989 through June 1989 are given below and are compared to the those generated during July 1988 through December 1988.
Expected gallons 2,000,000 162 QTR 1989 gallons 1,443,387 3a4 QTR 1988 gallons 3,888,856 Reduction in volume of liquid radwaste generated is due to fixing of leaking valves during the refueling outage in later part of 1988 and the installation of the'Radwaste Demineralization System.
The total curies of Fission and Activation Products and the dose to the public for the period January 1989 through June 1989 is given below and compared to the period of July 1988 through December 1988.
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Semiannual Radioactive Effluent Release Report January 1, 1989 to June 30, 1989 Appendix 7 (Continued): Major Modifications to the Radwaste System Technical Specification 6.15.1.
- Curies released 1S(2 QTR 1989 8.08E-2 curies 3@4 gm 1988 6.18E-2 curies'ose to Public Total Body Bone Liver Thyroid Kidney Lung
- GI-LLI 2.3E-3 mrem 4.7E-4 mrem 3.0E-3 mrem 2.2E-3 mrem 2.1E-3 mrem 1.8E-3 mrem 1.3E-2 mrem 5.3E-3 mrem 1.8E-3 mrem 6.0E-3 mrem 3.6E-3 mrem 3.7E-3 mrem 3.4E-3 mrem 1.5E-2 mrem
- Not including Tritium
- Dose to GI-LLI is greater than the dose to any other organ Total curies of Fission and Activation Products released from the site during the 1st and 2nd quarter of 1989 has slightly increased over the curies release during the previous Report period.
Dose to the public has decreased due to increased'dilution water flow from the Cooling Tower Blowdown System. All releases were with the limits of 10CFR20, 10CFR50 Appendix I, and SHNPP Technical Specification and are less than those predicted in Table 11.2.3-4 of the FSAR.
EXPOSURE TO PLANT OPERATING PERSONNEL It is estimated that exposure to plant operating personnel as a result of the installation of the RDS will decrease overall. It is expected that the dose to the Radwaste personnel who operate the system will increase due to manual filter change-out but this will be offset by lower maintenance personnel exposure.
SAFETY AND TECHNICAL REVIEWS Documented safety and technical reviews in accordance with Technical Specification 6.5.1,have been completed for this modification and vendor procedures for operation of the system.
The Plant Nuclear Safety Committee reviewed and approved this modification in January 1989.
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