ML20107M471
| ML20107M471 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/1995 |
| From: | Stafford A VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 96-186, NUDOCS 9604300412 | |
| Download: ML20107M471 (202) | |
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i Vruo1NIA I?LECTHIC AND POWER COMPANY Ricimoxo,VinorxtA 20201 April 22, 1996 l I
l United States Nuclear Regulatory Commission Serial No.96-186 I Attention: Document Control Desk NL/RPC i Washington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF 4 l NPF-7 Gentlemen:
i VIRGINIA ELECTRIC AND POWER COMPANY l NORTH ANNA POWER STATION UNITS 1 AND 2 l ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT !
Pursuant to Technical Specification 6.9.1.9, enclosed is the Annual Radioactive Effluent Release Report for North Anna Power Station Units 1 and 2 for the reporting period of January 1,1995 to December 31,1995.
If you have any questions or require additional information, please contact us.
Very truly yours, M. L. Bowling, Manager Nuclear Licensing and Operations Support Attachment i cc: U. S. Nuclear Regulatory Commission i Region ll :
101 Marietta Street, N. W. ;
Suite 2900 '
Atlanta, Georgia 30323 Mr. R. D. McWhorter NRC Senior Resident inspector l North Anna Power Station !
300.124 9604300412 951231 8 TfW DR ADOCK 0500 ll
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ATTACHMENT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NORTH ANNA POWER STATION UNITS 1 AND 2 LICENSE NOS. NPF-4 AND NPF-7
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AHNilAL RADIOACTIVE EFFLUENT RELEASE REPORT NORTH ANNA POWER STATION l
(JANUARY 01, 1995 TO DECEMBER 31, 1995) l 1
I PREPARED BYA -
Supervisor Radiological Analysis q
REVIEWED BY: *
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Supervisor Technical Services APPROVED BY:
Superi ndent Radiological Protection 1
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This report is submitted as required by Appendix A to Operating License Nos. NPF-4 and NPF-7, Technical Specifications for North Anna Power Station, Units 1 and.2,-Virginia Electric and Power Company, Docket Nos. 50-338, 50-339, Section 6.9.1.9.
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ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT FOR THE NORTH ANNA POWER STATION JANUARY 01, 1995 TO DECEMBER 31, 1995 INDEX Section No. Subiect Pace 1 EXECUTIVE
SUMMARY
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2 2 PURPOSE AND SCOPE..................... 2 -
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1 3 DISCUSSION............................ 4 -
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I l 4 SUPPLEMENTAL INFORMATION.............. 6 l l
Attachment 1 -
Effluent Release Data................. 7 !
Attachment 2 -
Annual and Quarterly Doses............ 8 1
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Attachment 3 -
Revisions to Offsite Dose Calculation Manual (ODCM)......................... 9 I
Attachment 4 -
Major Changes to Radioactive Liquid, l Gaseous, and Solid Waste Treatment Systems............................... 10 l Attachment 5 -
Inoperability of Radioactive Liquid l and Gaseous Effluent Monitoring l
Instrumentation....................... 11 l
l Attachment 6 -
Unplanned Releases.................... 12 Attachment 7 -
Lower Limits of Detection (LLD) for Effluent Sample Analysis.............. 13 - 14 I
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1.0 EXECUTIVE SUSOEARY l The Annual Radiological Effluent Release Report describes the
! radiological effluent control program conducted at the North Anna Power station during the 1995 calendar year. This document summarizes the quantities of radioactive liquid and gaseous effluents and solid waste released from the North Anna Power Station in accordance with R.G. 1.21 during the period January 1 through December 31, 1995, and includes an assessment of radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents.
i There were no unplanned liquid or gaseous effluent releases classified according to the criteria in the Offsite Dose Calculation Manual during this reporting period.
l Based on the 1995 effluent release data, 10 CFR 50, Appendix I dose l
calculations were performed in accordance with the Offsite Dose
, Calculation Manual. The results of these pathway dose ca?.culaLivus l
indicate the following:
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- 1. The total body dose due to liquid effluents was 2.81E-01 mrem, l l
l which is 4.68% of the dose limit and the critical organ (liver) l dose due to liquid effluents was 2.89E-01 mrem, which is 1.4% of the dose limit.
- 2. The air dose due to noble gases was 2.29E-02 mrad gamma, which is j 0.1% of the annual gamma dose limit, and 1.07E-02 mrad beta, which i
is .03% of the annual beta dose limit.
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1.O EXECUTIVE
SUMMARY
(cont.)
with half-lives greater than 8 days was 2.18E-02 mrem, which is l
.07% of the annual dose limit.
There war' (1) one major change to radioactive liquid, gaseous, and solid j waste treatment systems during this reporting period. This was installation of a high capacity blowdown capability for each unit's steam generators. A brief description of this system and a summary of the safety evaluations are provided in Attachment 4.
There were changes to the Offsite Dose Calculation Manual, VPAP-2103, during this reporting period which were implemented in PN-1 and Revision 7 on January 3, 1995 and October 31, 1995 respectively. Attachment 3 provides the changes to VPAP-2103.
l Based on the levels of radioactivity observed during this reporting
- period and the dose calculations performed, the operations of the North Anna Nuclear Power Station Units 1 and 2 have resulted in negligible
- dose consequences to the maximum exposed member of the public in l
unrestricted areas.
2.O PURPOSE AND SCOPE i
The Radioactive Effluent Release Report includes, in Attachment 1, a summary of the quantities of radioactive liquid and gaseous effluents and solid waste as outlined in Regulatory Guide 1.21, " Measuring, l
l Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of 4
Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", Revision 1, June 1974, with data 2
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2.0 PURPOSE AND SCOPE (cont).
, summarized on a quarterly basis following the format of Tables 1, 2 and l
3 of Appendix B thereof. The report submitted before May 1st of each year includes an assessment of radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents released from the site during the previous calendar year. The report also includes a list of unplanned releases during the reporting period, in Attachment 6. !
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i As required by Technical Specification 6.15, changes to the Offsite Dose Calculation Manual (ODCM) for the time period covered by this report are l
included in Attachment 3. ]
l l Major changes to radioactive liquid, gaseous and solid waste treatment I systems are reported in Attachment 4, as required by the ODCM, section 6.7.2.a.4. Information to support the reason (s) for the change (s) and a summary of the 10 CFR 50.59 evaluation are included. In lieu ' of reporting major changes in this report, major changes to the radioactive waste treatment systems may be submitted as part of the annual FSAR update.
As required by the ODCM, sections 6.2.2.b.2 and 6.3.2.b.3, a list and i t
! explanation for the inoperability of radioactive liquid and/or gaseous
( effluent monitoring instrumentation is provided in Attachment 5 of this
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l report.
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l l 3.0 DISCUSSION l
l The basis for the calculatibe of the percent of technical specification I for the critical organ in Table 1A of Attachment 1 is the ODCM, section l 6.3.1, which requires that the dose rate for iodine-131 & iodine-133, for tritium, and for all radionuclides in particulate form with half-l lives greater than 8 days shall be less than or equal to 1500 mrem /yr to the critical organ at or beyond the site boundary. The critical organ is the child's thyroid via the inhalation pathway.
The basis for the calculation of percent of technical specification for the total body and skin in Table 1A of Attachment 1 is the ODCM, section 6.3.1, which requires that the dose rate for noble gases to areas at or beyond site boundary shall be less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.
l j The basis for the calculation of the percent of technical specification l
in Table 2A in Attachment 1 is the ODCM, section 6.2.1, which states that the concentrations of radioactive material released in liquid effluents to unrestricted areas shall be limited to 10 times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-4 microcuries/ml.
Percent of technical specification calculations are based on the total gaseous or liquid effluents released for that respective quarter.
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i l 3.0 DISCUSSION (cont.) l l
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I The annual and quarterly doses, as reported in Attachment 2, were l 1
l calculated according to the methodology presented in the ODCM. The beta and gamma air doses due to noble gases released from the site were calculated at site' boundary. The maximum exposed member of the public
! from the releases of airborne iodine-131 & iodine-133, tritium and all l
l radionuclides in particulate form with half-lives greater than 8 days, is defined as an infant, exposed through the grass-cow-milk pathway, I
1 with the critical organ being the thyroid gland. The maximum exposed member of the public from radioactive materials in liquid effluents in unrestricted areas is defined as an rdult, exposed by either the invertebrate or. fish pathway, with the criuical organ being the liver.
The total body dose was also determined for this individual.
Presented in Attachment 6 is a list of unplanned gaseous and liquid releases meeting the requirements of 6.7.2.a.3 of the ODCM.
The typical Lower Limit of Detection (LLD) capabilities of the radioactive effluent analysis instrumentation are presented in-Attachment 7. These LLD values are based upon conservative conditions (i.e., minimum sample volume and maximum delay time prior to analysis).
Actual LLD values may be lower. If a radioisotope was not detected when effluent samples were analyzed, then the activity of that radioisotope was reported as Not Detectable (N/D) on Attachment 1 of this report. If i an analysis for an isotope was not performed, then the activity was reported as Not Applicable 'A) .
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4.0 SUPPLEMENTAL INFORMATION As required by the ODCM, section 6.6.2, evaluation of the Land Use Census is made to determine if new location (s) have been identified for the radiological environmental monitoring program pursuant to the ODCM, Section 6. 6.2 requirements . Evaluation of the Land Use Census conducted in 1995 identified no change in sample locations for the radiological environmental monitoring program.
Changes to some TLD locations were made to resolve differences in emergency plan sectors and radiological environmental monitoring sectors. These changes were implemented with PN-1 to VPAP-2101, Off site Dose Calculation Manual, included in Attachment 3 of this report.
Section 6.6.1.b.4 of the ODCM requires identification of the cause(s) l for the unavailability of milk or leafy vegetation samples, and the identification of new locations for obtaining replacement samples. Milk l
, samples, as required by the ODCM, section 6.6.1, were available during l 1
the time period covered by this report. The leafy vegetation samples l
for vegetation station 14, 15, 16, 21 and 23 were not collected for the months of January, February, March, November and December 1995 due to seasonal unavailability. All other samples were obtained and analyzed l
as required during the time period covered by this report.
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l ATTACBENT 1 EFFLUENT RELEASE DATA (01/95 -
12/95)
This attachment includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste, as outlined in Regulatory Guide 1.21, 1
Appendix B.
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' TABLE 1A NORTH ANNA POWER STATION ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SUMMATION OF ALL GASEOUS EFFLUENT RELEASES FOR (01/95 - 12/95)
Page 1 of 2 IST 2ND ESTIMATED TOTAL UNITS OUARTER OUARTER PERCENT ERROR (%)
A. Fission and Activation Gases:
- 1. Total Release. Curies 2.53E+1 3.49E+0 1.80E+1
- 2. Averace Release Rate for Period uCi/sec 3.25E+0 4.44E-1
- 3. Iodines:
- 1. Total Iodine-131 Release. Curies 7.85E-5 1.56E-4 2.80E+1
- 2. Averace Release Rate for Period uCi/sec 1.01E-5 1.99E-5 C. Particulates (TM > 8 days):
- 1. Total Particulate (TM > 8 days)
Release Curies 2.17E-6 6.35E-5 2.80E+1
- 2. Average Release Rate for Period uCi/sec 2.79E-7 8.07E-6
- 3. Gross Aloha Radioactivity Release Curies 8.52E-6 5.74E-6 D. Tritium:
- 1. Total Release Curies 1.14E+1 1.08E+2 3.10E+1
- 2. Average Release Rate for Period uCi/sec 1.47E+0 1.38E+1 E. Percentage of Technical Specification Limits
- 1. Total Body Dose Rate % 9.28E-3 1.98E-4
- 2. Skin Dose Rate % 2.32E-3 7.59E-5
- 3. Critical Organ Dose Rate % 1.02E-3 9.79E-3
TABLE 1A NORTH ANNA POWER STATION ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SUMMATION OF ALL GASEOUS EFFLUENT RELEASES FOR (01/95 - 12/95)
Page 2 of 2 3rd 4th ESTIMATED TOTAL UNITS OUARTER OUARTER PERCENT ERROR (%)
A. Fission and Activation Gases:
- 1. Total Release. Curies 2.24E+0 5.39E+0 1.80E+1 j 2. Average Release Rate for Period uCi/sec 2.82E-1 6.78E-1 i
B. Iodines:
- 1. Total Iodine-131 Release. Curies 8.18E-9 N/D 2.80E+1
- 2. Averace Release Rate for Period uCi/sec 1.03E-9 N/D __
l C. Particulates (TM > 8 days):
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- 1. Total Particulate (TM > 8 days) i Release Curies 1.02E-5 1.66E-7 2.80E+1
- 2. Average Release Rate for Period uCi/sec 1.29E-6 2.09E-8
- 3. Gross Alpha Radioactivity Release Curies 4.62E-6 3.48E-6 D. Tritium:
- 1. Total Release Curies 7.03E+1 1.27E+1 3.10E+1
- 2. Average Release Rate for Period uCi/sec 8.85E+0 1.60E+0 E. Percentaae of Technical Specification Limits
- 1. Total Body Dose Rate % 1.13E-4 7.85E-3
- 2. Skin Dose Rate % 8.79E-5 1.91E-3
- 3. Critical Organ Dose Rate % 6.01E-3 8.81E-4
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TABLE 2A NORTH ANNA POWER STATION ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT LIOUID EFFLUENTS-SUMMATION OF ALL RELEASES FOR (01/95 - 12/95)
Page 1 of 2 1st 2nd ESTIMATED TOTAL UNITS OUARTER OUARTER PERCENT ERROR (%)
A. Fission and Activation Products:
- 1. Total release (not including tritium, noble gas, and gross alpha). Curies 1.12E-1 7.47E-2 2.00E+1
- 2. Average diluted concentration during the oeriod. uCi/ml 2.51E-10 1.51E-10
- 3. Percent of acolicable limit (T.S.) % 1.85E-3 1.64E-4 D. Tritium:
- 1. Total release activity. Curies 3.49E+2 1.77E+2 2.00E+1
- 2. Average diluted concentration durina the oeriod. uCi/ml 7.81E-7 3.57E-7
- 3. Percent of anolicable limit (T.S.) % 7.82E-3 3.59E-3 C. Dissolved and Entrained Gases:
- 1. Total release activity. Curies 3.51E-4 4.60E-5 2.00E+1 .
- 2. Average diluted concentration during the period. uCi/ml 7.85E-13 9.27E-14
- 3. Percent of applicable limit (T.S.) % 3.92E-7 4.65E-8 D. Gross Alpha Radioactivity:
- 1. Total release activity. Curies 1.42E-4 N/D 2.00E+1 E. Volume of waste released: (prior to dilution). Liters 7.15E+7 5.90E+7 3.00E+0 F. Total volume of dilution water used durina the ceriod. Liters 4.47E+11 4.96E+11 3.00E+0
TABLE 2A NORTH ANNA POWER STATION ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT LIOUID EFFLUENTS-STNMATION OF ALL pur.w1SES FOR (01/95 - 12/95)
Page 2 of 2 3rd 4th ESTIMATED TOTAL UNITS OUARTER QUARTER PERCENT ERROR (%)
A. Fission and Activation Products:
- 1. Total release (not including tritium, noble gas, and gross alpha). Curies 3.66E-2 1.29E-1 2.00E+1
- 2. Average diluted concentration durina the ceriod. uCi/ml 4.20E-11 1.83E-10
- 3. Percent of applicable limit (T.S.) % 8.06E-5 2.37E-4 B. Tritium:
- 1. Total release activity. Curies 5.47E+1 3.96E+2 2.00E+1
- 2. Average diluted concentration durino the oeriod. uCi/ml 6.28E-8 5.63E-7
- 3. Percent of apolicable limit (T.S.) % 6.28E-4 5.62E-3 C. Dissolved and Entrained Gases:
- 1. Total release activity. Curies 3.00E-4 N/D 2.00E+1
- 2. Average diluted concentration during the ceriod. uCi/ml 3.44E-13 N/D
- 3. Percent of applicable limit (T.S.) % 1.72E-7 N/A D. Gross Alpha Radioactivity:
- 1. Total release activity. Curies N/D N/D 2.00E+1 '
E. Volume of waste released: (prior to dilution). Liters 7.08E+7 5.41E+7 3.00E+0 F. Total volume of dilution water used durina the period. Liters 8.71E+11 7.04E+11 3.00E+0
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TABLE 3 NORTH ANNA POWER STATION RADIOACTIVE EFFLUENT RELEASE REPORT SUMMATION OF SOLID RADIOACTIVE WASTE AND IRRADIATED FUEL SHIPMENTS FOR 01-01-95 THROUGH 12-31-95 Pace 1 of 2 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) l 12-Month Estimated Total
- 1. Type of Waste Unit Period Percent Error (%)
l
- a. Spent resins, sludges, m3 6.98E+1* 2.50E+1 filters, filter sludge, Ci 8.05E+2 2.50E+1 evaporator bottoms , etc,
- b. Dry compressible waste, m3 6.61E+2** 2.50E+1 l contaminated equipment, Ci 1.58E+1 2.50E+1
! etc.
- c. Irradiated components, m2 0.00E0 0.00E0 control rods, etc. Ci 0.00E0 0.00E0
- d. Other (describe) m) 4.10E+0*** 2.50E+1 Waste Oil Ci 1.91E-2 2.50E+1 l 2. Estimate of major nuclide composition (by type of waste)
- a. Ni-63 35.9% 2.87E+2 2.50E+1 Co-60 28.7% 2.29E+2 2.50E+1 Cs-137 5.7% 4.59E+1 2.50E+1 Fe-55 12.0% 9.56E+1 2.50E+1 Cs-134 2.8% 2.27E+1 2.50E+1 Co-58 13.1% 1.05E+2 2.50E+1 Mn-54 1.3% 1.03E+1 2.50E+1
- b. Mn-54 2.0% 3.15E-1 2.50E+1 Co 58 10.1% 1.59E0 2.50E+1 i Fe-55 77.6% 1.22E+1 2.50E+1 l
Co-60 4.3% 6.76E-1 2.50E+1 Ni-63 4.9% 7.77E-1 2.50E+1 c.
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i TABLE 3 NORTH ANNA POWER STATION RADIOACTIVE EFFLUENT RWTRASE REPORT SUMMATION OF SOLID RADIOACTIVE WASTE AND IRRADIATED FUEL SHIPMENTS FOR 01-01-95 THROUGH 12-31-95 Page 2 of 2
- 2. Estimate of major nuclide composition (by type of 12-Month Estimated Total waste) (cont.) Unit Period Percent Error (%)
- d. Ce-144 1.2% 2.24E-4 2.50E+1 Cs-137 9.6% 1.84E-3 2.50E+1 Co-60 3.1% 5.85E-4 2.50E+1 Ni-63 38.0% 7.27E-3 2.50E+1 Fe-55 32.8% 6.27E-3 2.50E+1 Sr-90 13.6% 2.60E-3 2.50E+1 Ac-110m 1.8% 3.48E-4 2.50E+1
- 3. Solid Waste Disposition Number of Shioments Mode of Transoortation Destination 11 Truck Barnwell, SC 14 Truck Oak Ridge, TN (SEG) 4 Truck Wampum,PA (ALARON)
B. Irradiated Fuel Shipments (Disposition)
Number of Shioments Mode of Transoortation Destination N/A N/A N/A 1 shipment of resin and 2 shipments of sludge were shipped from North Anna to a Licensed Waste Processor for volume reduction. Therefore, the volume listed for this type is not representative of actual volume buried. The total volume buried for this reporting period was 74.3 m2
- 16 shipments of dry compressible waste / contaminated equipment were shipped from North Anna to Licensed Waste Processors for volume reduction. Therefore, the volume listed for this type is not representative of the actual volume buried. The total volume buried for this reporting period was 54.1 m'.
- 2 shipments of waste oil and 1 shipment of grease were shipped f rom North Anna to a Licensed Waste Processor for incineration. Therefore, I the volume listed for this type is not representative of the actual volume buried. The total volume buried for this reporting period was 0.00 m). J
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l ATTACHMENT 2 .
ANNUAL AND OUARTERLY DOSES (01/95 - 12/95) 1 l
1 An assessment of radiation doses to the maximum exposed nember of the
. public due to radioactive liquid and gaseous effluents released from the site l
l for each calendar quarter for the calendar year of this report, along with an l
l annual total of each effluent pathway will be made pursuant to the ODCM Section 6.7.2.
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Liouid Effluents:
lat 2nd 3rd 4th Annual Quarter Quarter Quarter Quarter Total Total Body Dose (jagem) 9.90E-2 5.64E-2 1 C4F.-2 1.09E-1 2.81E-1 Critical Organ Dose (arem) 1.02E-1 6.09E-2 1.68E-J 1.10E-1 2.89E-1 l
- Gaseous Effluents
1st 2nd 3rd 4th Annual Ouarter Quarter Quarter Quarter Total l Noble Gas l
Gamma Dose (mrad) 1.21E-2 2.85E-4 1.51E-4 1.04E-2 2.29E-2 Noble Gas Beta Dose (mrad) 5.67E-3 7.16E-4 5.72E-4 3.74E-3 1.07E-2 Critical Organ Dose for I-131, I-133, H-3, 4.17E-3 1.31E-2 3.79E-3 6.19E-4 2.18E-2 i Particulates with TM > 8 days (arem) 8 i
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! ATTAC1 DENT 3 i REVISIONS TO OFFSITE POSE CALCULATION M7NUAL l
l (ODCM) l l (01/95 -
12/95) l As required by Technical Specification 6.15, revisions to the ODCM, effective for the time period covered by this report, are summarized in this
, attachment.
1 l
1 There are two procedure changes to Revision 6 and one procedure revision implemented during the period January 1 through December 31, 1995. Included in this attachment are the revision summaries and associated page changes to l
the ODCM corresponding to North Anna's procedure changes. Revision 7, effective October 31, 1995, incorporated the twc procedure changes and additional changes described in the revision suonary.
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i The changes which relate to North Anna Power Station are the following:
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- 1. 6-PN1 -
a correction of Emergency TLD locations and reference to Deviation Report, DR N94-1137, which was implemented January 3, 1995, and
- 2. Revision 7 - correction of units of I-131 concentration contained in the definition of Dose Equivalent I-131, which was implemented October 31, 1995.
9 1
FOR INFORMATION ONLY 9 Station Administrative Procedure C 1994 by Virginia Power. All Rights Reserved
Title:
Offsite Dose Calculation Manual Lead Department: Radiological Protection Procedure Number Revision Number Effective Date VPAP-2103 6-PNI l-3-95 Revision Summary
. Reference commitment 3.2.2 in Step 6.6.1.a.2 and Attachment 23 to ensure the commitment is not deleted at a later date.
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. Correct page number referenced in Attachment 6 page 1 of 5 definitions fi and Ai. Add definition of"21"in equation Ai on Attachment 6 page 2 of 5.
. On Attachment 7 corrected ' Total Body Ai and Critical Organ Ai" to " Total Body Bi and Liver Bi".
. Added alphabetic identifiers to first three items in first column on Attachment 10 and 11.
- Changed note 1 on Attachment 17 page 3 of 3 from " Automatic isolation of this pathway" to " Automatic actuation of the valves in this pathway".
. Change step 6.7.2.a.3 to include classification of unplanned liquid and gaseous effluent releases for the Annual Radioactive Effluent Release Report.
. Delete Bi-monthly River Water samples from Attachment 22 page 3 of 4.
. Change location of sample points for Oysters.
. Delete sediment requirements at Burwell's Bay and Newport News
. Delete sample point of clams at Jamestown. The number of required samples was reduced from 5 to 4.
. Change collection frequency for fish and invertebrates in sections a and b on Attachment 20 page 2 of 3 from Bi-monthly to Semi-Annually.
. Correct distance location of TLD on Attachment 22 page 1 of 4 from .33 to .29 n.iles.
. Incorporate PN&St. Change equipment mark numbers to new format.
. Changed Attachments 14 and 16 to include mark numbers for VG-RM-104 and ventilation flow rate monitors.
. E Par PNI Correct TLD locations on Attachment 23 and referenced DR N941137. North Anna Only Surry Power Station North Anna Power Station Approved by: Approved by:
M M W. R. Matthews 1-3-95 SNSOC Chairman Date SNSOC Chairman Date Approved by: Approved by:
E E J. A. Stall 1 9,5 Station Manager Date Station Manager Date Approved by: M M Vice President-Nuclear Operations Date
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- Station Administrative Procedure l VIRGINIA POWER e ms y y,,,,,m,,,. xii n,,su n n.a
Title:
Offsite Dose Calculation Manual Lead Department: Radiological Protection i
Procedure Number Revision Number Effective Date l VPAP-2103 7 10-31-95 l Revision Summary
! . Incorporates PSI: Changes ATI'ACHMENT 22, Surry Environmental Sampling Locations to
! replace Walnut Point oyster sampling location with Kingsmill oyster sampimg location on i page 128 l
- Incorporates PN1: Changes ATI'ACHMENT 23, North Anna Environmental Sampling i Locations to correct Emergency TLD locations on pages 129 and 130,
! adds reference 3.1.22, Deviation Report N94-1137, Improper Placement of Emergency TLDs to j Page 9 j d(N . Corrects units of I l31 concentration to pCi/cc at 4.5, Dose Equivalent I-131 on page 10 .
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- Reflects Stury Core Uprate to 2546 MWt at 4.13, Rated Thermal Power on page 12 )
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l Surry Power Station North Anna Power Station 1
Approved by: Approved by:
! h & 9f2ll%~ b 10. SAT j SNSOUChairman Date SNSOC Chairman Date
{ Appro ed by: Approved by:
Aq mf to/M M
- Station Manager Date i Station Manager Date l j
Approved by: /
j Vice President-Nuclear Operations Date k
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VIRGINIA VPAP-2103 l POWER REVISION 7 l PAGE 3 OF 156 i
TABLE OF CONTENTS Section Page 1.0 PURPOSE 7 2.0 SCOPE 7
3.0 REFERENCES
/ COMMITMENT DOCUMENTS 8 4.0 DEFINITIONS 9 5.0 RESPONSIBILITIES 13 6.0 INSTRUCTIONS 15 )
6.1 Sampling and Monitoring Criteria 15 l
6.2 Liquid Radioactive Waste Effluents 15
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6.2.1 Liquid Effluent Concentration Limitations 15
- O 6.2.2 Liquid Monitoring Instrumentation 16 lO 6.2.3 Liquid Effluent Dose Limit 6.2.4 Liquid Radwaste Treatment 20 23 6.2.5 Liquid Sampling 24 6.3 Gaseous Radioactive Waste Effluents 24 6.3.1 Gaseous Effluent Dose Rate Limitation 24 6.3.2 Gaseous Monitoring Instrumentation 27 6.3.3 Noble Gas Effluent Air Dose Limit 30 6.3.4 I-131,133, H 3 & Radionuclides in Particulate Form Effluent Dose Limit 33 l
l 6.3.5 Gaseous Radwaste Treatment 36 4 6.4 Radioactive Liquid and Gaseous Release Permits 38 6.4.1 Liquid Waste Batch Release Permits 38 6.4.2 Continuous Release Permit 39 6.4.3 Waste Gas Decay Tank (WGDT) Release Permit 39 6.4.4 Reactor Containment Release Permits 40
- ,O 6.4.5 Miscellaneous Gaseous Release Permit 40
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 4 OF 156 TAllLE OF CONTENTS (continued)
Section Page 6.4 Radioactive Liquid and Gaseous Release Permits (continued) 6.4.6 Radioactive Liquid and Gaseous Release Controls 40 6.5 Total Dose Limit to Public From Uranium Fuel Cycle Sources 41 6.6 Radiological Environmental Monitoring 43 6.6.1 Monitoring Program 43 6.6.2 Land Use Census 45 6.6.3 Interlaboratory Comparison Program 46 6.7 Reporting Requirements 47 6.7.1 Annual Radiological Environmental Operating Report 47 6.7.2 Annual Radioactive Effluent Release Report 48 6.7.3 Annual Meteorological Data 49 6.7.4 Changes to the ODCM 50 7.0 RECORDS 51 ATTACHMENTS 1 Surry Radioactive Liqcid Effluent Monitoring Instrumentation 53 2 North Aima Radioactive Liquid Effluent Monitoring Instrumentation 55 3 Surry Radioactive Liquid Effluent Monitoring Instrumentation 57 Surveillance Requirements 4 North Anna Radioactive Liquid Effluent Monitoring Instrumentation 59 Surveillance Requirements 5 Liquid Ingestion Pathway Dose Factors for Surry Station Units 1 and 2 61 6 North Anna Liquid Ingestion Pathway Dose Factor Calculation Units I and 2 63 !
l 7 North Anna Liquid Pathway Dose Commitment Factors for Adults 69 8 Surry Radioactive Liquid Waste Sampling and Analysis Program 71 l
9 North Anna Radioactive Liquid Waste Sampling and Analysis Program 75 10 Surry Radioactive Gaseous Waste Sampling and Analysis Program 79 11 North Anna Radioactive Gaseous Waste Sampling and Analysis Program 85 12 Gaseous Effluent Dose Factors for Surry 89 13 Gaseous Effluent Dose Factors for North Anna 93 1
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 5 OF 156 TABLE OF CONTENTS (continued)
I Section Page ATI'ACHMENTS (continued) 14 Surry Radioactive Gaseous Effluent Monitoring Instrumentation 97 15 Surry Radioactive Gaseous Effluent Monitoring Instrumentation 99
- 16 Surry Radioactive Gaseous Effluent Monitoring Instrumentation 105 Surveillance Requirements 17 North Anna Radioactive Gaseous Effluent Monitoring Instrumentation 107 Surveillance Requirements 18 Critical Organ and Inhalation Dose Factors for Surry 111 l 19 Critical Organ Dose Factors for North Anna 113 l 20 Surry Radiological Environmental Monitoring Program 115 21 North Anna Radiological Environmental Monitoring Program 119 22 Surry Environmental Sampling Locations 125 23 North Anna Environmental Sampling Locatimu 129 l 24 Detection Capabilities for Surry Environmental Sample Analysis 133 25 Detection Capabilities for North Anna Environmental Sample Analysis 135 26 Reporting Levels for Radioactivity Concentrations in Environmental 137 Samples at Surry
. 27 Reporting Levels for Radioactivity Concentrations in Environmental 139 Samples at North Anna 28 Surry Meteorological, Liquid, and Gaseous Pathway Analysis 141 29 North Anna Meteorological, Liquid, and Gaseous Pathway Analysis 149 4
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ae_ e.a4-_s44.aa-.u. man ea mA='-aevh. 4 .ma..e.mns>s.-L-wu a_ m m-h uAu as wh m m_smed.4+me,/.u4 aJ,rJae--4m-*-4E+A6 ---44 d-.J.4mm__wJ_mA4 A,..._m,m,4 -
% m am. aam 2__.m__ m_,a VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 6 OF 156 i
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 7 OF 156 l.0 PURPOSE The Offsite Dose Calculation Manual (ODCM) establishes requirements for the Radioactive Effluent and Radiological Environmental Monitoring Programs. Methodology and parameters
, are provided to calculate offsite doses resulting from radioactive gaseous and liquid effluents, to calculate gaseous and liquid effluent monitoring alarm / trip setpoints, and to conduct the Environmental Monitoring Program. Requirements are established for the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report required by Station Technical Specifications. Calculation of offsite doses due to radioactive liquid and gaseous effluents are performed to assure that:
. Concentration of radioactive liquid effluents to the unrestricted area will be limited to ten times the effluent concentration values of 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases and 2E-4 pCi/ml for dissolved or entrained noble gases.
- Exposure to the maximum exposed member of the public in the unrestricted area from radioactive liquid effluents will not result in doses greater than the liquid dose limits of p 10 CFR 50, Appendix 1 l V + Dose rate at and beyond the site boundary from radioactive gaseous effluents will be limited to:
.. Noble gases -less than or equal to a dose rate of 500 mrem /yr to the total body and less than or equal to a dose rate of 3000 mrem /yr to the skin
.1131, 1 133, and H3 , and all radionuclides in particulate form with half-lives greater than 8 days -less than or equal to a dose rate of 1500 mrem /yr to any organ
. Exposure from radioactive gaseous effluents to the maximum exposed member of the public in the unrestricted area will not result in doses greater than the gaseous dose limits of 10 CFR 50, Appendix 1, and
. Exposure to a real individual will not exceed 40 CFR 190 dose limits 4
2.0 SCOPE This procedure applies to the Radioactive Effluent and Environmental Monitoring Programs at Surry and North Anna Stations.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 8 OF 156
3.0 REFERENCES
/ COMMITMENT DOCUMENTS 3.1 References 3.1.1 10 CFR 20, Standards for Protection Against Radiation 3.1.2 10 CFR 50, Domestic Licensing of Production and Utilization Facilities 3.1.3 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations 3.1.4 TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites 3.1.5 Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, Rev.1, U.S. NRC, June 1974 3.1.6 Regulatory Guide 1.109, Calculation of Annual Doses to Man From Routine Rele.ases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR 50, Appendix I, Rev.1, U.S. NRC, October 1977 3.1.7 Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Rev.1, U.S. NRC, July 1977 3.1.8 Surry and North Anna Technical Specifications (Units 1 and 2) 3.1.9 NUREG-0324, XOQDOQ, Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, U.S. NRC, September 1977 i 3.1.10 NUREG/CR-1276, Users Manual for the LADTAP II Program, U.S. NRC, May,1980 3.1.11 NUREG-0597, User's Guide to GASPAR Code, U.S. NRC, June,1980 3.1.12 Radiological Assesament Branch Technical Position on Environmental Monitoring, November,1979, Rev.1 3.1.13 NUREG-0133, Preparation of Radiological Effluent Technical Specifications for :
Nuclear Power Stations, October,1978
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3.1.14 NUREG-0543, February 1980, Methods for Demonstrating LWR Compliance With I the EPA Uranium Fuel Cycle Standard (40 CFR Part 190) 3.1.15 NUREG-0472, Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors, Rev. 3, March 1982 3.1.16 Environmental Measurements Laboratory, DOE HASL 300 Manual 3.1.17 NRC Generic Letter 89-01, Implementation of Programmatic Controls for l Radiological Effluent Technical Specifications (RETS)in the Administrative Controls l Section of the Technical Specifications and the Relocation of Procedural Details of l RETS to the Offsite Dose Calculation Manual or to the Process Control Program ;
3.1.18 UFSAR (Surry and North Anna) l l
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1 VIRGINIA VPAP-2103 i POWER REVISION 7 PAGE 9 OF 156 t
% )) 3.1.19 Nuclear Reactor Environmental Radiation Monitoring Quality Control Manual, l
IWL-0032-361 1 3.1.20 VPAP-2802, Notifications and Reports 3.1.21 NAPS Circulating Water System Modifications
- a. DC-85-37-1 Unit 1
- b. DC-85-38-2 Unit 2 3.1.22 Deviation Report N94-1137, Improper Placement of Emergency TLDs l 3.2 Commitment Documents 3.2.1 Quality Assurance Audit Report Number C 90-22, Management Safety Review Committee, Observation 03C, January 17,1991 3.2.2 Quality Assurance Audit Report Number 91-03, Observation 08N 3.2.3 Quality Assurance Audit Report Number 92-03, Observation 02N 3.2.4 Quality Assurance Audit Report Number 92-03, Observation 04NS (Item 2) 4.0 DEFINITIONS 4.1 Channel Calibration m
Adjustment, as necessary, of the channel autr.it so it responds with the necessary range and accuracy to known values of the parameter the channel monitors. It encompasses the entire channel, including the sensor and alarm and/or trip functions and the Channel Functional Test.
The Channel Calibration can be performed by any series of sequential, overlapping, or total channel steps so the entire channel is calibrated.
4.2 Channel Check A qualitative assessment, by observation, of channel behavior during operation. This assessment includes, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.
4.3 Channel Functional Test There are two types of Channel Functional Tests.
4.3.1 Analog Channel Injection of a simulated signal into a channel, as close to the secar as practicable, to
(~} verify Operability, including alarm and/or trip functions.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 10 OF 156 4.3 Channel Functional Test (continued) -
4.3.2 Bistable Channel Injection of a simulated signalinto a sensor to verify Operability, including alarm and/
or trip functions.
4.4 Critical Organ That organ, which has been determined to be the maximum exposed organ based on an effluent pathway analysis, thereby ensuring the dose and dose rate limitations to any organ will not be exceeded.
4.5 Dose Equivalent I-131 That concentration of 1131 ( Ci/cc ) that alone would produce the same thyroid dose as the l 131 ,1132,g133,iIM, and 1135 actually present. Thyroid dose quantity and isotopic mixture of 1 conversion factors iat this calculation are listed in Table III of TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites. Thyroid dose conversion factors from NRC Regulatory Guide 1.109, Revision 1, may be used (Surry).
4.6 Frequency Notations g NOTE: Frequencies are allowed a maximum extension of 25 percent. I NOTATION FREQUENCY D - Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W - Weekly At least once per 7 days M - Monthly At least once per 31 days Q - Quarterly At least once per 92 days S A - Semi-annually At least once per 184 days R - Refueling At least once per 18 months S/U - Start-up Prior to each reactor start-up P - Prior to release Completed prior to each release N. A. - Not applicable Not applicable DR - During the release At least once during each release j l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 11 OF 156 bJ 4.7 Gaseous Radwaste Treatment System A system that reduces radioactive gaseous effluents by collecting primary coolant system l offgases from the primary system and providing delay or holdup to reduce total radioactivity prior to release to the environment. The system comprises the waste gas decay tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tanks, and waste gas diaphragm compressor (North Anna). l 1
General Nomenclature 4.8 x = Chi: concentration at a point at a given instant (curies per cubic meter) i i
D = Deposition: quantity of deposited radioactive material per unit area (curies per square meter)
Q = Source strength (instantaneous; grams, curies) j
= Emission rate (continuous; grams per second, curies per second) ;
= Emission rate (continuous line source; grams per second per meter) I l
4.9 Lower Limit of Detection (LLD) l The smallest concentration of radioactive material in a sample that will yield a net count (above g
V system background) that can be detected with 95 percent probability with only 5 percent j probability of falsely concluding that a blank observation represents a "real" signal. )
1 4.10 Members of the Public Individuals who, by virtue of their occupational status, have no formal association with the Station. This category includes non-employees of Virginia Power who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with Station functions. This category does not include non-employees such as vending machine servicemen or postal workers who, as part of their formal job function, occasionally enter an area that is controlled by Virginia Power to protect individuals from exposure to radiation and radioactive j materials. I 4.11 Operable - Operability A system, subsystem, train, component, or device is operable or has operability when it is l capable of performing its specified functions and all necessary, attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its functions are also capable of performing their related support functions.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 12 OF 156 4.12 Purge - Purging 9
Controlled discharge of air or gas from a confinement to maintain temperature, pressure, i humidity, concentration, or other operating condition, so that replacement air or gas is required to purify the confinement. ,
l 4.13 Rated Thermal Power Total reactor core heat transfer rate to reactor coolant.
. Surry - 2546 Megawatts Thermal (MWt) l l
. North Anna-2893 MWt 4.14 Site Boundary l l
The line beyond which Virginia Power does not own, lease, or otherwise control the land. l 4.15 Source Check A qualitative assessment of channel response when a channel sensor is exposed to radiation. I This applies to installed radiation monitoring systems.
4.16 Special Report A report to NRC to comply with Subsections 6.2,6.3, or 6.5 of this procedure. Also refer to VPAP-2802, Notifications and Reports.
4.17 Thermal Power Total reactor core heat transfer rate to the reactor coolant.
4.18 Unrestricted Area Any area at or beyond the site boundary, access to which is neither limited nor controlled by Virginia Power for purposes of protection of individuals from exposure to radiation and I radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional or recreational purposes.
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VIRGINIA VPAP-2103 POWER REVISION 7 i PAGE 13 OF 156 )
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O 4.19 Ventilation Exhaust Treatment System l l A system that reduces gaseous radioiodine or radioactive material in particulate form in l
effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and High l
l Efficiency Particulate Air (HEPA) filters to remove iodines and particulates from a gaseous exhaust stream prior to release to the environment (such a system is not considered to have any l effect on noble gas effluents). Engineered Safety Feature (ESP) atmospheric cleanup systems )
I l are not Ventilation Exhaust Treatment System components.
i 5.0 RESPONSIBILITIES 5.1 Superintendent Radiological Protection The Superintendent Radiological Protection is responsible for:
5.1.1 Establishing and maintaining procedures for surveying, sampling, and monitoring l radioactive effluents and the environment.
l' 5.1.2 Surveying, sampling, and analyzing plant effluents and environmental monitoring, and documenting these activities.
l (d 5.1.3 Analyzing plant effluent trends and recommending actions to correct adverse trends.
5.1.4 Preparing Effluent and Environmental Monitoring Program records.
5.2 Superintendent Operations The Superintendent Operations is responsible for requesting samples, analyses, and l authorization to release effluents.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 15 OF 156 O 6.0 INSTRUCTIONS NOTE: Meteorological, liquid, and gaseous pathway analyses are presented in Attachments 28 and 29, Meteorological, Liquid, and Gaseous Pathway Analysis.
6.1 Sampling and Monitoring Criteria 6.1.1 S urveys, sampling, and analyses shall use instruments calibrated for the type and range of radiation monitored and the type of discharge monitored.
6.1.2 Installed monitoring systems shall be calibrated for the type and range of radiadon or parameter monitored.
6.1.3 A sufficierit number of survey points shall be used or samples taken to adequately assess the status of the discharge monitored.
6.1.4 Samples shall be representative of the volume and type of discharge monitored.
6.1.5 Surveys, sampling, analyses, and monitoring records shall be accurately and legibly j p documented, and sufficiently detailed that the meaning and intent of the records are !
O clear.
l 6.1.6 Surveys, analyses, and monitoring records shall be reviewed for trends, completeness, and accuracy.
l 6.2 Liquid Radioactive Waste Effluents 6.2.1 Liquid Effluent Concentration Limitations
- a. Liquid waste concentrations discharged from the Station shall not exceed the following limits: j
- 1. For radionuclides (other than dissolved or entrained noble gases), liquid effluent concentrations released to unrestricted areas shall not exceed ten times l the effluent concentradon values specified in 10 CFR 20, Appendix B, Table 2, Column 2.
- 2. For dissolved or entrained noble gases, concentrations shall not exceed 2E-4 pCi/ml.
- b. If the concentration ofliquid effluent exceeds the limits in 6.2.1.a., promptly reduce
( concentrations to within limits.
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VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 16 OF 156
! 6.2.1 Liquid Effluent Concentration Limitations (continued)
O l c. Daily concentrations of radioactive materials in liquid waste released to ,
unrestricted areas shall meet the following: l l
l Volume of Waste Discharged + Volume of Dilution Water (;) ,
1 Ci/ml.' )
Volume of Waste Discharged x [ ACW. I I i 1
l l
l where:
1 pCi/mli = the concentration of nuclide iin the liquid effluent discharge ACW; = ten times the effluent concentration value in unrestricted areas of nuclide i, expressed as pCi/ml from 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than noble gases, and 2E-4 Ci/ml for dissolved or entrained noble gases 6.2.2 Liquid Monitoring Instrumentation g
- a. Radioactive Liquid Effluent Monitoring Instrumentation W Radioactive liquid effluent monitoring instrumentation channels shown on Attachments 1 and 2, Radioactive Liquid Effluent Monitoring Instrumentation, i shall be operable with their alarm / trip setpoints set to ensure that 6.2.1.a. limits are not exceeded.
- 1. Alarm / trip setpoints of these channels shall be determined and adjusted in accordance with 6.2.2.d., Setpoint Calculation.
- 2. If a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint is less conservative than required by 6.2.2.a., perform one of the following:
- Promptly suspend release of radioactive liquid effluents monitored by the affected channel l
- Declare the channelinoperable !
+ Change the setpoint to an acceptable, conservative value O
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 17 OF 156 O b. Radioactive Liquid Emuent Monitoring Instrumentation Operability Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated operable by performing a Channel Check, Source Check, Channel Calibration, and Channel Functional Test at the frequencies shown in Attachments 3 and 4, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.
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- 1. If the number of cperable channels is less than the minimum required by the l
tables in Attachment 1 or 2, perform the action shown in those tables.
- 2. Attempt to retum the instruments to operable status within 30 days. If unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- c. Applicable Monitors Liquid effluent monitors for which alarm / trip setpoints shall be determined are:
Release Point Instrument Number )
North Anna Surry V Liquid Radwaste Effluent Line 1-LW-RM-111 N/A ;
i Service Water System Effluent Line 1-SW-RM-108 1-SW-RM-107 A, '
B,C,D Condenser Circulating Water Line 1-SW-RM-130 1-SW-RM-120 2-SW-RM-230 2-SW-RM-220 Radwaste Facility Effluent Line N/A 1-RRM-RITS-131 l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 18 OF 156 6.2.2 Liquid Monitoring Instrumentation (continued)
- d. Setpoint Calculation NOTE: This methodology does not preclude use of more conservative setpoints.
- 1. Maximum setpoint values shall be calculated by:
CF D S= p (2) ,
E l where: ,
1 S = the setpoint, in pCi/ml, of the radioactivity monitor measuring the l radioactivity concentration in the effluent line prior to dilution C= the effluent concentration limit for the monitor used to implement 10 CFR 20 for the Station,in pCi/ml FE= maximum design pathway effluent flow rate Fo = dilution water flow rate calculated as:
(Surry) D = F E+ (200,000 gpm x number of cire. pumps in service)
(N. Anna) D = Fs + (218,000 gpm x number of cire, pumps in service)
- 2. Each of the condenser circulating water channels (Surry: SW-120, SW-220)
(North Anna: SW-130, SW-230) monitors the effluent (service water, including component cooling service water, circulating water, and liquid radwaste) in the l circulating water discharge tunnel beyond the last point of possible radioactive l material addition. No dilution is assumed for this pathway. Therefore, Equation l (2) becomes: l S=C (3)
I The setpoint for Station monitors used to implement 10 CFR 20 for the site becomes the effluent concentration limit. ;
- 3. In addition, for added conservatism, setpoints shall be calculated for the liquid radwaste effluent line (North Anna: LW-111), the service water system effluent line (Surry: SW-107 A, B, C, and D, North Anna: SW-108), and the Radwaste Facility effluent line (Surry: RRM-131).
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 19 OF 156
- 4. For the liquid radwaste effluent line, Equation (2) becomes:
CFD Kgw S= (4)
FE where:
Kw t
= The fraction of the effluent concentration limit, used to implement 10 CFR 20 for the site, attributable to the liquid radwaste effluent line pathway
- 5. For the service water system effluent line, Equation (2) becomes:
CFn3K ,
S= (5)
F E
where:
Ksw = The fraction of the effluent concentrat'on limit, used to implement s 10 CFR 20 for the Station, attributable to the service water effluent line pathway
- 6. For the Radwaste Facility effluent line, Equation (2) becomes:
CFDKRW S= (6)
FE where:
Kgw = The fraction of the effluent concentration limit, used to implement 10 CFR 20 attributable to the Radwaste Facility effluent line ,
pathway
- 7. The sum Kew + Ksw + K W R shall not be greater than 1.0.
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VIRGINIA VPAP-2103 j
POWER REVISION 7 PAGE 20 OF 156 l
6.2.3 Liquid Effluent Dose Limit 0
- a. Requirement At least once per 31 days, perform the dose calculations in 6.2.3.c. and 6.2.3.d. to ensure the dose or dose commitment to the maximum exposed member of the public from radioactive materials in liquid releases (from each reactor unit) to unrestricted areas is limited to: )
l
- 1. During any calendar quarter: 1 1
. Less than or equal to 1.5 mrem to the total body
. Less than or equal to 5 mrem to the critical organ 1 I
( 2. During any calendar year: ;
t
. Less than or equal to 3 mrem to the total body
- Less than or equal to 10 mrem to the critical organ l l
- b. Action
! If the calculated dose from release of radioactive materials in liquid effluents exceeds any of the above limits, prepare and submit to the NRC, within 30 days, a l special report in accordance with VPAP-2802, Notifications and Reports, that l identifies causes for exceeding limits and defines corrective actions taken to reduce l releases of radioactive materials in liquid effluents to ensure that subsequent releases will be in compliance with the above limits.
l c. Surry Dose Contribution Calculations '
l l
NOTE: Thyroid and GI-LLI organ doses must be calculated to determine which is the critical 1
organ for the period being considered. l Dose contributions shall be calculated for all radionuclides identified in liquid effluents released to unrestricted areas based on the equation:
D = t F M [C;A; (7) i j 1
where- i Subscripts = i, refers to individual radionuclide
VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 210F 156
!O D= the cumulative dose commitment to the total body or. critical organ from the j liquid effluents for the period t, in mrem i t = the period for which C and i F are averaged for allliquid releases,in hours M= the mixing ratio (reciprocal of the dilution factor) at the point of exposure, dimensionless,0.2 from Appendix 11 A, Surry UFS AR F = the near field average dilution factor for Ci during any liquid effluent release; the ratio of the average undiluted liquid waste flow during release to the average flow from the site discharge structure to unrestricted areas C= i the average concentration of radionuclide, i, in undiluted liquid effluent
- during the period t, frcm all liquid releases, in Ci/ml
- A; = the site-related ingestion dose commitment factor to the total body or critical 4
organ of an adult for each identified principal gamma and beta emitter in mrem-ml per hr-pCi. Values for A iare given in Attachment 5, Liquid
) Ingestion Pathway Dose Factors For Surry Power Station.
A; = 1.14 E+05 (21BF; + 5BI;) DF; (8)
A where:
, V 1.14 E+05 = 1 E+06 pCi/pCi x 1 E+03 ml/kg/(8760 hr/yr), units conversion factor 21 = adult fish consumption, kg/yr, from NUREG-0133 5 = adult invertebrate consumption, kg/yr, from NUREG-0133 Bl i = the bioaccumulation factor for nuclide i, in invertebrates, pCi/kg per pCi/1, from Table A-1 of Regulatory Guide 1.109, Rev.1 BFi= the bioaccumulation factor for nuclide i, in fish, pCi/kg per pCi/1, from
, Table A-1 of Regulatory Guide 1.109, Rev. I j DFi= the critical organ dose conversion factor for nuclide i, for adults, in mrem /pCi, from Table E-11 of Regulatory Guide 1.109, Rev.1 O
i VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 22 OF 156 6.2.3 Liquid Effluent Dose Limit (continued)
- d. North Anna Dose Contribution Calculations ;
1 NOTE: Attachment 6, North Anna Liquid Ingestion Pathway Dose Factor Calculation !
provides the derivation for Equation (9). l Dose contribution shall be calculated for all radionuclides identified in liquid i effluents released to unrestricted areas based on:
(9) l D = [Q;x B; l i Where:
Subscripts = i, refers to individual radionuclide l
D= the cumulative dose commitment to the total body or critical organ from the liquid effluents for the period t, in mrem Bi = Dose Commitment Factors (mrem /Ci) for adults. Values for Bi are provided in Attachment 7, North Anna Liquid Ingestion Pathway Dose Commitment i Factors for Adults Qi = Total released activity for the considered period and the ith nuclide Q; = t x C; x Waste Flow (10) l Where:
l t = the period for which C; and F are averaged for allliquid releases, in hours C=i the average concentration of radionuclide, i, in undiluted liquid effluent during the period, t, from any liquid releases, in pCi/ml 1
- e. Quarterly Composite Analyses I
For radionuclides not determined in each batch or weekly composite, dose contribution to current monthly or calendar quarter cumulative summation may be approximated by assuming an average monthly concentration based on previous
! monthly or quarterly compcsite analyses. However, for reporting purposes, calculated dose contribution shall be based on the actual composite analyses.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 23 OF 156 A
G 6.2.4 Liquid Radwaste T:eatment
- a. Requirement
- 1. The Liquid Radwaste Treatment System and/or the Surry Radwaste Facility Liquid Waste System shall be used to reduce the radioactive materials in liquid waste prior to discharge when projected dose due to liquid effluent, from each reactor unit, to unrestricted areas would exceed 0.06 mrem to total body or 0.2 mrem to the critical organ in a 31-day period.
- 2. Doses due to liquid releases shall be projected at least once per 31 days.
- b. Action If radioactive liquid waste is discharged without treatment and in excess of the 1
above limits prepare and submit to the NRC, within 30 days, a special report in j accordance with VPAP-2802, Notifications and Reports, that includes the following:
- 1. An explanation of why liquid radwaste was being discharged without treatment, I g identification of any inoperable equipment or sub-system, and the reason for the C inoperability. I
- 2. Actions taken to restore inoperable equipment to operable status.
- 3. Summary description of actions taken to prevent recurrence,
- c. Projected Total Body Dose Calculation
- 1. Determine DTB, the total body dose from liquid effluents in the previous 31-day period, per Equation (7) or Equation (9) (Surry and North Anna, respectively).
- 2. Estimate R ,i the ratio of the estimated volume ofliquid effluent releases in the present 31-day period to the volume released in the previous 31-day period.
- 3. Estimate F 3, the ratio of the estimated liquid effluent radioactivity concentration in the present 31-day period to liquid effluent concentration in the previous 31-day period ( Ci/ml).
- 4. Determine PD a, T the projected total body dose in a 31-day period.
PDTB
- TB@l 1) (II)
(3 G
VIRGINIA VPAP-2103 l POWER REVISION 7 PAGE 24 OF 156 6.2.4 Liquid Radwaste Treatment (continued)
- d. Projected Critical Organ Dose Calculation l
Historical data pertaining to the volumes and radioactivity of liquid effluents )
l released in connection with specific Station functions, such as maintenance or l refueling outages, shall be used in projections as appropriate.
l
- 1. Determine D o, the critical organ aose from liquid effluents in the previous l
31-day period, per Equation (7) or Equation (9) (Surry and North Anna, respectively).
l 2. Estimate R ias in 6.2.4.c.2. l 1
- 3. Estimate F ias in 6.2.4.c.3.
l
- 4. Determine PD o= projected critical organ dose in a 31-day period.
PD g = Dg(R 1Fi )
6.2.5 Liquid Sampling l Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis requirements in Attachments 8 and 9, Radioactive Liquid Waste h
Sampling and Analysis Program (Surry and North Anna, respectively).
l 6.3 Gaseous Radioactive Waste Effluents 6.3.1 Gaseous Effluent Dose Rate Limitation
- a. Requirement Dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to:
- 1. The dose rate limit for noble gases shall be s 500 mrem / year to the total body l and s 3000 mrem / year to the skin.
- 2. The dose rate limit for 113I, II33, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days shall be s 1500 mrem / year i to the critical organ.
i
- b. Action
- 1. If dose rates exceed 6.3.1.a. limits, promptly decrease the release rate to within the above limits.
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VIRGINIA VPAP-2103 POWER REVISION 7 i PAGE 25 OF 156 l
- 2. Dose rates due to noble gases in gaseous effluents shall be determined,
! continue usly, to be within 6.3.1.a. limits.
I 3. Dose rates due to 1131, 1833, tritium, and all radionuclides in particulate form 1
l with half-lives greater than 8 days,in gaseous effluents shall be determined to j I be within the above limits by obtaining representative samples and performing l analyses in accordance with the sampling and analysis program specified on l Attachments 10 and 11, Radioactive Gaseous Waste Sampling and Analysis Program.
I l
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l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 26 OF 156 6.3.1 Gaseous Effluent Dose Rate Limitation (continued)
- c. Calculations of Gaseous Effluent Dose Rates
- 1. The dose rate limit for noble gases shall be determined to be within the limit by limiting the release rate to the lesser of:
l l
[K;yyd;yy+K;pyd;py] s 500 mrem /yr to the total body (13) l 1 UR ,
l hLiyy + 1.1 M;yy ) divy + (Lipy + 1.1 Mjp y)dipy] 5 3000 mrem /yr to the skin (14) l where: l l
Subscripts = vv, refers to vent releases from the building ventilation vent, I i.1cluding Radwaste Facility Ventilation Vent; py, refers to the vent releases from the process vent; i, refers to individual radionuclide Kiyy,Kipy = The total body dose factor for ventilation vents or process vent release due to gamma emissions for each identified noble gas radionuclide i, in mrem /yr per Curie /sec. Factors are listed in Attachments 12 and 13, Gaseous Effluent Dose Factors (Surry and North Anna, respectively)
Livy,L py = The skin dose factor for ventilation vents or process vent release due to beta emissions for each identified noble gas radionuclide i, in mrem /yr per Curie /sec. Factors are listed in Attachments 12 and 13 Miyy,M py = The air dose factor for ventilation vents or process vent release due to gamma emissions for each identified noble gas radionuclide, i, in mrad /yr per Curie /sec. Factors are listed in Attachments 12 and 13 d.d,p, = The release rate for ventilation vents or process vent of noble gas radionuclide i, in gaseous effluents in Curie /sec (per site) 1.1 = The unit conversion factor that converts air dose to skin dose, in mrem / mrad l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 27 OF 156
- 2. The dose rate limit for1131, 1133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days, shall be determined to be within the limit by restricting the release rate to:
[P;,d;yy + P;pyd y]p5; 1500 mrem /yr to the critical organ (15)
I where:
Piyy, Pipy = an dose factor for ventilation vents or process The critical org33, ventfor1131,I 3 H , and all radionuclides in partic with half-lives greater than 8 days, for the inhalation pathway, in mrem /yr per Curie /sec. Factors are listed in Attachments 12 and 13 diodig = The release rate for ventilation vents or process vent of 1131, I I33, H3 , and all radionuclides i, in particulate form with half-lives greater than 8 days,in gaseous effluents in Curie /sec (per site)
- 3. All gaseous releases, not through the process vent, are considered ground level 3 and shall be included in the determination of dm .
(d 6.3.2 Gaseous Monitoring Instrumentation
- a. Requirement
- 1. The radioactive gaseous effluent monitoring instrumentation channels shown in Attachment 14 or 15, Radioactive Gaseous Effluent Monitoring Instrumentation, shall be operable with alarm / trip setpoints set to ensure thct 6.3.1.a. noble gas limits are not exceeded. Alarm / trip setpoints of these etiannels shall be determined and adjusted in accordance with 6.3.2.d.
- 2. Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated operable by Channel Checks, Source Checks, Channel Calibrations, and Channel Functional Tests at the frequencies shown in Attachment 16 or 17, Radioactive Gaseous Effluent Monitoring Instrumenunion Saveillance Requirements.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 28 OF 156 6.3.2 Gaseous Monitoring Instrumentation (continued)
- b. Action
- 1. If a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint is less conservative than required by 6.3.2.a.1, promptly:
. Suspend the release of radioactive gaseous effluents monitored by the affected channel and declare the channel inoperable
. or
. Change the setpoint so it is acceptably conservative
- 2. If the number of operable channels is less than the minimum required by tables in Attachment 14 and 15, take the action shown in those tables.
- 3. Return instruments to operable status within 30 days. If unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- c. Applicable Monitors Rt.dioactive gaseous effluent monitors for which alarm / trip setpoints shall be g determined are: W Release Point Instrument Number North Anna Surry Process Vent l-GW-RM-102 1-GW-RM-102 1-GW-RM-178-1 1-GW-RM-130-1 Condenser Air Ejector 1-SV-RM-121 1-SV-RM-111 2-SV-RM-221 2-SV-RM-211 Ventilation Vent A 1-VG-RM-104 N/A 1-VG-RM-179-1 Ventilation Vent B l-VG-RM-113 N/A 1-VG-RM-180-1 Ventilation Vent No.1 N/A 1-VG-RM-104 Ventilation Vent No. 2 N/A 1-VG RM-110 1-VG-RM-131-1 Radwaste Facility Vent N/A RRM-101 O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 29 OF 156 O d. Setpoint Calculations
- 1. Setpoint calculations for each monitor listed in 6.3.2.c. shall maintain this relationship:
D2Dpy + Dc,, + D yy (16) where:
D = Step 6.3.1.a. dose limits that implement 10 CFR 20 for the Station, mrem /yr D py = The noble gas site boundary dose rate from process vent gaseous i
effluent releases, mrem /yr D cae = The noble gas site boundary dose rate from condenser air ejector gaseous effluent releases, mrem /yr D yy = The noble gas site boundary dose rate from:
Surry: Sumanation of the Ventilation Vents 1,2, and the Radwaste Facility vent gaseous effluent releases, mrem /yr North Anna: Summation of Ventilation Vent A plus B gaseous
,q effluent releases, mrem /yr NJ l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 30 OF 156
- 2. Setpoint values shall be determined by:
0 R* x 2.12 E-03 Cm =
p (17) m where:
m = The release pathway, process vent (pv), ventilation vent (vv) condenser air ejector (cae), or Radwaste Facility (rv)
Cm = The effluent concentration limit implementing 6.3.1.a. for the Station, Ci/ml Rm = The release rate limit for pathway m determined from methodology in 6.3.1.c., using Xe l33 as nuclide to be released, pCi/sec 2.12E-03 = CFM per ml/sec Fm = The maximum flow rate for pathway m, CFM NOTE: According to NUREG-0133, the radioactive effluent radiation monitor alarm / trip setpoints should be based on the radioactive noble gases. It is not practicable to apply instantaneous alarm / trip setpoints to integrating monitors sensitive to radiciodines, radioactive materials in particulate form, and radionuclides oilur than noble gases.
6.3.3 Noble Gas Effluent Air Dose Limit
- a. Requirement
- 1. The air dose in unrestricted areas due to noble gases released in gaseous effluents from each unit at or beyond the site boundary shall be limited to:
. During any calendar quarter: $5 mrads for gamm radiation and 510 mrads for beta radiation
. During any calendar year: $ 10 mrads for gamma radiation and $20 mrads for beta radiation
- 2. Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with 6.3.3.c. at least once per 31 days.
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VIRGINIA VPAP-2103
! POWER REVISION 7 l
PAGE 31 OF 156 O b. Action
! If the calculated air dose from radioactive noble gases in gaseous effluents exceeds any of the above limits, prepare and submit to the NRC, within 30 days, a special l report in accordance with VPAP-2802, Notifications and Reports, that identifies the I causes for exceeding the limits and defines corrective actions that have been taken to reduce releases and the proposed corrective actions to be taken to assure that l subsequent releases will be in compliance with the limits in 6.3.3.a.
l l c. Noble Gas Effluent Air Dose Calculation Gaseous releases, not through the process vent, are considered ground level and shall be included in the determination of Qivv.
j The air dose to areas at or beyond the site boundary due to noble gases shall be l
determined by the followc --
For gamma radiation:
D = 3.17E-08 [M;yyQyy+M;pyU;py]
8 l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 32 OF 156 For beta radiation:
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[N;yyD;yy+N;pyQjpy] (19)
Db = 3.17E-08 1
Where:
Subscripts = vv, refers to vent releases from the building ventilation vents, including the Radwaste Facility Ventilation Vent and air ejectors py, refers to the vent releases from the process vent i, refers to individual radionuclide D8 = the air dose for gamma radiation, in mrad Db = the air dose for beta radiation, in mrad Mivy, Mipv = the air dose factors for ventilation vents or process vent release due to gamma emissions for each identified noble gas radionuclide i, in mrad /yr per Curie /sec. Factors are listed in Attachments 12 and 13 N ivy, Nipv = the air dose factor for ventilation vents or process vent release due to beta emissions for each identified noble gas radionuclide i, in mrad /yr per Curie /sec. Factors are listed in Attachments 12 and 13 Qiyy, Qipy = the release for ventilation vents or process vent of noble gas radionuclide i,in gaseous effluents for 31 days, quarter, or year as appropriate in Curies (per site) 9
l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 33 OF Ifa O 6.3.4 I-131,133, H-3 & Radionuclides In Particulate Form Effluent Dose Limit
- a. Requirement
- 1. Methods shall be implemented to ensure that the dose to any organ of a member of the public from 1131, I133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days,in gaseous effluents released from the site to l unrestricted areas from each reactor unit shall be:
. During any calendar quarter: $ 7.5 mrem to the critical organ
. During any calendar year: s 15 mrem to the critical organ
- 2. Cumulative dose contributions to a member of the public from Il31, 1133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents released to unrestricted areas for the current calendar quarter and current calendar year shall be determined at least once per 31 days in accordance with 6.3.4.c. or 6.3.4.d.
- b. Action If the calculated dose from the release of II31, Il33, tritium, and radionuclides in particulate form, with half-lives greater than 8 days, in p.secus effluents exceeds any of the above limits, prepare and submit to the NRC within 30 days, a special report in accordance with VPAP-2802, Notifications and Reports, that contains the:
- 1. Causes for exceeding limits.
- 2. Corrective actions taken to reduce releases.
- 3. Proposed corrective actions to be taken to assure that subsequent releases will be in compliance with limits stated in 6.3.4.a.
- c. Surry Dose Calculations Gaseous releases, not through the process vent, are considered ground level and shall be included in the determination of b . Historical data pertaining to the volumes and radioactive concentrations of gaseous effluents released in connection to specific Station functions, such as containment purges, shall be used in the estimates, as appropriate, i
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 34 OF 156
- 1. The dose to the maximum exposed member of the public, attributable to gaseous effluents at and beyond the site boundary that contain 1131,1133, tritium, and particulate-form radionuclides with half-lives greater than 8 days, shall be i determined by:
Dr = 3.17E-08 [(RM;yyd yy + RM ,d;p.)+(RI yy d yy + Rl;pydjpy)] (20) i Where:
Subscripts = vv, refers to vent releases from the building ventilation vents, including the Radwaste Facility Ventilation Vent and air ejectors; pv, refers to the vent releases from the process vent Dr = the dose to the critical organ of the maximum exposed member of the public in mrem RMivy, RMipv= the cow-milk pathway dose factor for ventilation vents or process vent release due to 1131, 1133, tritium, and from all particulate-form radionuclides with half-lives greater than 8 days, in mrem /yr per Curie /sec. Factors are listed in Attachment 18, Critical Organ and Inhalation Dose Factors For Surry RIivy, Riipv = the inhalation pathway dose factor for ventilation vents or process vent release due to 1131, 1133, tritium, and from all particulate-form radionuclides with half-lives greater than 8 days, in mrem /yr per Curie /sec. Factors are listed in Attachment 18
- d. dip = the release for ventilation vents or process vent of I131, 1133, tritium, and from all particulate-form radionuclides with half-lives greater than 8 days in Curies 3.17 E-08 = the inverse of the number of seconds in a year O
_ _ _ _ . - . - . ~ . - - . - . _ - - - . , , _ - _ - _ _
l VIRGINIA VPAP-2103 i POWER REVISION 7 PAGE 35 OF 156 O 6.3.4 I-131,I-133, H-3, and Radionuclides In Particulate Form Emuent Dose Limit (continued)
- d. North Anna Dese Calculations Gaseous releases, not through the process vent, are considered ground level and shall be included in the determination of dm. Historical data pertaining to the
. volumes and radioactive concentrations of gaseous effluents released in connection to specific Station functions, such as containment purges, shall be used in the estimates as appropriate.
- 1. The dose to the maximum exposed member of the public, attributable to gaseous effluents at and beyond the site boundary, that contain 1131, Il33, tritium, and particulate-form radionuclides with half-lives greater than 8 days, shall be determined by:
'r = 3.17E-08 [RM;yydiyy + W;pyd;py. (21)
' i Where: l O Subscripts = vv, refers to vent releases from the building ventilation vents; l py, refers to the vent releases from the process vent Dr = the dose to the critical organ of the maximum exposed member of the public,in mrem l RMivy, RMipv =the cow-milk dose factor for ventilation vents or process vent !
release due to 1131, Il33, tritium, and from all particulate-form radionuclides with half-lives greater than 8 days,in mrem /yr per Curie /sec. Factors are listed in Attachment 19, Critical Organ Dose Factors for North Anna d.d,, = the release for ventilation vents or process vent of 1131,;133, tritium, and from all particulate-form radionuclides with half-lives greater than 8 days,in Curies 3.17 E-08 = the inverse of the number of seconds in a year O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 36 OF 156 6.3.5 Gaseous Radwaste Treatment 0 ,
1 Historical data pertaining to the volumes and radioactive concentrations of gaseous "
I effluents released in connection with specific Station functions, such as containment l I
purges, shall be used to calculate projected doses, as appropriate.
- a. Requirement j
~ ' 1. 'The Gaseous Radwaste Treatment System and the Ventilation Exhaust l Treatment System shall be used to reduce radioactive materialin gaseous waste before its discharge, when projected gaseous effluent air doses due to gaseous effluent releases, from each unit to areas at and beyond the site boundary, would l exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation, averaged i over 31 days. (North Anna)
- 2. Appropriate portions of the Gaseous Radwaste Treatment System shall be used to reduce radioactive materials in gaseous waste before its discharge, when the projected gaseous effluent air doses due to gaseous effluent releases, from each !
unit to areas at and beyond the site boundary, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation, averaged over 31 days. (Surry) g
- 3. The Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste before its discharge, when the projected doses due to gaseous effluent releases, from each unit to areas at and beyond the site boundary, would exceed 0.3 mrem to the critical organ, averaged over 31 days.
- 4. Doses due to gaseous releases from the site shall be projected at least once per 31 days, based on the calculations in 6.3.5.c., and 6.3.5.d.
- b. Action if gaseous waste that exceeds the limits in 6.3.5.a. is discharged without treatment, prepare and submit to the NRC within 30 days, a special report in accordance with VPAP-2802, Notifications and Reports, that includes:
l 1. An explanation why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.
- 2. Actions taken to restore the inoperable equipment to operable status.
l
- 3. Summary description of actions taken to prevent recurrence. h l
VIRGINIA VPAP-2103 POWER '
REVISION 7 PAGE 37 OF 156 6.3.5 Gaseous Radwaste Treatment (continued)
- c. Projected Gamma Dose
- 1. Determine D ,E the 31-day gamma air dose for the previous 31-day period, per Equation (18).
- 2. Estimate R ,g the ratio of the estimated volume of gaseous effluent in the current
~~
31-day period to the volume released during the previous 31-day period.
- 3. Estimate F ,gthe ratio of the estimated noble gas effluent activity in the current 31-day period to the noble gas effluent activity during the previous 31-day period (pCi/ml).
- 4. Determine PD ,s the projected 31-day gamma air dose.
PD 8
= D8 (R8 x F8 ) (22)
- d. Projected Beta Dose
- 1. Determine Db, the 31-day beta air dose in the previous 31 days, per Equation (19).
- 2. Estimate R 8and F as 8 in 6.3.5.c.2. and 6.3.5.c.3.
- 3. Determine PD ,b the projected 31-day beta air dose.
PD b = Db (Rg x F8 ) (23)
- e. Projected Maximum Exposed Member of the Public Dose
- 1. Determine Dmax, the 31-day maximum exposed member of the public dose in the previous 31-day period, per Equation (20) or Equation (21), where Dr = 1 Dmax.
- 2. Estimate F i , the ratio of the estimated activity from I133, 1133, radioactive l materials in particulate form with half-lives greater than 8 days, and tritium in l the current 31-day period to the activity of 1133, 1133, radioactive materials in l
particulate form with half-lives greater than 8 days, and tritium in the previous l 31-day period ( Ci/ml).
b lO
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2 VIRGINIA VPAP-2103 !
POWER REVISION 7 PAGE 38 OF 156
- 3. Determine PDmax, the projected 31-day maximum exposed member of the 9
public dose. ,
PD max =Dmax(Rg x F.i ) (24) 6,4 Radioactive Liquid and Gaseous Release Permits RP shall' maintain procedures for Liquid and Gaseous Release Permits to ensure effluent dose limits are not exceeded when making releases.
6.4.1 Liquid Waste Batch Release Permits Operations shall obtain PS authorization before initiating batch releases of radioactive liquids. Examples of batch releases include:
- a. Surry Batch Releases Release of contents from the following tanks / sumps other than transfers to the Surry Radwaste Facility shali liave a Liquid Waste Batch Release Permit before the discharge:
Boron Recovery Test Tank (BRTT)
. Low Level Waste Drain Tank (LLWDT)
- . High Level Waste Drain Tank (HLWDT)
. Liquid Waste Test Tank (LWTD Contaminated Drain Tank (CDT)
. Laundry Drain Surge Tank (LDST)
. Turbine Building Sumps when RP determines that source activity requires placing pumps in manual mode
. Condensate Polishing Building Sumps when RP determines th'e presence of contamination from primary-to-secondary leakage O
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I VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 39 OF 156 !
lO 6.4.1 Liquid Waste Batch Release Permits (continued)
- b. North Anna Batch Releases NOTE: If the clarifier is in service, releases from tanks processed through the clarifier are l
considered continuous releases.
i A Batch Release Permit is required for a release from any tanks / sumps which contain (or potentially contain) radioactive liquid. Tanks / sumps include:
.BRTT l
. LLWDT
, .HLWDT l
l . Turbine Building Sumps when secondary coolant activity exceeds 1.0 E-5 Ci/ml 1 CDT :
l 6.4.2 Continuous Release Permit Operations shall obtain RP authorization before initiating continuous releases of
[ radioactive liquids.
i a. Surry Continuous Releases A Continuous release permit is required at Surry for:
. Steam generator blowdown l . Component Cooling Water (CCW) heat exchanger to service water leakage, if 1
( applicable
! . Turbine Building sumps and/or subsurface drains if source activity concentrations i are sufficiently low to allow continuous release
- b. North Anna Continuous Releases l
A Continuous Release Permit is required at North Anna for:
. Clarifier, unless being bypassed
. Steam generator blowdown when clarifier is bypassed
. Containment mat sumps and service water reservoir when clarifier is bypassed 6.4.3 Waste Gas Decay Tank (WGDT) Release Permit Operations shall obtain RP authorization before initiating WGDT releases.
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l VIRGINIA VPAP-2103 j POWER REVlSION 7 PAGE 4G OF 156 6.4.4 Reactor Containment Release Permits 0
Operations shall obtain authorization .'ron. RP before initiating containment purges or containment hogging. Reactor Comainment Release Permits shall be valid from start of purge / hog until: .
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. Routine termination
. Terminated for cause by RP l
. Receipt of Radiation Monitoring System (RMS) Containment Gas Monitor high alarm l
6.4.5 Miscellaneous Gaseous Release Permit Operations shall obtain RP authorization before initiating releases of noble gases that may not be accounted for by routine sampling, or any planned release not being routed through the Process Vent or Ventilation Vents (e.g., steam driven auxiliary feedwater pump testing if primary to secondary leakage exists).
6.4.6 Radioactive Liquid and Gaseous Release Controls
appropriate release permit. Operations shall provide the necessary information to complete the required release permit.
- b. A representative sample shall be obtained of the source to be released.
- 1. Operations shall provide RP with liquid samples and sample information (e.g.,
time of sample) for samples obtained outside the Primaty Sample Room, except Clarifier Proportional Tank and Clarifier Grab Samples at North Anna.
- 2. Chemistry shall provide RP with liquid samples and sample information for samples obtained from inside the Primary Sample Room.
- 3. RP shall obtain gaseous samples.
- c. RP shall perform required sample analyses.
- d. RP shall calculate and record the following information on a release permit:
. Maximum authorized release rate
. Maximum authorized release rate in percentage of limits specified by the ODCM
. Applicable conditions or controls pertaining to the release
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1 VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 41 OF 156 O 6.4.6 Radioactive Liquid and Gaseous Release Controls (continued) e.~ RP shall notify the Shift Supervisor ifit is determined that a release may not be within the effluent dose limits.
- f. Upon receipt of a release permit from RP, Operations shall:
- 1. Verify the conect source is authorized for release.
- 2. Note maximum authorized release rate.
- 3. Note percent of Technical Specification limits the release represents.
- 4. Note and ensure compliance with any indicated controls or conditions applicable to the release. i l
- g. When commencing release, Operations shall provide RP with required information.
As appropriate, required information shall include:
. Date and time release was started
. Starting tank / sump level
. Beginning pressure
. Release flow rate
. Dilution water flow rate
- h. Upon terminating the release, Operations shall return the permit to RP and provide information necessary for completion of permit. As appropriate, required information shallinclude:
. Date and time release was stopped
. Tank / sump ending level
. Release flow rate just prior to termination I
. Ending pressure
. Volume released 6.5 Total Dose Limit to Public From Uranium Fuel Cycle Sources 6.5.1 Requirement The annual (calendar year) dose or dose commitment to a real individual due to releases of radioactivity and radiation from uranium fuel cycle sources shall not i exceed 25 mrem to the total body or the critical organ (except the thyroid, which shall not exceed 75 mrem).
l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 42 OF 156 O;
6.5.2 Action l
- a. If the ca'culated doses from release of radioactive materials in liquid or gaseous effluents exceed twice the limits in 6.2.3.a.,6.3.3.a., or 6.3.4.a., calculate ;
(including direct radiation contribution from the units and from outside storage tanks) whether limits in 6.5.1 have been exceeded.
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- b. If the limits in 6.5.1 have been exceeded, prepare and submit to the NRC within 30 l days, a special report in accordance with VPAP-2802, Notifications and Reports, t
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! that defines the corrective action to be taken to reduce subsequent releases and to i 1
prevent recurrence, and includes a schedule for achieving conformance with the I limits. Special reports, as defined in 10 CFR 20.2203(a)(4), shall include:
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- 1. An analysis that estimates the radiation exposure (dose) to a real individual from uranium fuel cycle sources, including all effluent pathways and direct ,
I radiation, for the calendar year that includes the releases covered by the report.
- 2. A description of the levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. l
- 3. If the estimated dose exceeds the limits in 6.5.1, and if die release condition that violates 40 CFR 190 has not already been corrected, the special report shall include a request for a variance in accordance with the provisions of v
40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. l i
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 43 OF 156 O 6.6 Radiological Environmental Monitoring 6.6.1 Monitoring Program
- a. Requirement
- 1. The Radiological Environmental Monitoring Program shall be conducted as a
specified in Attachments 20 or 21, Radiological Environmental Monitoring Program.
- 2. Samples shall be collected from specific locations specified in Attachment 22 or 23, Environmental Sample Locations. [ Commitment 3.2.2]
- 3. Samples shall be analyzed in accordance with:
- Attachment 20 or 21 requirements
. Detection capabilities required by Attachment 24 or 25, Detection
- Capabilities for Environmental Sample Analysis
- Guidance of the Radiological Assessment Branch Technical Position on Environmental Monitoring dated November,1979, Revision No. I
- b. Action O
Q l. If the Radiological Environmental Monitoring Program is not being conducted as required in 6.6.1.a., report the situation in accordance with VPAP-2802, Notifications and Reports, by preparing and submitting to the NRC. a the Annual Radiological Environmental Operating Report required by i hnical Specification (Surry Technical Specification 6.6.B.2 and North Anna'l .shnical Specification 6.9.1.8), a description of the reasons for not conducting the program as required, and the plan for precluding recurrence.
- 2. If, when averaged over any calendar quarter, radioactivity exceeds the reporting levels of Attachment 26 or 27, Reporting Levels for Radioactivity Concentrations in Environmental Samples, prepare and submit to the NRC within 30 days, a special report in accordance with VPAP-2802, Notifications and Reports, that:
. Identifies the causes for exceeding the limits, and
. Defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the J calendar year limits of 6.2.3,6.3.3, and 6.3.4
l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 44 OF 156 When more than one of the radionuclides listed in Attichment 26 or 27 are O l l
detected in the sampling medium, the report shall be submitted if: -
concentration (1) concentration (2)
(25) reporting level (2) + .. 21.0 n: porting level (1) l
- 3. When radionuclides other than those listed in Attachments 26 and 27 are detected and are the result of plant effluents, the report shall be submitted if the potential annual dose to a member of the public is equal to or greater than the calendar year limits of 6.2.3,6.3.3, and 6.3.4. The report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, report and describe the condition in the Annual Radiological Environmental Operating Report in accordance with VPAP-2802, Notifications and Reports.
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- 4. If milk or fresh leafy vegetable samples are unavailable from one or more of the sample locations required by Attachment 20 or 21, identify locations for obtaining replacemer.t samples and add them to the radiological environmental l monitoring program within 30 days. The specific locations from which samples l were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new locations for obtaining replacement samplesin the next Annual Radioactive Effluent Release Report in accordance with VPAP-2802, Notifications and Reports. Include in the report a revised figure and table for the ODCM to reflect the new locations.
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l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 45 OF 156 6.6.2 Land Use Census .
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! a. Requirement l A land use census shall be conducted and shall identify within a distance of 8 km l (5 miles) the location in each of the 16 meteorological sectors of the following: J l
. Nearest milk animal I
. Nearest residence
- Nearest garden greater than 50 m2 (500 ft2) that produces broad leaf vegetation l 1. The land use census shall be conducted during the growing season, at least once l per 12 months, using methods that will provide the best results (e.g., door-to-door survey, aerial survey, local agriculture authorities). Land use census results l shall be included in the Annual Radiological Environmental Operating Report )
in accordance with VPAP-2802, Notifications and Reports.
- 2. In lieu of the garden census, road leaf vegetation sampling of at least three different kinds of vegetation may be performed at the site boundary in each of two different direction sectors with the highest predicted ground deposition (D/Qs). Specifications for broad leaf vegetation sampling in Attachment 20 or 21 shall be followed, including analysis of control samples.
- b. Action
- 1. If a land use census identifies locations that yield a calculated dose or dose commitment greater than the values currently being calculated in 6.3.4.a.2, I identify the new locations in the next Annual Radioactive Effluent Release Report in accordance with VPAP-2802, Notifications and Reports.
l 2. If a land use census identifies locations that yield a calculated dose or dose l
commitment (via the same exposure pathway) 20 percent (Surry) or 25 percent (North Anna) greater than at a location from which samples are currently being obtained, add the new locations to the Radiological Environmental Monitoring Program within 30 days. Sampling locations, excluding the control station l
location, that have the lowest calculated dose or dose commitments (via the same exposure pathway) may be deleted from the monitoring program. Identify new locations in the next Annual Radioactive Effluent Release Report and include in the report revised figures and tables reflecting the new locations in accordance with VPAP-2802, Notifications and Reports. [ Commitment 3.2.4]
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! VIRGINIA VPAP-2103 l POWER REVISION 7 PAGE 46 OF 156 6.6.3 Interlaboratory Comparison Program -
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- a. Requirement Radioactive materials (which contain nuclides produced at the Stations), supplied as part of an Interlaboratory Comparison Program that has been approved by the NRC, shall be analyzed.
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- b. Action
! 1. Analyses shall be pe formed as part of the Environmental Protection Agency's Environmental Radioactivity Laboratory Intercomparison Studies (Cross l Check) Program and include Program Cross-Check of Milk 1131, Gamma, K, Sr89 and Sr 90 f l
Water Gross Beta, Gamma,I 131 H3 (Tritium),Sr89 and Sr90(blind-any combinations of above radionuclides) l l Air Filter Gross Beta, Gamma, Sr90 l
- 2. If analyses are not performed as required by 6.6.3.b., report in the Annual Radiological Environmental Operating Report in accordance with VPAP-2802, ,
1 l Notifications and Reports, the corrective actions taken to prevent recurrence. i l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 47 OF 156 6.6.3 Interlaboratory Comparison Program (continued)
- c. Methodology and Results
- 1. Methodology and results of the cross-check program shall be maintained in the contractor-supplied Nuclear Reactor Environmental Radiation Monitoring Quality Control Manual, IWL-0032-361.
- 2. Results shall be reported in the Annual Radiological Environmental Monitoring Report in accordance with VPAP-2802, Notifications and Reports.
6.7 Reporting Requirements 6.7.1 Annual Radiological Environmental Operating Report Routine Radiological Environmental Operating Reports covering the operation of the units during the previous calendar year shall be submitted prior to May 1 of each year.
A single submittal may be made for the Station. Radiological Environmental Operating Reports shallinclude: I
- a. Summaries, interpretations, and analysis of trends of results of radiological g environmental surveillance activities for the report period, including:
J 5._) . A comparison (as appropriate) with preoperational studies, operational controls, and previous environmental surveillance reports
. An assessment of the observed impacts of the plant operation on the environment
- Results of land use census per 6.6.2
- b. Results of analysis of radiological environmental samples and of environmental radiation measurements taken per 6.6.1, Monitoring Program. Results shall be summarized and tabulated in the format of the table in the Radiological Assessment Branch Technical Position on Environmental Monitoring.
- 1. If some individual results are not available for inclusion with the report, the report shall be submitted, noting and explaining reasons for missing results.
- 2. Missing data shall be submitted in a supplementary report as soon as possible.
- c. A summary description of the radiological environmental monitoring program.
- d. At least two legible maps covering sampling locations, keyed to a table giving distances and directions from the centerline of one reactor. One map shall cover stations near the site boundary; a second shall include more distant stations.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 48 OF 156
- e. Results of Station participation in the Interlaboratory Comparison Program, per 6.6.3.
- f. Discussion of deviations from the Station's environmental sampling schedule per Attachment 20 or 21.
- g. Discussion of analyses in which the lower limit of detection (LLD) required by Attachment 24 or 25 was not achievable.
6.7.2 Annual Radioactive Effluent Release Report
- a. Requirement Radioactive Effluent Release Reports covering operation of the units during the previous 12 months of operation shall be submitted before May 1 of each year. A single submittal may be made for the Station and should combine those sections that are common to both units. Radioactive Effluent Release Reports shall include:
- 1. A summary of quantities of radioactive liquid and gaseous effluents and solid ,
waste released. Data shall be summarized on a quarterly basis following the j format of Regulatory Guide 1.21, Appendix B. l
- 2. An assessment of radiation doses to the maximum exposed members of the @1 public due to the radioactive liquid and gaseous effluents released from the Station during the previous calendar year. This assessment shall be in accordance with 6.7.2.b.
- 3. A list and description of unplanned releases from the site to' unrestricted areas, during the reporting period, which meet the following criteria:
. Unplanned releases that exceeded the limits in 6.2.1 and 6.3.1 l l
. Unplanned releases which require a Deviation Report and involve the l discharge of contents of the wrong Waste Gas Decay Tank or the wrong liquid radwaste release tank
. Unplanned releases from large leaks due to unexpected valve or pipe failures l that result in a quantity of release such that a 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors or 10 CFR 50.73, Licensee Event Report System, report is required
. Unplanned releases as determined by Radiation Protection Supervision, l which may or may not require a Deviation Report
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 49 OF 156 6.7.2 Annual Radioactive Emuent Release Report (continued)
- 4. Major changes to radioactive liquid, gaseous, and solid waste treatment systems during the reporting period.
- 5. Changes to VPAP-2103, Offsite Dose Calculation Manual (see 6.7.4).
- 6. A listing of new locations for dose calculations or environmental monitoring identified by the land use census (see 6.6.2).
- b. Dose Assessment
- 1. Radiation dose to individuals due to radioactive liquid and gaseous effluents
. from the Station during the previous calendar year shall either be calculated in accordance with this procedure or in accordance with Regulatory Guide 1.109.
Population doses shall not be included in dose assessments.
- 2. The dose to the maximum exposed member of the public due to radioactive ,
liquid and gaseous effluents from the Station shall be incorporated with the dose assessment performed above. If the dose to the maximum exposed member of the public exceeds twice the limits of 6.2.3.a.1,6.2.3.a.2,6.3.3.a.1, or 6.3.4.a.1, the dose assessment shall include the contribution from direct radiation.
NOTE: NUREG-0543 states: "There is reasonable assurance that sites with up to four operating reactors that have releases within Appendix I design objective values are also in conformance with the EPA Uranium Fuel Cycle Standard,40 CFR Part 190."
- 3. Meteorological conditions during the previous calendar year or historical annual average atmospheric dispersion conditions shall be used to determine gaseous pathway doses.
6.7.3 Annual Meteorological Data l
- a. Meteorological data collected during the previous year shall be in the form ofjoint frequency distributions of wind speed, wind direction, and atmospheric stability. i
- b. Meteorological data shall be retained in a file on site and shall be made available to NRC upon request.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 50 OF 156 6.7.4 Changes to the ODCM 0
Changes to the ODCM shall be:
- a. Reviewed and approved by SNSOC and the Station Manager before implementation.
- b. Documented. Records of reviews shall be retained as Station records.
Documentation shall include:
- 1. Sufficient information to support changes, together with appropriate analyses or evaluations justifying changes.
- 2. A determination that a change will not adversely impact the accuracy or reliability of effluent doses or setpoint calculations, and will maintain the level of radioactive effluent control required by:
10 CFR 20 Subpart D 40 CFR 190 10 CFR 50.36a 10 CFR 50, Appendix I
- c. Submitted to NRC in the form of a complete, legible copy of the entire ODCM as a part of, or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
- d. Submitted to the Management Safety Review Committee (MSRC) Coordinator.
[ Commitment 3.2.1]
- e. Submitted to NRC in accordance with VPAP-2802, Notifications and Reports.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 51 OF 156 O 7.0 RECORDS 1
7.1 The following individual and packaged documents and copies of any related correspondence completed as a result of the performance or implementation of this procedure are records. They shall be submitted to Records Management in accordance with VPAP-1701, Records Management. Prior to transmittal to Records Management, the sender shall assure that:
. Each record is packaged when applicable,
. QA program requirements have been fulfilled for Quality Assurance records,
. Each record is legible, completely filled out, and adequately identifiable to the item or activity involved,
. Each record is stamped, initialed, signed, or otherwise authenticated and dated, as required by this procedure.
7.1.1 Individual Records
. None 7.1.2 Record Packages f- s
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- Records of changes to the ODCM in accordance with 6.7.4
. Records of meteorological data in accordance with 6.7.3
. Records of sampling and analyses
. Records of radioactive materials and other effluents released to the environment
. Records of preventive ma'mtenance, surveillances, and calibrations 7.2 The following documents completed as a result of the implementation of this procedure are not records and are not required to be transmitted to Records Management.
None O
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} VIRGINIA VPAP-2103 i POWER REVISION 7 i PAGE 52 OF 156 O
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VIRGINIA VPAP-2103 POWER REVISION 7 i PAGE 53 OF 156 !
O ATTACHMENT 1 i
I (Page 1 of 1)
Surry Radioactive Liquid Emuent Monitoring Instrument /. tion l
1 1 Instrument Minimtim Action i Operabfe !
Channels !
- 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM i AND AUTOMATIC TERMINATION OF RELEASE l (a) Radwaste Facility Liquid Effluent Line l RM-RRM-131 } }
- 2. GROSS BETA OR GAMMA RADIOACTIVITY MONITOR $ 4 PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC l 1
TERMINATION OF RELEASE (a) Circulating Water Discharge Line Unit 1: 1 SW-RM-120 2 2 i I
Unit 2: 2-SW-RM-220 (b) Component CoolinE Service Water Effluent Line V l-SW-RM-107A 4 2 1-SW-RM 107B l-SW-RM-107C l l-SW-RM-107D I
- 3. FLOW RATE MEASUREMENT DEVICES !
Radwaste Facility Liquid Effluent Line Instrument Loop RLW-153 1 3 ACTION 1: If the number of operable channels is less than required, effluent releases shall be suspended. l ACTION 2: If the number of operable channels is less than required, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are co2ected and analyzed for principal gamma emitters, as defined in Attachment 8, Surry Radioactive Liquid Waste Samplir.g and Analysis Program.
ACTION 3: If the number of operable channels is less than required, effluent releases via this pathway shall be suspended.
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,u -,n-e. 4a,a,- n- ,A1.s a s, wuea- ,L,z e --nm- 2-m, -a -n--ma,J---,2---a u an-1. >ss- A _ a A-- 2 --a.-mam. - . VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 54 OF 156 9
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 55 OF 156 b1 V A'ITACHMENT 2 (Page 1 of 2)
North Anna Radioactive Liquid Emuent Monitoring Instrumentation Minimum Instrument Operable Action Channels _
- 1. Liquid Radwaste Emuent (a) 1-RM-LW ill, Liquid Radwaste Effluent Monitor . I 1 (b) 1 LW-FT-104, Liquid Radwaste Effluent Total Flow Measuring } 9 Device (c) 1-LW-SOV-121, Clarifier Effluent Line Continuous Composite } }
Sampler and Sampler Flow Monitor (d) 1-LW-TK-20, Liquid Waste Effluent Sample Vessel 1 1 (e) 1 LW 1130, Liquid Waste Efiluent Proportional Sample Valve 1 1 (f) 1 RM-SW-108, Service Water Effluent Monitor 1 1 (g) 1-RM-SW 130, Unit 1 Circulating Water System Effluent Line i 4 Monitor
.O (h) 2-RM-SW-230, Unit 2 Circulating Water System Effluent Line } 4 V Monitor
- 2. Tank Level Indicating Devices (Note 1)
(a) Refueling Water Storage Tanks Unit I l-QS-LT-100A 1 3 1-QS-LT-100B l-QS-LT-100C 1-QS-LT-100D Unit 2 2-QS-LT-200A 1 3 2-QS-LT-200B 2-QS-LT 200C 2-QS-LT-200D (b) Casing Cooling Storage Tanks Unit I l-RS LT-103A 1 3 1 RS-LT 103B Unit 2 2-RS-LT 203A 1 3 2-RS-LT-203B (c) PG Water Storage Tanks (Note 2) 1 BR-LT-il6A(1 PG-TK 1A) 1 3 1-BR-LT-ll6B (1-PG-TK 1B) 1 3 (d) Boron Recovery Test Tanks (Note 2) l-BR-LT-112A (1 BR TK-2A) ] 3 O 1 BR-LT-il2B (1 BR TK 2B) 1 3
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 56 OF 156 ATTACIIMENT 2 (Page 2 of 2)
North Anna Radioactive Liquid Emuent Monitoring Instrumentation ACTION 1: If the number of operable channels is less than required, effluent releases via this pathway may continue if, at least once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta and gamma) at an LLD of at least lx10-7 Ci/g or an isotopic radioactivity at an LLD of at least 5x10-7 pCi/g.
ACTION 2: If the number of operable channels is less than required, effluent releases via this pathway may continue if the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Design capacity performance curves generated in situ may be used to estimate flow.
ACTION 3: If the number of operable channels is less than required, liquid additions to this tank may continue if the tank liquid level is estimated during all liquid additions to the tank.
ACTION 4: If the number of op able channels is less than required, make repairs as soon as possible. Grab .camples cannot be obtained via this pathway.
NOTE 1: Tanks included in his requirement are those outdoor tanks that are not sanounded by liners, dikes, or waJs capable of holding the tank contents, and do not have overflows and surrounding area drain; connected to the liquid radwaste treatment system.
NOTE 2: This is a shared system between Unit I and Unit 2.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 57 OF 156 ATTACHMENT 3 (Page1of1)
Surry Radioactive Liquid Emuent Monitoring Instrumentation Surveillance Requirements l
1 Channel Description Channel Source Channel Channel l Check Check Calibration Functional Test
- 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC
'IERMINATION OF RELEASE (a) Radwaste Facility Liquid Effluent Line 1 l
l RM-RRM 131 o p g g i
- 2. GROSS BETA OR GAMMA RADIOACTIV-ITY MONITORS PROVIDING ALARM BUT l NOT PROVIDING AUIOMATIC TERMI-NATION OF RELEASE (a) Circulating Water Discharge Line Unit 1: 1-SW-RM-120 D M R Q Unit 2: 2-SW-RM-220 (b) Component Cooling Service Water Efflu-
! ent Line 1-SW-RM 107A D M R Q 1-SW-RM-107B 1-SW-RM-107C 1-SW RM-107D l
- 3. FLOW RATE MEASUREMENT DEVICES Radwaste Facility Liquid Effluent Line Instrument Loop RLW-153 DR N/A R N/A l l
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- a. a mJ A. -,E a. e eam-,-- 2A e ,,-s-- a--_s4 &_ _ _a,,m,,B n.&5-m-am-a.--,-
- A=. ,Anun,,mm a_ammasu,,D-e-1m-a e.4 A,am-4= mm+ a --,nmmn,m----,a mm-.an,s-4,, m- , _ _
VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 58 OF 156 !
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 59 OF 156 ATTACHMENT 4 (Page 1 of 2)
North Anna Radioactive Liquid Emuent Monitoring Instrumentation Surveillance Requirements Channel Description Channel Source Channel Channel Check Check Calibration Functional '
Test
- l. Liquid Radwaste Emuent (a) 1-RM LW 111.LiquidRadwasteEffluent D D R Q (NOTE 1)
Monitor (b) 1-LW-FT-104, Liquid Radwaste Effluent D (NOTE 3) N/A R Q Total Flow Measuring Device (c) 1-LW-SOV-121 Clarifier Effluent Line N/A N/A R N/A Continuous Composite Sampler and Sampler Flow Monitor (d) 1 LW-TK-20, Liquid Waste Effluent D (NOTE 9) N/A N/A N/A Sample Vessel (e) 1-LW-1130, Liquid Waste Effluent D (NOTE 9) N/A N/A N/A Proportional Sample Valve (f) 1 RM-SW-108, Service Water System D M R Q (NOTE 2)
Effluent Monitor O' (g) 1-RM-SW 130 Unit 1 Circulating Water System Effluent Line Monitor D M R Q (NOTE 2)
(h) 2-RM-SW-230, Unit 2 Circulating Water D M R Q (NOTE 2)
System Effluent Line Monitor
- 2. Tank Level Indicating Device (NOTE 6)
(a) Refueling Water Storage Tanks Unit I l-QS-LT 100A D (NOTE 4) N/A R Q (NOTE 7) 1-QS-LT-100B l-QS-LT-100C 1-QS-LT-100D Unit 2 2-QS-LT-200A D (NOTE 4) N/A R Q (NOTE 7) 2-QS LT 200B 2-OS-LT-200C 2-QS-LT-200D (b) Casing Cooling Storage Tanks Unit I l-RS-LT-103 A D (NOTE 4) N/A R Q (NOTE 7) 1-RS-LT-103B Unit 2 2-RS LT-203A D (NOTE 4) N/A R Q (NOTE 7) 2-RS-LT-203B (c) PO Water Storage Tanks (NOTE 5) 1 BR-LT-116A (1-PO-TK 1 A) D (NOTE 4) N/A R Q (NOTE 8) 1-BR-LT-116B (1-PO-TK 1B) D (NOTE 4) N/A R Q (NOTE 8)
(d) Boron Recovery Test Tanks (NOTE 5) p l BR-LT-112A (1-BR-TK-2A) D (NOTE 4) N/A R Q (NOTE 8) 1 BR-LT-112B (1 BR TK-2B) D (NOTE 4) N/A R Q (NOTE 8) i I
VIRGLNIA VPAP-2103 POWER REVISION 7 PAGE 60 OF 156 l ATTACHMENT 4 O
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North Anna Radioactive Liquid Effluent Monitoring Instrumentation Surveillance l Requirements NOTE 1: The Channel Functional Test shall demonstrate:
l a. Automatic isolation of this pathway and Control Room alarm annunciation occur if the instrument indicates measured levels abovc alarm / trip setpoint.
- b. Alarm annunciation occurs if the instrument controls are not set in " operate" mode.
NOTE 2: The Channel Functional Test shall demonstrate that Control Room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the alarm / trip setpoint.
- b. Instrument controls not set in " operate" mode.
NOTE 3: Channel Check shall consist of verifying indication of flow during periods of release.
Channel Check shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.
NOTE 4: During liquid additions to the tank, verify indication of level change.
NOTE 5: This is a shared system between Unit I and Unit 2.
NOTE 6: Tanks included in this requirement are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have overflows and surrounding area drains connected to the liquid radwaste treatment system.
NOTE 7: The Channel Functional Test shall demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occur if instrument indicates measured levels outside the alarm / trip setpoint. Demonstration of automatic isolation may consist of verifying the appropriate signal is generated. Valves need not be operated for this test.
NOTE 8: The Channel Functional Test shall demonstrate that Control Room alarm annunciation occurs if the instrument indicates measured levels are outside alarm setpoint.
NOTE 9: Channel Check shall consist of verifying that proportional flow exceeds 0.5 mis / gallon.
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, POWER REVISION 7 l PAGE 61 OF 156 I
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v ATTACHMENT 5
! (Page 1 of 1)
Liquid Ingestion Pathway Dose Factors for Surry Stat:on Units I and 2 Total Body Ai Thyroid Ai GI-LLI Radionuclide mrem /hr mrem /hr mrem /hr l
pCi/ml pCi/ml pCi/mi H-3 2.82E-01 2.82E-01 2.82E-01 Na-24 4.57E-01 4.57E-01 4.57E-01 Cr-51 5.58E+00 3.34E-01 1.40E+03 Mn-54 1.35E+03 - 2.16E+M Fe-55 8.23E+03 - 2.03E+04 Fe-59 7.27E+04 - 6.32E+05 Co-58 1.35E+03 -
1.22E+04 Co-60 3.82E+03 -
3.25E+04 Zn-65 2.32E+05 -
3.23E+05 Rb-86 2.91E+02 - 1.23E+02 Sr-89 1.43E+02 - 8.00E+02 Sr-90 3.01E+04 -
3.55E+03 Y-91 2.37E+00 -
4.89E+04 Zr-95 3.46E+00 - 1.62E+04 O' Zr-97 Nb-95 8.13E-02 1.34E+02 5.51E+04 1.51E+06 Mo-99 2.43E+01 - 2.%E+02 Ru-103 4.60E+01 - 1.25E+04 Ru-106 2.01 E+02 -
1.03E+05 Ag-110m 8.60E+02 - 5.97E+05 Sb-124 1.09E+02 6.70E-01 7.84E+03 Sb-125 4.20E+01 1.79E-01 1.94E+03 Te-125m 2.91E+01 6.52E+01 8.66E+02 Te 127m 6.68E+01 1.40E+02 1.84E+03 Te-129m 1.47E+02 3.20E+02 4.69E+03 Te-131m 5.71E+01 1.08E+02 6.80E+03 Te-132 1.24E+02 1.46E+02 6.-24E+03 I-131 1.79E+02 1.02E+05 8.23E+01 1-132 9.%E+00 9.%E+02 5.35E+00 1133 3.95E+01 1.90E+04 1.16E+02 I-134 5.40E+00 2.62E+02 1.32E-02 l-135 2.24E+01 4.01E+03 6.87E+01 Cs-134 1.33E+04 - 2.85E+02 Cs-136 2.04 E+03 - 3.21 E+02 Cs-137 7.85E+03 -
2.32E+02 Cs-138 5.94E+00 -
5.12E-05 l
Ba-140 1.08E+02 -
3.38E+03
! La-140 2.10E-01 -
5.83E+04
! Cc 141 2.63E-01 - 8.86E+03 Cc-143 4.94E-02 - 1.67E+04 Os Ce-144 Np 239 9.59E+00 1.91 E-03 ,
6.NE+04 7.1 IE+02
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 63 OF 156 ATTACHMENT 6 (Page 1 of 5)
North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 1.0 Equation (6)
D = t F[f;C;A; (6-1) i where:
D = cumulative dose commitment to the total body or critical organ, from the liquid effluents for the period t,in mrem t = period for which C; and F are averaged for all liquid releases, in hours F = the near field average dilution factor for C; during any liquid effluent release. Defined as the ratio of the average undiluted liquid waste flow during release to the average flow from the Station discharge structure to unrestricted areas fi = the individual dilution multiplication factor to account for increases in concentration of long-lived nuclides due to recirculation, listed on page 5 of this attachment. "fi"is the ratio of the total dilution flow over the effective dilution flow C; = average concentration of radionuclide i, in undiluted liquid effluent during the period t, from any liquid releases,in Ci/ml A; = the site-related ingestion dose commitment factor to the total body or critical organ of an adult for each identified principal gamma and beta emitter listed on page 5 of this attachment,in mrem /hr per pCi/ml A; = 1.14 E+05 (730/Dw+ 21 BF;/D,) DF; (6-2) where:
1.14 E+05 = (1 E+06 pCi/ Ci x 1 E+03 ml/kg)/8760 hr/yr, units conversion factor 730 = adult water consumption, kg/yr, from NUREG-0133 O
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l VIRGINIA VPAP-2103 l POWER REVISION 7 l
PAGE 64 OF 156 ATI'ACHMENT 6 91 l (Page 2 of 5)
North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 Dw = dilution factor from the near field area within one-quarter mile of the release point to the potable water intake for the adult water consumption.
D, includes ti.e dilution contributions from the Lake Anna Dam to -
Doswell (0.73), the WHTF (CijCc), and Lake Anna (Cp/CL). The potable water mixing ratio is cah lated as: ,
1 1/(Cg/C C)(C R/C )g x 0.73 = C C/(C xg 0.73 ) (6-3) where Cc, CL and CRare the respective concentrations for the considered nuclide in the discharge channel, WHTF (Lagoon) and the Lake.
Calculation is per expressions 11.2-5,11.2-6, and 11.2-8 of the North Anna UFSAR l
21 DFi
=
=
adult fish consumption rate, kg/yr, from NUREG 0133 the bioaccumulation factor for nuclide i, in fish, pCi/kg per pCill, from 9l l l
l Table A-1 of Regulatory Guide 1.109, Rev.1 Da = dilution factor for the fish pathway, calculated as 1/(CL /Cc) where CL and Cc are the concentrations for the considered nuclide in the discharge channel and the WHTF (Lagoon). Calculation is per Expressions 11.2-5, and 11.2-6 of North Anna's UFSAR DFi = the critical organ dose conversion factor for nuclide i, for adults, in mrem /
pCi, from Table E-11 of Regulatory Guide 1.109, Rev.1 l
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VIRGINIA VPAP-2103 l POWER REVISION 7 i PAGE 65 OF 156 ATTACHMENT 6 (Page 3 of 5) l North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 l 2.0 Equation (9)
Equation (6) is simplified for actual dose calculations by introducing:
WASTE FLOW ~
WASTE FLOW .
(6-4) 1 F = CIRC. (WATER) FLOW + WASTE FLOW CIRC. FLOW and !
CIRC. FLOW l (6-5) f i*
1 EFFECTIVE DIL. FLOW. 1 i
Effective dilution flow rates for individual nuclides "i" are listed on Attachment 7, North Anna Liquid Pathway Dose Commitment Factors for Adults. Then the total released activity (Qi) for the considered period and the ith nuclide is written as: l r\
O Q; = t x C; x WASTE FLOW (6-6) and Equation (6) reduces to:
A'.
D=[Q i iEFF. DIL. FLOW. 1 (6-7) l i
For the long lived, dose controlling nuclides, the effective dilution flow is essentially the over (dam) flow rate out of the Lake Anna system (i.e., the liquid pathway dose is practically i
independent from the circulating water flow rate. However, to accurately assess long range average effects of reduced circulating water flow rates during outages or periods oflow lake water temperatures, calculations are based on an average of 7 out of 8 circulating water pumps running at 218,000 gpm = 485.6 cft/sec per pump.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 66 OF 156 l ATTACIIMENT 6 (Page 4 of 5)
North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 NOTE: The 218,000 gpm flow rate per Circulating Water pump is based on Reference 3.1.21.
The choice of seven Circulating Water pumps is considered realistic. Compared to this, the NAPS UFS AR, Chapter 11.2 (Reference 3.1.18), contains an extremely conservative consideration based on the minimum flow in accordance with Reference 3.1.21 with only two Circulating Water pumps operating. Even at such a low flow rate, which cannot be sustained during power generation, liquid pathway effluent dose factors increase only slightly for the dose controlling nuclides (i.e., Cs*
l37 15 percent).
19 percent, Cs By defining B i= Ai / EFF. DIL. FLOW , ithe dose calculation is reduced to a two factor formula:
D= Q; x B; (6-8)
Values for Bi(mrem /Ci) and EFF. DIL. FLOW iare listed in Attachment 7.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 67 OF 156 O
V ATTACHMENT 6 (Page 5 of 5)
North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 Individual Dilution Total Body Ai Liver Ai Radionuclide Multiplication Factor mrem /hr mrem /hr (f i) pCi/mi pCi/ml H3 14.9 6.18 E+00 6.18E+00 Na-24 1.0 3.71E+01 3.71E+01 Cr-51 1.7 1.10E+00 -
Mn-54 7.0 8.62E+02 4.52E+03 Fe-55 113 130E+02 5.56E+02 Fe-59 2.2 9.47E+02 2.47E+03 Co-58 2.8 2.49E+02 1.11E+02 Co-60 13 3 8.27E+02 3.75 E+02 Zn-65 6.1 3.28E+04 7.25E+0*
Rt>86 1.5 3.53E4)4 7.59E+04 Sr-89 23 8 ~ JE+02 -
Sr-90 15.8 239E+05 -
Y 91 2.5 3.42E-01 -
Zr-95 ,
2.7 2.98E-01 4.41E-01 Zr 97 1.0 1.50E-04 3.27E-04 i
\ N b-95 1.9 1.13E+02 2.10E+02 Mo-99 1.0 7.48E+00 3.93E+01 Ru-103 2.0 4.10E+00 -
Ru-106 7.6 2.65E+01 -
Ag.110m 6.2 4.94E+00 832E+00 Sb-124 2.6 437E+01 2.08E+00 l Sb-125 11.4 2.46E+01 1.16E+00 Te-125m 2.5 3.23Ed)2 8.73E+02 Te-127m 3.7 7.82E+02 2.29E+03 ,
Te-129m 1.9 1.52E+03 3.58E+03 Te-131m 1.0 1.i2E+02 135E+02 Te-132 1.0 5.04E+02 537E+02 1131 1.2 9.66E+01 1.69E+02 1132 1.0 1.03E-01 2.95E.01 1 133 1.0 3.47E+00 1.14E+01 1-134 1.0 2.15E-02 6.00E-02 1135 1.0 6.58E-01 1.78E+00 Cs-134 103 5.80E+05 7.09E+05 Cs-136 13 6.01 E+04 8.35E+04 Cs-137 15.8 3.45E+05 5.26E+05 Cs-138 1.0 9.18E-01 1.85E+00 B a-140 13 2.65 E+01 5.08E-01 La 140 1.0 4.47E-03 1.69E-02 Ce-141 1.8 2.14E-02 1.89E-01 w Ce 143 1.0 135E-N 1.22E+00 Ce 144 6.6 1.41E+00 1.10E+01 Np-239 1.0 5.13E-04 931E-N
i VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 68 OF 156 O'
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VIRGINIA VPAP-2103 POWER REVISION ~1 PAGE 69 OF 156 Os L' ATTACHMENT 7 (Page1of1)
North Anna Liquid Pathway Dose Commitment Factors for Adults Bi= A F i / CIRC FLOW = (A / Effective Dilution Flowi) X 9.81E-3 hr ft3 pCi / sec ml Ci
. Effective Dilution Flow Total Body Bi Liver Bi Radionuclide (cft/sec) (mrem /Ci) (mrem /Ci)
H-3 2.28E+02 2.66E-G4 2.66E-04 Na 24 339E+03 1.07E-G4 1.07E-04 Cr-51 1.99E+03 5.445-06 N/A Mn-54 4.88E+02 1.73E-02 9.08E-02 Fe-55 3.01E+02 4.23E-03 1.81E-02 Fe-59 1.57E+03 5.93E-03 1.55E-02 Co-58 1.20E+03 2.04E-03 9.10E-04 Co-60 2.55E+02 3.18E-02 1.44E-02 Zn-65 5.60E+02 5.74E-01 1.27E+00 Rb-86 2.34E+03 1.48E-01 3.18E-01 Sr-89 1.46E+03 5.84E-03 N/A St-90 2.16E+02 1.09E+01 N/A Y-91 1.34E+03 2 50E-06 N/A Zr-95 1.27E+03 2.30E-06 3.40E-06 g Zr-97 339E+03 433E-10 9.46E-10
( Nb-95 1.78E+03 6.24E-04 1.16E-03 Mo-99 330E+03 2 22E-05 1.17E-Ot Ru-103 1.68E+03 2.40E-05 N/A Ru 106 4.48E+02 5.80E-Ot N/A Ag 110m 5.52E+02 8.78E-05 1.48E-04 Sb.124 1.32E+03 3.25E-04 1.55E-05 Sb-125 2.98E+02 8.10E-04 3.80E-05 Te-125m 135E+03 235E-03 6.35E-03 Te 127m 9.16E+02 837E-03 2.46E-02 Te 129m 1.82E+03 8.19E-03 1.93E-02 Te-131m 338E+03 3.27E-04 3.92E-04 Te-132 3.27E+03 1.51E-03 1.61E-03 1.I31 2.94E+03 3.22E-04 5.62E-G8 l132 3.40E+03 2.98E-07 8.51E-07 I-133 3.39E+03 1.00E-05 3.29E-05 -
I134 3.40E+03 6.19E-08 1.73E-07 I-135 3.40E+03 1.90E-06 5.15E-06 Cs.134 3.29E+02 1.73 E+01 2.11E+01 Cs-136 2.62E+03 2.25E-01 3.12E-01 Cs 137 2.15E+02 1.57E+01 2.40E+01 Cs-138 3.40E+03 2.65E-06 534E-06 Ba-140 2.65E+03 9.83E-05 1.hdE-06 La 140 336E+03 131E-08 4.94E-08 Ce-141 1.85 E+03 1.14E-07 1.00E-06 Ce-143 337E+03 3.93E-10 3.55E-06 Ce-144 5.14E+02 2.70E-05 2.10E-04 i
Np-239 332E+03 1.51E-09 2.75E-09
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 71 OF 156 O ATTACHMENT 8 (Page 1 of 3) 1 Surry Radioactive Liquid Waste Sampling and Analysis Program er i to Liquid Release Sampling Minimum Analy- Type of Activity
, 9 D Type Frequency sis Frequency Analysis '
( Ci/ml),(Note 1) p p Principle Gamma 5 x 10'7 Emitters (Note 3) 131 (Each Batch) (Each Batch) 1 1 x 10-6 Dissolved and Batch Releases M Entrained Gases 1x17 5 (One B tch/M)
(Gamma Emitters)
(Note 2) P M Composite H3 1 x 10'3 (Each Batch) (Note 4) Gross Alpha 1 x 10'7 P Q Composite Sr89and Sr90 5 x 10-8 (Each Batch) (Note 4) Fe55 1 x 10-6 Continuous W Composite Pal Gamm 5 x 10'7 t 6 131 1 x 10-6 (Note 6) (Note 6) 1 Dissolved and Continuous M 1 x 10 5 Entrained Gases i Releases Grab Sample (Gamma Emitters) )
l 1 l (Note 5) Continuous M Composite H 3 1 x 10-5 )
(Note 6) (Note 6) Gross .^Jpha 1 x 10-7 Continuous Q Composite Sr89and Sr90 5 x 10-8 ,
I 55 1 x 10-6 (Note 6) (Note 6) Fe
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 72 OF 156 ATTACHMENT 8 (Page 2 of 3)
Surry Radioactive Liquid Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):
4.66 s b
LLD = (8-1)
E = V e 2.22E+06
- Y e e-( AAt)
Where:
LLD = the "a priori" (before the fact) Lower Limit of Detection (as microcuries per unit mass or volume)(see 4.8) sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm)
E = the counting efficiency (as counts per disintegration)
V = the sample size (in units of mass or volume) 2.22E+06 = the number of disintegrations per minute (dpm) per microcurie ;
l Y = the fractional radiochemical yield (when applicable)
A = the radioactive decay constant for the particular radionuclide At = the elapsed time between the midpoint of sample collection and time of counting l Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori" (before the fact) limit representing the capability of a measurement system and not a "posteriori" (after the fact) limit for a particular measurement.
l NOTE 2: A batch release is the discharge of liquid wastes of a discrete volume. Before samt ung for analyses, each batch shall be isolated, and appropriate methods will be used to obttln a representative sample for analysis.
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i VIRGINIA VPAP-2.103 i POWER REVIS'ON 7 PAGE 73 OF 156
% ATTACHMENT 8 ;
(Page 3 of 3)
Surry Radioactive Liquid Waste Sampling and Analysis Program 1 NOTE 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn 54 , pe 59
, Zn65, gg99, Cs I34 , Cs137, Cel41, and
, CoS8, Co 60 l44 . This list does not mean that only these nuclides are to be detected and reported.
Ce Other peaks that are measurable and identifiable, at levels exceeding the LLD, together !
with the above nuclides, shall also be identified and reported.
NOTE 4: A composite sample is one in which the quantity of liquid sampled is proportional to the quantity ofliquid waste discharged and for which the method of sampling employed results in a specimen that is representative of the liquids released.
NOTE 5: A continuous release is the discharge of liquid wastes of a non-discrete volume, e.g., from ,
a volume of a system that has an input flow during the continuous release. l l
NOTE 6: To be representative of the quantities and concentrations of radioactive materials in liquid J effluents, composite sampling shall employ appropriate methods which will result m a p( specimen representative of the effluent release. )
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POWTR REVISION 7 PAGE 74 OF 156 O
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v ATTACHMENT 9 l
j (Page 1 of 3)
North Anna Radioactive Liquid Waste Sampling and Analysis Program l
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Liquid Release Sampling Minimum Analy- Type of Activity Lower Limit of Type Frequency sis Frequency Analysis Detection (LLD)
(pCi/ml),(Note 1)
! P P Principle Gamma 5 x 10'7 Emitters (Note 3) i (Each Batch) (Each Batch) g131 1 x 10-6 Batch Releases P M Dissolved and (One Batch /M) Entrained Gases 1 x 10-5 (Gamma Emitters)
(Notes 2 and 7) P M Composite H3 1 x 10-5 (Each Batch) (Note 4) Gross Alpha 1 x 10~7 i P Q Composite Sr89 and Sr* 5 x 10-8 l (Each Batch) (Note 4) pe 55 1 x 10-6 Principal Gamma 5 x 10-7 Emitters (Note 6) ,
Continuous W Composite 1131 1 x 10~6 Continuous (Note 6) (Note 6) Dissolved and Releases Entrained Gases 1 x 10-5 (Gamma Emitters)
(Note 5) Continuous M Composite H3 1 x 10-5 (Note 6) (Note 6) Gross Alpha 1 x 10~7 Continuous Q Composite Sr89 and Sr* 5 x 10-8 (Note 6) (Note 6) p,55 1 x 10-6 lO W
., e,
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 76 OF 156 ATTACHMENT 9 (Page 2 of 3)
North Anna Radioactive Liquid Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):
4.66 s b LLD = (9,3)
E
- V e 2.22E+06
- Y
- e-( AAt)
Where:
LLD = the "a priori"(before the fact) Lower Limit of Detection as defined above (as microcuries per unit mass or volume) (see 4.8) sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm)
E = the counting efficiency (as counts per disintegration)
V = the sample size (in units of mass or volume) 2.22E+06 = the number of disintegrations per minute (dpm) per microcurie Y = the fractional radiochemical yield (when applicable)
A = the radioactive decay constant for the particular radionuclide At = the elapsed time between the midpoint of sample collection and time of counting Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori"(before the fact) limit representing the capability of a measurement system and not a "posteriori" (after the fact) limit for a particular measurement.
NOTE 2: A batch release is the discharge of liquid wastes of a discrete volume. Before sampling for analyses, each batch shall be isolated, and then thoroughly mixed as the situation permits, to assure representative sampling.
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i VIRGINIA VPAP-2103 i POWER REVISION 7 PAGE 77 OF 156 i I
ATTACHMENT 9 l (Page 3 of 3)
North Anna Radioactive Liquid Waste Sampling and Analysis Program I
NOTE 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn54 pe59 ,Coss, Co60
, Zn65, Mo99, Csl34, Csl37, Cel41, and l44 . This list does not mean that only these nuclides are to be detected and reported.
Ce Other peaks that are measurable and identifiable, at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported.
l NOTE 4: A composite sample is one in which the quantity of liquid sampled is proportional to the quantity ofliquid waste discharged and for which the method of sampling employed results in a specimen that is representative of the liquids released.
NOTE 5: A continuous release is the discharge of liquid wastes of a non-discrete volume, e.g., from l a volume of a system that has an input flow during the continuous release. l NOTE 6: To be representative of the quantities and concentrations of radioactive materials in liquid
- effluents, samples shall be collected continuously in proportion to the rate of flow of the l
v effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent releases.
NOTE 7: Whenever the secondary coolant activity exceeds 10 5 Ci/ml, the turbine building sump l pumps shall be placed in manual operation and samples shall be taken and analyzed prior to release. Secondary coolant activity samples shall be collected and analyzed on a weekly basis. These samples are analyzed for gross activity or gamma isotopic activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Surry Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sampling Minimum Type of Activity Lower Limit of Type Frequency Analysis Analysis Detection (LLD)
Frequency (pCi/ml), (Note 1) rt Releam A. Waste Gas Prior to Release Principal Gamma 4 Storage Tank I 1 x 10 g (Each Tank) Emitters (Note 2) 3 Prior to Release iP d Prior to Release 1 x 10 ers o Purge h P RGE) H3 fGabg )) (Each PURGE) 1 x 10-6 C. Ventilation Weekly Principle Gamma 4 Weekly 1 x 10 (1) Process Vent (Crab Sample) Emitters (Note 2)
(2) Vent Vent #1 (3) Vent vent #2 (Note 3) (Note 3) H3 1 x 10 6 O (4)SRF Vent V
Il31 1 x 1012 Continuous Weekly (Note 5)
(Note 4) (Charcoal Sample) ;133 1 x 1010 Continuous Weekly (Note 5) Principal Gamma 1 x 1011 All Release (Note 4) Particulate Sample Emitter (Note 2)
Weekly Types as listed
."" "8 ~
Composite Gross Alpha 1 x 10~l1 4)
Particulate Sample in A, B, and C Quarterly Continuous Composite Sr 89 and Sr* 1 x 10~11 Particulate Continuous Noble Gas NobleGasesGross 1 x 10'6 (Note 4) Monitor Beta and Gamma Principle Gamma Weekly Weekly 1 x 10 4 Condenser Air Emitters (Note 2)
Grab Sample (Note 3)
Ejector H3 1 x 10-6 (Note 3)
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 80 OF 156 ATTACIIMENT 10 (Page 2 of 5)
Surry Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sampling Minimum Type of Activity Lower Limit of Type Frequency Analysis Analysis Detection (LLD)
Frequency (pCi/ml), (Note 1)
Prior to Release nc P g Gamma 4 Prior to Release 1 x 10 (Grab Sample) (Each Release) H3 1 x 10-6 Continuous Charcoal Sample Containment Continuous Particulate Principal Gamma llog Depres- 1 x 10.to (Note 4) Sample (Note 6) Emitter (Note 2) surization Composite Co .nuous Particulate Gross Alpha 1 x 1010 Sample (Note 6)
Connuous farticul te Sr89and Sr90 1 x 10'10 Sample (Note 6)
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 81 OF 156 ATTACHMENT 10 (hge 3 of 5)
Surry Radioactive Gaseous Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):
l I
4.66 s 0 LLD = (10 1)
E
- V = 2.22E+06
- Y
- e-( AAt)
Where:
LLD = the "a priori"(before the fact) Lower Limit of Detection as defined above (as microcuries per unit mass or volume) (see 4.8).
sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm).
E = the counting efficiency (as counts per disintegration).
V = the sample size (in units of mass or volume).
2.22E+06 = the number of disintegrations per minute (dpm) per microcurie.
Y = the fractional radiochemical yield (when applicable).
1 = the radioactive decay constant for the particular radionuclide.
At = the elapsed time between the midpoint of sample collection and time of counting.
Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori" (before the fact) limit representing the capability of a measurement system and not a "posteriori" (after the fact) limit for a particular measurement.
d
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 82 OF 156 ATTACHMENT 10 ,
(Page 4 of 5)
Surry Radioactive Gaseous Waste Sampling and Analysis Program NOTE 2: The principal gamma emitters for which the LLD specification applies exclusively are the 88 following radionuclides: Kr ,87Kr ,Xel33, Xe133m, Xe135, Xe135m, and Xe138 for gaseous emissions and Mn , Cosa Co60 , Zn65, Mo99, Csl34, Csl37, Cel41 and Ce l44 for 54, pe59 particulate emissions.This list does not mean that only these nuclides are to be detected and reported. Other nuclides with half lives greater than 8 days, that are measurable and identifiable at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported.
NOTE 3: Sampling and analysi:, shall also be performed following shutdown, start-up, and whenever a thermal power change exceeding 15 percent of the rated thermal power occurs within any one-hour period, when:
- a. Analysis shows that the dose equivalent 1 131 concentration in the primary coolant has increased more than a factor of 3; and
- b. The noble gas activity monitor shows that effluent activity has increased by more than a factor of 3. ;
NOTE 4: The ratio of the sample flow rate to the sampled stream flow rate shall be known for the .
period covered by each dose or dose rate calculation made in accordance with 6.3.1, 6.3.3, and 6.3.4.
NOTE 5: Samples shall be changed at least once per seven days and analyses shall be completed !
l within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least seven days following each shutdown, start-up, or thermal power change exceeding 15 percent of rated thermal power in one hour, and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if:
131
- a. Analysis shows that the dose equivalent 1 concentration in the primary coolant has l
increased by a factor of 3; and
- b. Noble gas monitor shows that effluent activity has increased more than a factor of 3.
O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 83 OF 156 ATTACHMENT 10 (Page 5 of 5)
Surry Radioactive Gaseous Waste Sampling and Analysis Program NOTE 6: To be representative of the quandties and concentrations of radioactive materials in gaseous effluents, composite sampling shall employ appropriate methods that will result in a specimen representative of the effiuent release.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 84 OF 156 l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 85 OF 156 ATTACHMENT 11 (Page 1 of 4)
North Anna Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sampling Minimum Type of Activity Lower Limit of Type Frequency Analysis Analysis Detection (LLD)
Frequency (pCi/ml). (Note 1)
A. Waste Gas Prior to Release Principal Gamma 4 ach 'I'a 1 x 10 Storage Tank (EacWad) Entters @ote 2)
Principle Gamma 4 Prior to Release Prior to Release 1 x 10 ,
B. Containment Emitters (Note 2) J Purge I H3 1 x 10 4 G ab Samp e
( ach WRGE)
C. Ventilation Monthly Principle Ganuna d Monthly 1 x 10 (1) Process Vent (Grab Sample) Emitters (Note 2)
(2) Vent Vent A (Notes 3,4, and (Note 3) H3 1 x 10-6 (3) Vent vent B 5)
Continuous Weekly I I31 1 x 10-12 (Note 4) (Charcoal Sample) g133 1 x 10-10 Continuous Weekly Principal Gamma 1 x 10-11 All Release (Note 4) Particulate Sample Emitter (Note 2)
Monthly
"" us Types as listM Composite Gross . Alpha 1 x 10~l1
- 4) Particulate Sample in A, B, and C Quarterly Continuous Composite Sr89and Sr90 1 x 10'11 (Note 4) Particulate Continuous Noble Gas NobleGasesGross 1 x 10 6 (Note 4) Monitor Beta or Gamma Condenser Air Principle Gamma 4 Weekly Weekly 1 x 10 Ejector / Steam Emitters (Note 7)
Generator Grab Sample H3 1 x 10-6 Blowdown Vent (Note 6)
Containment Principle Gamma 4 Prior to Release Prior to each 1 x 10 Vacuum Steam Emitters (Note 2)
Ejector (Hogger) (Grab Sample)
H3 1 x 10-6 Release (Note 8)
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 86 OF 156 ATTACHMENT 11 (Page 2 of 4)
North Anna Radioactive Gaseous Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):
4.66 s D LLD = M E e V e 2.22E+06
- Y e e-(Aat)
Where:
LLD = the "a priori"(before the fact) Lower Limit of Detection as defined above (as microcuries per unit mass or volume) (see Subsection 4.9) sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm)
E = the counting efficiency (as counts per disintegration)
V = the sample size (in units of mass or volume) 2.22E+06 = the number of disintegrations per minute (dpm) per microcurie Y = the fractional radiochemical yield (when applicable) 1 A = the radioactive decay constant for the particular radionuclide 1
At = the elapsed time between the midpoint of sample collection and time of l counting l Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori"(before the fact) limit representing the capability of a !
measurement system and not as "posteriori"(after the fact) limit for a particular l measurement. !
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 87 OF 156 ATTACHMENT 11 l
(Page 3 of 4)
J North Anna Radioactive Gaseous Waste Sampling and Analysis Program l NOTE 2: The principal gamma emitters for which the LLD specification applies exclusively are the 88 followingradionuclides:Kr 87 ,Kr ,Xel33, Xel335, Xel35, Xe135m,and Xe138 for gaseous emissions and Mn 54
, pe 59 s8 , Co
, Co 60 , Zn65
, Mo99, Cs l34 , Csl37, Cel41 and Ce l44 for particulate emissions.This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported.
NOTE 3: Sampling and analysis shall also be performed following shutdown, start-up, and whenever a thermal power change exceeding 15 percent of the rated thermal power occurs within any one-hour period, if:
- a. Analysis shows that the dose equivalent 1 131 concentration in the primary coolant is greater than 1.0 Ci/gm; and g b. The noble gas activity monitor shows that effluent activity has increased by more than l C a factor of 3.
NOTE 4: The ratio of the sample flow rate to the sampled stream flow rate shall be known for the period covered by each dose or dose rate calculation made in accordance with 6.3.1,6.3.3, and 6.3.4.
NOTE 5: Samples shall be changed at least once per seven days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least seven days following each shutdown, start-up or thermal power change exceeding 15 percent of rated thermal power in one hour and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. 'llis requirement applies if:
- a. Analysis shows that the dose equivalent 1 131 concentratica in the primary coolant is greater than 1.0 pCi/gm and;
- b. Noble gas monitor shows that effluent activity has increased more than a factor of 3.
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 88 OF 156 ATTACHMENT 11 (Page 4 of 4)
North Anna Radioactive Gaseous Waste Sampling and Analysis Program NOTE 6: Whenever the secondary coolant activity exceeds 10-5 Ci/ml, samples shall be obtained and analyzed weekly. Secondary coolant activity samples shall be collected and analyzed on a weekly basis. These samples are analyzed for gross activity or gamma isotopic activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NOTE 7: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr ,87Kr88, xe133, Xel335, Xel35, Xe135m, and Xel38 for gaseous emissions. This list does not mean that only these nuclides are to be detected and reported.
Other peaks that are measurable and identifiable, at levels exceeding the LLD together with the above nuclides, shall also be identified and reported.
NOTE 8: If the secondary coolant activity level in any Steam Generator supplying steam to the Hogger exceeds 1.0E-5 pCi/ml, Steam Generator samples shall be obtained and analyzed prior to release.
O O
! l VIRGINIA VPAP-2103 i POWER REVISION 7 l PAGE 89 OF 156 ATTACHMENT 12 (Page 1 of 3)
Gaseous Effluent Dose Factors for Surry l
(Gamma and Beta Dose Factors) y/Q = 6.0E-05 sec/m3 at 499 meters N Direction Dose Factors for Ventilation Vent l
Noble Gas Kw Lw Miw Niw Radionuclide Total Body Skin Gamma Air Beta Air mrem /vr mrem /yr mrad /yr mrad /vr Curie /sec Curie /sec Curie /sec Curie /sec Kr-83m 4.54E+00 -
1.16E+03 1.73E+04 ,
i Kr-85m 7.02E+04 8.76E+04 7.38E+04 1.I8E+05 Kr-85 9.66E+02 8.04E+04 1.03E+03 1.17E+05 Kr-87 3.55E+05 5.84E+05 3.70E+05 6.18E+05 l'
' Q) Kr-88 8.82E+05 1.42E+05 9.12E+05 1.76E+05 l
Kr-89 9.96E+05 6.06E+05 1.04E+06 6.36E+05 Kr-90 9.36E+05 4.37E+05 9.78E+05 4.70E+05 l Xe-131m 5.49E+03 2.86E+04 9.36E+03 6.66E+04 Xe-133m 1.51E+04 5.96E+04 1.96E+04 8.88E+04 Xe-133 1.76E+04 1.84E+04 2.12E+04 6.30E+04 Xe-135m 1.87E+05 4.27E+04 2.02E+05 4.43E+04 l Xe-135 1.09E+05 1.12E+05 1.15E+05 1.48E+05 l l
Xe-137 8.52E+04 7.32E+05 9.06E+04 7.62E+05 l Xe-138 5.30E+05 2.48E+05 5.53E+05 2.85E+05 l Ar-41 5.30E+05 1.61E+05 5.58E+05 1.97E+05 l
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VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 90 OF 156 l ATTACHMENT 12 (Page 2 of 3)
Gaseous Effluent Dose Factors for Surry (Gamma and Beta Dose Factors) l 3 '
x/Q = 1.0E-06 sec/m at 644 meters N Direction Dose Factors for Process Vent Noble Gas K ipv Lipv Mjpy Njpv Radionuclide Total Body Skin Gamma Air Beta Air mrem /yr mrem /vr mrad /yr mrad /vr J
Curie /sec Curie /sec Curie /sec Curie /sec Kr-83m 7.56E-02 -
1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 l Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 l Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 i
Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 0
i VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 91 OF 156 im ATTACHMENT 12 (Page 3 of 3)
Gaseous Effluent Dose Factors for Surry (Inhalation Pathway Dose Factors) 3 Ventilation Vent X/Q = 6.0E-05 sec/m at 499 meters N Direction 3
Process Vent %/Q = 1.0E-06 sec/m at 644 meters S Direction Radionuclide P yy Pipy mrem /vr mrem /vr Curie /sec Curie /sec H-3 6.75E+04 1.12E+03 Cr-51 5.13E+03 8.55E+01 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND
, Rb-86 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-127m 3.64E+05 6.07E+03 Te-129m 3.80E+05 6.33E+03 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND I-131 9.75E+08 1.62E+07 I-133 2.31E+08 3.85E+06 ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1 I
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 92 OF 156 O
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 93 OF 156 l k ATTACHMENT 13 l (Page 1 of 3)
Gaseous Emuent Dose Factors for North Anna l l
(Gamma and Beta Dose Factors) 3 X/Q = 9.3E-06 sec/m at 1416 meters SE Direction Dose Factors for Ventilation Vent Noble Gas Kiyy L yy Miyy N iyy Radionuclide Total Body Skin Gamma Air Beta Air mrem /vr mrem /vr mrad /yr mrad /vr j Curie /sec Curie /sec Curie /sec Curie /sec i Kr-83m 7.03E-01 -
1.79E+02 2.68E+03 l l
Kr-85m 1.09E+04 1.36E+04 1.14E+04 1.83E+04 j Kr-85 1.50E+02 1.25E+04 1.60E+02 1.81E+04 Kr-87 5.51E+04 9.05E+04 5.74E+04 9.58E+04 )
Kr-88 1.37E+05 2.20E+04 1.41E+05 2.72E+04 Kr-89 1.54E+05 9.39E+04 1.61E+05 9.86E+04 Kr-90 1.45E+05 6.78E+04 1.52E+05 7.28E+04 Xe-131m 8.51E+02 4.43E+03 1.45E+03 1.03E+04 Xe-133m 2.33E+03 9.24E+03 3.04E+03 1.38E+04 Xe-133 2.73E+03 2.85E+03 3.28E+03 9.77E+03 ;
Xe-135m 2.90E+04 6.61E+03 3.12E+04 6.87E+03 l
Xe-135 1.68E+04 1.73E+04 1.79E+04 2.29E+04 Xe-137 1.32E+04 1.13E+05 1.40E+04 1.18E+05 Xe-138 8.21E+04 3.84E+04 8.57E+04 4.42E+04 Ar-41 8.22E+04 2.50E+04 8.65E+04 3.05E+04 .
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 94 OF 156 ATTACHMENT 13 (Page 2 of 3)
Gaseous Effluent Dose Factors for North Anna (Gamma and Beta Dose Factors) y/Q = 1.2E-06 sec/m3at 1513 meters S Direction Dose Factors for Process Vent Noble Gas Ki Li M,ip N ipy Radionuclide Total body Skin Gamma Air Beta Air mrem /vr mrem /vr mrad /vr mrad /vr Curie /sec Curie /sec Curie /sec Curie /sec Kr-83m 9.07E-02 -
2.32E+01 3.46E+02 Kr-85m 1.40E+03 1.75E+03 1.48E+03 2.36E+03 Kr-85 1.93E+01 1.61E+03 2.06E+01 2.34E+03 Kr-87 7.10E+03 1.17E+04 7.40E+03 1.24E+04 Kr-88 1.76E+04 2.84E+03 1.82E+04 3.52E+03 Kr-89 1.99E+04 1.21E+04 2.08E+04 1.27E+04 Kr-90 1.97F+04 8.75E+03 1.96E+04 9.40E+03 Xe-131m 1.10E+02 5.71E+02 1.87E+02 1.33E+03 Xe-133m 3.01E+02 1.19E+03 3.92E+02 1.78E+03 i Xe-133 3.53E+02 3.67E+02 4.24E+02 1.26E+03 Xe-135m 3.74E+03 8.53E+02 4.03E+03 8.87E+02 Xe-135 2.17E+03 2.23E+03 2.30E+03 2.95E+03 ,
Xe-137 1.70E+03 1.46E+04 1.81E+03 1.52E+04 Xe-138 1.06E+04 4.96E+03 1.llE+04 5.70E+03 Ar-41 1.06E+04 3.23E+03 1.12E+04 3.94E+03
1 VIRGINIA VPAP-2103 l POWER REVISION 7 PAGE 95 OF 156 l l
(-
'% > ATTACHMENT 13 i (Page 3 of 3)
Gaseous Emuent Dose Factors for North Anna 1
(Inhalation Pathway Dose Factors)
Ventilation Vent y/Q = 9.3E-06 sec/m3at 1416 meters SE Direction 3
l Process Vent x/Q = 1.2E-06 sec/m at 1513 meters S Direction Radionuclide P iyy P ipv mrem /vr mrem /vr Curie /sec Curie /sec H-3 1.05E+04 1.35E+03 Cr-51 7.95E+02 1.02E+02 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND I Rb-86 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-127m 5.64E+04 7.28E+03 Te-129m 5.88E+04 7.59E+03 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND
~
Ce-141 ND ND Cc-144 ND ND 6 I-131 1.51E+08 1.95E+07 I-133 3.58E+07 4.62E+06 ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1
_ _ _ _ _ . . - - _ . - . - - ..-.~. . - - . _ - . . . __ ..
1 i
! VIRGINIA VPAP-2103 i
POWER REVISION 7 PAGE 96 OF 156 k
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 97 OF 156
/O V ATTACHMENT 14 (Page 1 of 3)
Surry Radioactive Gaseous Emuent Monitoring Instrumentation MINIMUM INSTRUMENT OPERABLE ACTION CHANNELS
- 1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release 1-GW-RM-102 1 1 1-GW-RM-130-1 (b) Iodine Sampler Process Vent Continuous HP Sampler, or 1-GW-RM-130-1 1 2 (c) Particulate Sampler Process Vent Continuous HP Sampler, or 1 2 1-GW-RM-130-1 (d) Process Vent Flow Rate Monitor 1-GW-FT-100 1 3 (e) Sampler Flow Rate Measuring Device KAMAN Flow Rate Measuring Device 1 3 (Parameter #19), or HP Sampler Rotometer
- 2. CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor I~ ~
III 2 (one per unit) 1 2-SV-RM-211 (b) Air Ejector Flow Rate Measuring Device Unit 1: 1-VP-F1-1 A 1-VP-FI-1 B 2 (one per unit) 3 Unit 2: 2-VP-F1-1 A 2-VP-FI-l B
- 3. VENTILATION VENT SYSTEM (a) Noble Gas Activity Monitor SRF: RRM-101 1 1 SPS: Vent #1 1-VG-RM-104 1 1 Vent #2,1-VG-RM -110, or 1 1 1-VG-RM-131-1 O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 98 OF 156 l
ATTACHMENT 14 (Page 2 of 3)
Surry Radioactive Gaseous Emuent Monitoring Instrumentation l
l MINIMUM I INSTRUMENT OPERABLE ACTION
! CHANNELS (b) lodine Sampler SRF: RRM-101 1 2 l SPS: Vent #1,1-VG-RM-104 1 2 Vent #2, Continuous HP Sampler, or 1 2 l
l-VG-RM-131-1 (c) Particulate Sampler SRF: RRM-101 1 2 SPS: Vent #1, VG-RM-104 1 2 Vent #2, HP Continuous Sampler, or 1 2 1-VG-RM-131-1 (d) Ventilation Vent Flow Rate Monitor SRF: 01-RHV-FT-156 1 3 SPS: Vent #1,1-VS-FT-119 1 3 Vent #2,1-VS-FT-116 1 3 (e) Sampler Flow Rate Measuring Device l SRF: RRM-101 1 3 SPS: Vent #1,1-VG-RM-104 1 3 Vent #2, KAMAN Flow Rate Measuring Device 1 3 (Parameter #19), or HP Sampler Rotometer i 1
I O1l l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 99 OF 156 ATTACHMENT 14 (Page 3 of 3)
Surry Radioactive Gaseous Emuent Monitoring Instrumentation ACTION 1: If the number of operable channels is less than required, effluent releases via this path may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 2: If the number of operable channels is less than required, effluent releases via the l effected path may continue provided samples are continuously collected within one hour with auxiliary sampling equipment as required in Attachment 10.
ACTION 3: If the number of operable channels is less than required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 101 OF 156
/~N ATTACHMENT 15 (Page 1 of 3)
]
North Anna Radioactive Gaseous Emuent Monitoring Instrumentation i
l l
INSTRUMENT MINIMUM ACTION l OPERABLE CHANNELS
- 1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor 1-RM-GW-102 1 2,4 1-RM-GW-178-1 (b) Iodine Sampler 1-RM-GW-178-1 1 2, 5 Process Vent Continuous HP Sampler (c) Particulate Sampler f 1-RM-GW-178-1 2' 5 Q, Process Vent Continuous HP Sampler 1
(d) Total Flow Monitor 1-GW-FT-108 1 1 (e) Sampler Flow Rate Measuring Device l KAMANS Flow Rate Measuring Device (Parameter 19)
I I HP Sampler Rotameter
- 2. CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor UnitI l-SV-RM-121 1 3 Unit 2 2-SV-RM-221 (b) Flow Rate Measuring Device UnitI l-SV-FI-100A 1-SV-FI-101 A 1 SV-FI-100B 1(NOTE 1) 1 1-SV-FI-101B Unit 2 2-SV-FI-200A 2-SV-FI 201 A 2-SV-FI-200B 1 (NOTE 2) 1 A 2-SV-FI-201B U
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 102 OF 156 ATTACHMENT 15 e '
(Page 2 of 3)
North Anna Radioactive Gaseous Emuent Monitoring Instrumentation MINIMUM INSTRUMENT OPERABLE ACTION CHANNELS
- 3. VENTILATION VENT A (a) Noble Gas Activity Monitor 1-RM-VG-104 1 2 1-RM VG-179-1 (b) Iodine Sampler I l-RM VG-179-1 1 2 Vent Vent A Continuous HP Sampler (c) Particulate Sampler l-RM-VG-179-1 1 2 Vent Vent A Continuous HP Sampler j (d) Total Flow Monitor l 1-HV-FT-1212A 1 1 l (e) Sampler Flow Rate Measuring Device i l KAMANS Flow Rate Measuring Device (Parameter 19) ; 3 l HP Sampler Rotameter l 4. VENTILATION VENT B j (a) Noble Gas Activity Monitor
! l RM-VG-113 2 1
1-RM-VG-180-1 (b) Iodine Sampler l l-RM-VG 180-1 1 2 Vent Vent B Continuous HP Sampler l
l (c) Particulate Sampler 1-RM-VG-180-1 1 2 Vent Vent B Continuous HP Sampler i I
(d) Total Flow Monitor l-HV-FT-1212B 1 1 (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring Device (Parameter 19) ; ;
HP Sampler Rotameter
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 103 OF 156 C
ATTACHMENT 15 (Page 3 of 3)
North Anna Radioactive Gaseous Emuent Monitoring Instrumentation AGION 1: If the number of operable channels is less than required, effluem releases, via this path, may continue if the flow rate is estimated at least once per four hours.
ACION 2: If the number of operable channels is less than required, effluent releases, via this path, may continue if grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity or gamma isotopic activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
AGION 3: If the number of operable channels is less than required, effluent releases, via this path, may continue if the frequency of the grab samples provided by Technical Specification requirement 4.4.6.3.b is increased to at least once per four hours and these samples are analyzed for gross activity or gamma isotopic activity within eight hours.
ACION 4: If the number of operable channels is less than required, the contents of the Waste Gas Decay Tanks may be rdcased to the environment provided that prior to initiation of the release:
- a. At least two independent samples of the tank's contents are analyzed, and:
- b. At least two technically qualified members of the Station staffindependently verify the release rate calculations and discharge valve lineup.
ACTION 5: If the number of operable channels is less than required, effluent releases from the Waste Gas Decay Tank may continue provided samples are continuously collected with auxiliary sampling equipment as required in Attachment 11.
NOTE 1: A channel shall consist of:
- a. The flow instrument installed in the ejector through which the discharge is routed; either Train A (1-SV-FI-100A,101 A), or Train B (1-SV-FI-100B,101B) or both.
- b. Flow instruments 101 A and 101B provide low range measurement. Flow instruments 100A and 100B provide high range measurement.
NOTE 2: A channel shall consist of:
- a. The flow instrument installed in the ejector through which the discharge is routed; either Train A (2-SV-FI-200A,201 A), or Train B (2-SV-FI-200B,201B) or both,
- b. Flow instruments 201 A and 201B provide low range measurement. Flow instruments 200A and 200B provide high range measurement.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 105 OF 156 ATTACHMENT 16 .
(Page 1 of 2)
Surry Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTI NAL
- 1. PROCESS VENT SYNitM (a) Noble Gas Activity Monitor-Providing Alarm and Automatic Termination of Release 1-GW-RM 102 l-GW-RM 130-1 D M'* R O 1 (b) Iodine Sampler Process Vent Continuous HP Sampler, or 1-GW-RM.130-1 W N/A N/A N/A (c) Particulate Sampler Process Vent Continuous HP Sampler, or 1-GW RM- 130-1 W N/A N/A N/A O (d) Process Vent Flow Rate Monitor
( l-GW-FI'-100 D N/A R N/A (e) Sampler Flow Rate Measuring Device HP Sampler Rotometer, or D N/A SA N/A KAMAN Flow Rate Measuring D N/A R N/A Device (Parameter #19)
- 2. CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor Unit 1: 1 SV-RM-Ill D M R Q Unit 2: 2-SV-RM-211 ,
(b) Air Ejector Flow Rate Measuring Device Unit 1: 1-VP-F1-1 A D N/A R N/A Unit 2: 2-VP-F1-1A 2-VP-F1-1B
- 3. VENTILATION VENT SYSTEM (a) Noble Gas Activity Monitor SRF: RRM 101 SPS: 1-VG-RM -110 D M R Q 1 VG-RM 131-1 O
1 VIRGINIA VPAP-2103 i POWER REVISION 7 l PAGE 106 OF 156 ATTACHMENT 16 ,
l (Page 2 of 2) l Surry Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL SOURCE CIIANNEL CH N DESCRIPTION CIIECK CHECK CALIBRATION FUN O AL TEST (b) Iodine Sampler l SRF: RRM-101 l SPS: Vent #1,1-VG-RM IN Vent #2, Continuous IIP Sampler, or 1-VG-RM-131- N/A @ M 1 i 1
(c) Particulate Sampler SRF: RRM 101 SPS: Vent #1,1-VG-RM 104 Vent #2, Continuous HP l
Sampler, or 1-VG.RM 131-1 (d) Ventilation Vent Flow Rate Monitor g SRF: Ol RHV-FT-156 W !
SPS: Vent #1,1-VS-FT-119 D N/A R N/A Vent #2,1-VS-FT-116 (e) Sampler Flow Rate Measuring Device SRF: RRM 101 D N/A R N/A SPS: Vent #1,1-VG-RM-lM D N/A R N/A Vent #2,KAMAN Flow Rate D N/A R N/A Measuring Device
, (Parameter #19), or HP D N/A N/A Sampler Rotometer S/A
- Prior to each Waste Uas Decay lank release l
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VIRGINIA VPAP 2103 POWER REVISION 7 PAGE 107 OF 156 ATTACHMENT 17 (Page 1 of 3)
North Anna Radioactive Gaseous Emuent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION CHECK CHECK CALIBRATION FUNCTIONAL TEST
- 1. PROCESS VENT SYSTEM (a) Noble Gas Activity Monitor 1-RM-GW-102 D M (NOTE 5) R Q (NOTE 1) 1 RM-GW-178-1 D M (NOTE 5) R Q (NOTE 1)
(b) Iodine Sampler 1-RM-GW-178-1 W N/A N/A N/A Process Vent Continuous HP Sampler D (NOTE 3) N/A N/A N/A (c) Particulate Sampler 1-RM-GW-178-1 W N/A N/A N/A Process Vent Continuous HP O( Sampier D(NOTE 3) N/A N/A N/A (d) Total Flow Monitor 1-GW-FT-108 D N/A R Q (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring D (NOTE 3) N/A R N/A Device (Parameter 19)
HP Sampler Rotameter D (NOTE 3) N/A SA N/A
- 2. CONDENSER AIR EJECTOR SYSTEM (a) Noble Gas Activity Monitor Unit I l SV-RM-121 D M R Q(NOTE 1)
Unit 2 2 SV-RM-221 (b) Flow Rate Measuring Device Unit I l-SV-F1-100A 1-SV FI-101 A D N/A R N/A 1-SV-FI 100B 1-SV-FI-101B Unit 2 2 SV-FI-200A
^ 2-SV-FI-201 A D N/A R N/A t 2 SV-FI-200B 2-SV FI-201B
1 VIRGINIA VPAP-2103 l POWER REVISION 7 PAGE 108 OF 156 ATTACHMENT 17 !
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North Anna Radioactive Gaseous Emuent Monitoring Instrumentation Surveillance Requirements CHANNEL CIIANNEL SOURCE CHANNEL W O AL DESCRIPTION CHECK CHECK CALIBRATION TEST
- 3. VENTILATION VENT A (a) Noble Gas Activity Monitor l l-RM-VG 104 D M R Q (NOTE 2) 1-RM-VG 179-1 D M (NOTE S) R Q (NOTE 2)
(b) lodine Sampler i 1-RM-VG-179- 1 W N/A N/A N/A l
Vent Vent A Continuous HP Sampler D (NOTE 3) N/A N/A N/A (c) Particulate Sampler 1 RM-VG-179-1 W N/A N/A N/A l Vent Vent A Continuous HP Sampler D (NOTE 3) N/A N/A N/A ,
(d) Total Flow Monitor l l-HV-FT-1212A D N/A R Q j (c) Sampler Flow Rate Measuring l l
Device KAMANS Flow Rate Measuring D (NOTE 3) N/A R N/A Device (Parameter 19)
HP Sampler Rotameter D (NOTE 3) N/A SA N/A
- 4. VENTILATION VENT B (a) Noble Gas Activity Mordtor l l RM-VG-113 D M R Q (NOTE 4) l l-RM VG 180-1 D M (NOTE S) R Q (NOTE 2)
(b) lodine Sampler 1-RM-VG-180-1 W N/A N/A N/A Vent Vent B Continuous HP Sampler D (NOTE 3) N/A N/A N/A (c) Particulate Sampler 1-RM VG-180-1 W N/A N/A N/A Vent Vent B Continuous HP Sampler D (NOTE 3) N/A N/A N/A
- (d) Total Flow Monitor l l-HV-FT-1212B D N/A R Q
, (e) Sampler Flow Rate Measuring l Device KAMANS Flow Rate Measuring D (NOTE 3) N/A R N/A Device (Parameter 19) l HP Sampler Rotameter D (NOTE 3) N/A SA N/A
. .. . .. . _ - . .-~___- .- - . . . - - -_.-- _
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 109 OF 156 '
ATTACHMENT 17 (Page 3 of 3)
North Anna Radioactive Gaseous Emuent Monitoring InstrumentationSurveillance Requirements NOTE 1: The Channel Functional Test shall demonstrate:
- a. Automatic actuation of the valves in this pathway and Control Room alarm annunciation occur if the instrument indicates measured levels above the alarm / trip setpoint.
- b. Alarm annunciation occurs if the instrument controls not set in " operate" mode.
NOTE 2: The Channel Functional Test shall demonstrate:
- a. Control Room alarm annunciation occurs if the instrument indicates measured levels are above the alarm / trip setpoint.
- b. Alarm annunciation occurs if the instrument controls not set in " operate" mode.
^
NOTE 3: Channel Checks shall consist of verifying indication of flow during periods of release.
Q Channel Checks shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.
NOTE 4: The Channel Functional Test shall demonstrate that:
- a. Control Room alarm annunciation occurs if the instrument indicates measured levels are above alarm / trip setpoint.
- b. The Instrument mode selection control automatically resets to " operate" mode when released.
NOTE 5: Monitors 1-RM-GW-178-1,1-RM-VG-179-1, and 1-RM-VG-180-1 are electronically source checked using an LED.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 111 OF 156 ATTACHMENT 18 (Page 1 of 2)
Critical Organ and Inhalation Dose Factors for Surry (Critical Pathway Dose Factors)
Ventilation Vent D/Q = 9.0E-10 m-2 at 5150 meters S Direction Process Vent D/Q = 4.3E-10 m-2 at 5150 meters S Direction Radionuclide RMin RMj p mrem /yr mrem /yr Curie /sec Curie /sec H-3 7.20E+02 3.12E+02
~~
Mn-54 ND ND Fe-59 ND ND Cr-51 6.45E+01 3.08E+01 Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND O Sr-89 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND ,
Ru-103 ND ND Ru-106 ND ND )
Ag-110m ND ND Te-127m 8.06E+04 3.85E+04 Te-129m 1.25E+05 5.98E+04 I-131 6.21E+08 2.97E+08 I-133 5.79E+06 2.77E+06 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Cc-141 ND ND g
\ Cc-144 ND ND ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 112 OF 156 ATTACHMENT 18 (Page 2 of 2)
Critical Organ and Inhalation Dose Factors for Surry (Inhalation Pathway Dose Factors) 3 Ventilation Vent x/Q = 3.0E-07 sec/m at 5150 meters S Direction Process Vent y/O = 1.3E-07 sec/m3 at 5150 meters S Direction Radionuclide RMiyy RMipy mrem /vr mrem /vr Curie /sec Curie /sec H-3 1.94E+02 8.41E+01 Cr-51 1.73E+01 7.48E+00 Mn-54 ND ND l l
Fe-59 ND ND i Co-58 ND ND Co-60 ND ND Zn-65 ND ND l Rb-86 ND ND Sr-89 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-127m 1.46E+03 6.33E+02 Te-129m 1.64E+03 7.12E+02 1-131 4.45E+06 1.93E+06 I-133 1.07E+06 4.63E+05 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Cc-141 ND ND
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Cc-144 ND ND ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1
VIRGINIA VPAP-2103
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4 ATTACHMENT 19 (Page 1 of 1)
Critical Organ Dose Factors for North Anna o
(Critical Pathway Dose Factors)
Ventilation Vent D/Q = 2.4E-09 m 2 at 3250 meters N Direction Process Vent D/Q = 1.lE-09 m-2 at 3250 meters N Direction
, Radionuclide RM yy RM mrem /vr mred; vr 3
Curie /sec Curie /sec i H-3 1.73E+03 9.36E+02
- Mn-54 ND ND Fe-59 ND ND Cr-51 1.50E+02 6.89E+01
~
Co-58 ND ND Co-60 ND ND Zn-65 ND ND Rb-86 ND ND Sr-89 ND ND Sr-90 ND ND Y-91 ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-127m 1.97E+05 9.04E+04 Te-129m 2.95E+05 1.35E+05 I-131 1.45E+09 6.72E+08 I-133 1.33E+07 6.12E+06 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Cc-141 ND ND l \ Ce-144 ND ND ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1 j
VIRGINIA VPAP-2103
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! ATTACHMENT 20 (Page 1 of 3)
Surry Radiological Environmental Monitoring Program l Exposure Pathway Number of Sample and Collection Type and Frequency of l and/or Sample Sample Location Frequency Analysis
- 1. DIRECT
, RADIATION l About 40 Roudne i Monitoring Stations to be placed as follows:
- 1) Inner Ring in general 6 area of site boundary GAMMA DOSE with station in each sector
- 2) Outer Ring 6 to 8 km
! from the site with a Quarterly Quarterly
! station in each sector l ^ 3) The balance of the 8
!( dosimeters should be placedin specialinterest areas such as population l centers, nearby residents, schools, and in 2 or 3 areas to serve as controls
- 2. AIRBORNE Samples from 7 locations:
j a) I sample from close to
! the site boundary
! location of the highest Radiciodine Canister l calculated annual 1131 Analysis Weekly l average ground level Continuous Radiciodines and D/Q Sampler Particulates b) 5 sample locadons 6-8 operadon with Particulate Sampler km distance located in a samplecollection Gross beta radioactivity concentric ring around weekly analysis following filter
, the Stadon change; c) I sample from a control . .
i location 15-30 km Gammaisotopic analys.is
"'P 8 b c ground a "II ") IY
VIRGINTA VPAP-2103 POWER REVISION 7 PAGE 116 OF 156 ATTACHMENT 20 (Page 2 of 3)
Surry Radiological Environmental Monitoring Program Expcaure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis
- 3. WATERBORNE Gamma isotopic analysis a) I sample upstream a) Surface Monthly Sample *Composite "OIY: for tritium .
b) I sample downstream analysis quarterly Sample from 1 or 2 sources Gannaisotopicand tritium b) Ground Quarterly ,
analysis quarterly c) Sediment I sample from downstream
. Gamma isotopic analysis from area with existing or Semi-Annually semi-annually shoreline Potential recreational value 5 samples from vicinity of Gamma isotopic analysis d) Silt the Station Semi-Annually g, g;
- 4. INGESTION a) Milk a) 3 samples from milking (NOTE 1) animals,in the vicinity of Ganna isotopic and I*
b) sample rom milking Monthly animals at a control analys,s i montNy location (15-30 km distant) a) 2 samples of oysters in the vicinity of the Semi-Annually Gamma isotopic on edibles Station b) 4 samples of clams in the vicinity of the Semi-Annually Gamma isotopic on edibles b) Fish and Station Invertebrates c) I sampling of crabs from the vicinity of the Annually Gamma isotopic on edibles Station d) 2 samples of fish from Semi-Annually Gamma isotopic on edibles t n atf'sh hite perch, cel)
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1 VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 117 OF 156 l r l
ATTACHMENT 20 ~
(Page 3 of 3) j Surry Radiological Environmental Monitoring Program i
l Exposure Pathway Number of Sample and Collection Type and Frequency of I and/or Sample Sample Location Frequency Analysis i
- 4. INGESTION (Continued)
Gamma isotopic on edible
) sam le oybeans Annually c) I sample peanuts d) I sample of a broadleaf vegetation of two different available offsite locations with l highest annual average l c) Food ground levelD/Q,if one Products or more milk samples Monthly, if Gamma isotopic and 1 131 lq
- are unavailable available, or at .
an lysis lV e) I sample of a broadleaf harvest l vegetation grown 15-30 kmin the available,least prevalent wind direction,if one or more l milk samples are unavailable l
NOTE 1: If milk sampling cannot be performed, use item 4.c (d).
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ATTACHMENT 21 (Page 1 of 5)
North Anna Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location (NOTE 2) Frequency Analysis
- 1. DIRECT RADIATION (NOTE 3) 36 routine monitoring stations, either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously, to be placed as follows:
- 1) An inner ring of stations, one in each emergency GAMMA DOSE
,) meteorological sector within the site boundary
- 2) An outer ring of stations, one in each emergency Quarterly Quarterly meteorological sector within 8 km range from the site
- 3) The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations O) v
VIRGINIA VPAP-2103 POWER REVISION 7 l PAGE 120 OF 156 A'lTACHMENT 21 l (Page 2 of 5) !
1 North Anna Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location (NOTE 2) Frequency Analysis
- 2. AIRBORNE Samples from 5 locations:
a) 3 samples from close to the 3 site boundary ,
locations (in different .. !
ster i sectors) of the highest Radigiodine Cam.,
I Analysis, weekly I calculated historical annual average ground Continuous Radiciodines and level D/Q sampler, Particulates b) 1 sample from the operation with vicinity of a community sample Particulate Sampler having the highest collection Gross beta radioactivity calculated annual weekly analysis following filter average ground level change; (NOTE 4)
D/Q c) I sample from a control Gammaisotopic analysis location 15-40 km of composite (by distant and in the least location) quarterly prevalent wind direction (NOTE 5)
- 3. WATERBORNE Samples from 3 locations:
Gamma isotopic analysis a) I sample upstream monthly;(NOTE 5) a) Surface b) I sample downstream Grab Monthly .
Composite for tritium c) 1 sample from cooling analysis quarterly lagoon Sample from 1 or 2 sources Gammaisotopic and tritium b) Ground Grab Quarterly only iflikely to be affected analysis quarterly (NOTE 5) 1 sample from downstream i Gamma isotopic analysis c) Sediment area with existing or Semi-Annually semi-annually (NOTE 5) potential recreational value i
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VIRGINIA VPAP-2103 i POWER REVISION 7 l PAGE 1210F 156 i
(VD ATTACHMENT 21 (Page 3 of 5) i North Anna Radiological Environmental Monitoring Program 1
Exposure Pathway Numberof Sampleand Sample Collection Type and Frequency of and/or Sample Location (NOTE 2) Frequency Analysis
- 4. INGESTION a) Samples from milking i, animals in 3 locations within I 5 km that have the highest potential. If there are none, then I sample from milking i arircMs in each of 3 areas a) Milk between 5 :o 8 km where Monthly atall Gammaisotopic(NOTE 5)
(NOTE 7) doses are calculated to be times and1131 analysis monthly greater than 1 mrem per yr (NOTE 6)
- b) 1 sample from milking l
- /~ animals at a controllocation
(
l (15-30 km in the least prevalent wind direction) l s a) I sample of commercially and I 1 recreationally important species (bass, sunfish, catfish) l b) Fish and in vicinity of plant discharge Gamma isotopic on edible Semiannually Invertebrates area portions b) I sample of same species in areas not influenced by plant discharge a) Samples of an edible broad leaf vegetation grown nearest each of two different offsite locations of highest predicted historical annual average Monthly if c) Food ground level D/Q if milk . Gammaisotopic(NOTE 5) available, or Products sampling is not performed and1 131 analysis at harvest b) 1 sample of broad leaf vegetation grown 15-30 kmin the least prevalent wind direction if milk sampling is not performed
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 122 OF 156 ATTACIIMENT 21 (Page 4 of 5)
North Anna Radiological Environmental Monitoring Program NOTE 1: The number, media, frequency, and location of samples may vary from site to site. This table presents an acceptable minimum program for a site at which each entry is applicable.
Local site characteristics must be examined to determine if pathways not covered by this table may significantly contribute to an individual's dose and be included in the sampling program.
NOTE 2: For each and every sample location in Attachment 21, specific parameters of distance and direction sector from the centerline of the reactor, and additional description where pertinent, shall be provided in Attachment 23. Refer to Radiological Assessment Branch Technical Positions and to NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plant. Deviations are permitted from the required sampling schedule if specimens are unattainable due to hazardous coadi tions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons.
If specimens are unattainable due to sampling equipment malfunction, every effort shall be made to complete corrective action before the end of the next sampling period. All deviations from the sampling schedule shal.! be documented in the Annual Radiological Environmental Operating Report pursuant te 6.7.1. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable attemative media and locations may be chosen for the particular pathw ay in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report and pursuant to 6.7.2, identify the cause of the unavailability of samples for that pathway and identify the new locations for obtaining replacement samples in the next Annual Radioactive Effluent Release Report, and include revised figures and tables from the ODCM j eflecting the new locations in the report.
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! ATTACHMENT 21
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(Page 5 of 5)
North Anna Radiological Environmental Monitoring Program NOTE 3: One or more instruments, such as a pressurized ion chamber, for measuring and recording j dose rate continuously may be used in place of, or in addition to, integrating dosimeters.
l For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.
Film badges shall not be used as dosimeters for measuring direct radiation. The 36 stations l
l are not an absolute number. 'Ihe number of direct radiation monitoring stations may be reduced according to geographical limitations, e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly.The frequency of l
analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.
l l I NOTE 4: Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
NOTE 5: Gamma isotopic analysis is the identification and quantification of gamma-emitting radionuclides that may be attributable to effluents from the facility.
NOTE 6: The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.
NOTE 7: If milk sampling cannot be performed, use item 4.c (Page 3 of 5, Attachment 21) l l
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VIRGINLA VPAP-2103 POWER REVISION 7 PAGE 125 OF 156 V) r ATTACHMENT 22 1
(Page 1 of 4) !
Surry Environmental Sampling Locations l
l SAMPLE LOCATION DISTANCE DIRECTION REMARKS I MEDIA (MILES)
Air Charcoal Site Boundary and Particulate Surry Station (SS) 0.37 NNE Location at Sector with Highest D/Q Hog Island Reserve (HIR) 2.0 NNE Bacons Castle (BC) 4.5 SSW Alliance (ALL) 5.1 WSW Colonial Parkway (CP) 3.7 NNW Dow Chemical (DOW) 5.1 ENE '
Fort Eustis (FE) 4.8 ESE Newport News (NN) 16.5 ESE Control Location Environmental Control (00) Onsite ** I
( TLDs West North West (02) 0.17 WNW Site Boundary Surry Station Discharge 0.6 NW Site Boundary (03)
North North West (04) 0.4 NNW Site Boundary North (05) 0.29 N Site lloundary i l
North North East (06) 0.28 NNE Site Boundary North East (07) 0.31 NE Site Boandary East North East (08) 0.43 ENE Site Boundary i East (Exclusion) (09) 0.31 E Onsite West (10) 0.40 W Site Boundary West South West (11) 0.45 WSW Site Boundary South West (12) 0.30 SW Site Boundary )
South South West (13) 0.43 SSW Site Boundary South (14) 0.48 S Site Boundary ,
0.74 SSE Site Boundary l South South East (15)
South East (16) 1.00 SE Site Boundary l East (17) 0.57 E Site Boundary Station Intake (18) 1.23 ESE Site Boundary Hog Island Reserve (19) 1.94 NNE Near Resident l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 126 OF 156 ATTACHMENT 22 (Page 2 of 4)
Surry Environmental Sampling Locations SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES)
Environmental Bacons Castle (20) 4.45 SSW Approx. 5 miles TLDs Route 633 (21) 3.5 SW Approx. 5 miles Alliance (22) 5.1 WSW Approx. 5 miles Surry (23) 8.0 WSW Population Center Route 636 and 637 (24) 4.0 W Approx. 5 miles Scotland Wharf (25) 5.0 WNW Approx. 5 miles Jamestown (26) 6.3 NW Approx. 5 miles Colonial Parkway (27) 3.7 NNW Approx. 5 miles Route 617 and 618 (28) 5.2 NNW Approx. 5 miles Kingsmill (29) 4.8 N Approx. 5 miles Williamsburg (30) 7.8 N Population Center Kingsmill North (31) 5.6 NNE Approx. 5 miles Budweiser (32) 5.7 NNE Population Center Water Plant (33) 4.8 NE Approx. 5 miles Dow (34) 5.1 ENE Approx. 5 miles Lee Hall (35) 7.1 ENE Population Center Goose Island (36) 5.0 E Approx. 5 miles Fort Eustis (37) 4.8 ESE Approx. 5 miles Newport News (38) 16.5 ESE Population Center James River Bridge (39) 14.8 SSE Control l Benn's Church (40) 14.5 S Control Smithfield (41) 11.5 S Control Rushmere (42) 5.2 SSE Approx. 5 miles Route 628 (43) 5.0 S Approx. 5 miles Milk Epp's 4.8 SSW Colonial Parkway 3.7 NNW l Judkin's 6.2 SSW William's 22.5 S Control Location l
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5 ATTACHMENT 22 (Page 3 of 4)
Surry Environmental Sampling Locations l . SAMPLE LOCATION DISTANCE DIRECTION REMARKS l MEDIA (MILES)
Well Water Surry Station Onsite* *
- Hog Island Reserve 2.0 NNE l Crops (Corn, Slade's Farm 2.4 S Peanuts, Soybeans) Brock's Farm 3.8 S Crops Spratley's Garden 3.2 S (Cabbage, Kale) Carter's Grove Garden 4.8 NE
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Lucas's Garden !
(Chester, Va.) '
River Water Suny Discharge 0.17 NW (Monthly) Scotland Wharf 5.0 WNW Control Location l V Sediment Chickahominy River 11.2 WNW Control Location l (Silt) Surry Station Intake 1.9 ESE Surry Station Discharge 1.0 NNW Hog Island Point 2.4 NE l Point of Shoals 6.4 SSE l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 128 OF 156 ATTACHMENT 22 (Page 4 of 4)
Surry Environmental Sampling Locations SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES)
Clams Chickahominy River 11.2 WNW Control Location Surry Station Discharge 1.3 NNW Hog Island Point 2.4 NE Lawne's Creek 2.4 SE Oysters Kingsmill 2.9 NE l Mulberry Point 4.9 EESE Crabs Surry Station Discharge 0.6 NW Fish Surry Station Discharge 0.6 NW Shoreline Hog Island Reserve 0.8 N Sediment
- Onsite Location -in Lead Shield
- Onsite sample of Well Water-taken from tap-water at Surry Environmental Building 4
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 129 OF 156 ATTACHMENT 23 -
(Page 1 of 4)
North Anna Environmental Sampling Locations 1
Distance and Direction From Unit No. I Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Environment NAPS SewageTreatment 01 0.20 NE Quaedy & On-Site al Plant Annually TLDs Frederick's Hall 02 5.30 SSW Q dy &
y Mineral,VA 03 7.10 WSW
- Od"*
Wares Crossroads 04 5.10 WSW &
Qdam Route 752 05 4.20 NNE 9 E Sturgeon's Creek Marina OSA 3.20 N Q &
Levy,VA 06 4.70 ESE 9 jYy A Bumpass, VA 07 7.30 SSE Y 9d* y #
End of Route 685 21 1.00 WNW a y& Site Boundary y
Route 700 22 1.00 WSW 9 dY # Site Boundary y
Qu y
" Aspen HiIls" 23 0.93 SSE hy & Site Boundary Orange, VA 24 22.00 NW Od"*yYE Control Bearing Cooling Tbwer N-1/33 0.06 N Quarterly On-Site Sturgeon's Creek Marina N-2/34 3.20 N Quanerly Parking Lot "C" NNE-3/35 0.24 NNE Quarterly On-Site l Good Hope Church NNE-4/36 4.96 NNE Quanerly Parking Lot "B" NE-5/37 0.20 NE Quarterly On-Site :
l Bogg's Drive NE-6/38 1.46 NE Quanerly l WeatherTower Fence ENE-7/39 0.36 ENE Quarterly On-Site Route 689 ENE 8/40 2.43 ENE Quarterly NearTraining Facility E-9/41 0.30 E Quanerly On-Site "Moming Glory Hill" E-10/42 2.85 E Quarterly Island Dike ESE-II/43 0.12 ESE Quanctly On-Site Route 622 ESE-12/44 4.70 ESE Quarterly i
, - - . - - --.e , _.,,i - - - --
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 130 OF 156 ATTACHMENT 23 .,
(Page 2 of 4)
North Anna Environmental Sampling Locations l
Distance and Direction From Unit No. I Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency mnment Biology Lab SE-13/45 0.75 SE Quarterly On-Site TLDs Route 701 (Dam Entrance) SE-14/46 5.88 SE Quanerly
" Aspen Hills" SSE-15/47 0.93 SSE Quanerly Site Boundary Elk Creek SSE-16/48 2.33 SSE Quanerly NAPS Access Road S-17/49 0.47 S Quanerly On-Site l Elk Creek Church S-18/50 1.55 S Quanerly NAPS Access Road SSW-19/51 0.42 SSW Quanerly On-Site l Route 618 SSW-20/52 5.30 SSW Quanerly 500KV Tower SW-21/53 0.60 SW Quanerly On-Site l Route 700 SW-22/54 4.36 SW Quanerly NAPS RadioTower WSW-23/55 0.38 WSW Quanerly On-Site l Route 700 WSW-24/56 1.00 WSW Quarterly Site Boundary South Gate of Switchyard W-25/57 0.32 W Quanerly On-Site l Route 685 W-26/58 1.55 W Quanerly End of Route 685 WNW-27/59 1.00 WNW Quanerly Site Boundan l Route 685 WNW-28/60 1.40 WNW Quanerly l l Laydown Area Nonh Gate NW-29/61 0.45 NW Quanerly On-Site l Lake Anna Campground NW-30/62 2.54 NW Quanerly
- 1/#2 Intake NNW-31/63 0.07 NNW Quanerly On-Site i Route 208 NNW-32/64 3.43 NNW Quanerly Bumpass Post Office C-1/2 7.30 SSE Quanerly Control Orange, VA C-3/4 22.00 NW Quanerly Control Mineral, VA C-5/6 7.10 WSW Quanerly Control Louisa, VA C-7/8 11.54 WSW Quanerly Control l
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J VIRGINIA VPAP-2103 1 POWER REVISION 7 !
PAGE 131 OF 156 A
ATTACHMENT 23 I (Page 3 of 4)
North Anna Environmental Sampling Locations l Distance and Direction From Unit No.1 l Sample Location Station Distance Direction Collection REMARKS :
Media No. (Miles) Frequency '
Airbome NAPS SewageTreatment 01 0.20 NE Weekly On-Site Plant Particulate Frederick's Hall 02 5.30 SSW Weekly and Mineral. VA 03 7.10 WSW Weekly l Radiciodine Wares Cmssroads 04 5.10 WNW Weekly I l
, Route 752 05 4.20 NNE Weekly Sturgeon's Creek Marina 05A 3.20 N Weekly Levy.VA 06 4.70 ESE Weekly Bumpass, VA 07 7.30 SSE Weekly ,
End of Route 685 21 1.00 WNW Weekly Site Boundary l 22 1.00 WSW Weekly Site Boundary I Route 700
" Aspen Hills" 23 0.93 SSE Weekly Site Boundary ;
Orange. VA 24 22.00 NW Weekly Control Q Surface Water Waste HeatTreatment Facility (Second Cooling 08 1.10 SSE Monthly
[ Commitment 3.2.2] Lagoon)
North Anna River .
(upstream)Rt 669 Bridge 09A 12.9 WNW Monthly Control I (Brook's Bndge)
Nonh Anna River 11 5.80 SE Monthly (downstream)
Aquatic fI Biology Lab Waste Heat Treatment OlA 0.75 SE Quarterly Sediment Facility (Second Cooling 08 1.10 SSE Semi-Annually i Lagoon) l North Anna River (upstream)Rt 669 Bridge 09A 12.9 WNW Semi-Annually Contml (Brook's Bridge)
North Anna River 11 5.80 SE Semi Annually (downstream)
Shoreline Soil Lake Anna (upstream) 09 2.20 NW Semi Annually Soil NAPS SewageTreatment 01 0.20 NE Once per 3 yrs On-Site Plant O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 132 OF 156 ATTACHMENT 23 (Page 4 of 4)
North Anna Environmental Sampling Locations Distance and Direction From Unit No. I Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Soil Fredericks Hall 02 5.30 SSW Once per 3 yrs (continued) Mineral VA 03 7.10 WSW Once per 3 yrs Wares Crossroads 04 5.10 WNW Once per 3 yrs
~
Route 752 05 4.20 NNE Once per 3 yrs Sturgeon's Crees Marina 05A 3.20 N Once per 3 yrs Levy,VA 06 4.70 ESE Once per 3 yrs Bumpass, VA 07 7.30 SSE Once per 3 yrs End of Route 685 21 1.00 WNW Once per 3 yrs Site Boundary Route 700 22 1.00 WSW Once per 3 yrs Site Boundary
" Aspen Hills" 23 0.93 SSE Once per 3 yrs Site Boundary Orange, VA 24 22.00 NW Once per 3 yrs Control Milk Holladay Dairy 12 8.30 NW Monthly (R.C. Go(xiwin)
Terrell's Dairy 13 5.60 SSE Monthly (Frederick's Hall)
Fish Waste Heat Treatment Semi-Facility (Second Cooling 08 1.10 SSE Annually Lagoon) 25 NW Semi-Lake Orange 16.50 Control Annually Food Products Route 713 14 vanes NE (Broad Leaf Route 614 15 varies SE Monthly vegetation) Route 629/522 16 varies NW if available, or Control l Route 685 21 varies WNW at harvest
" Aspen Hills" Area 23 varies SSE O
VIRGINIA VPAP-2103 POWER REVISION 7 )
PAGE 133 OF 156 '
N ATTACHMENT 24 (Page 1 of 2) ;
l' Detection Capabilities for Surry Environmental Sample Analysis 1
LOWER LIMIT OF DETECTION (LLD) l Airborne Food .
Sediment Analysis Water Particulate Milk Products (pC g) I EI (NOTE 2) (pCi/l) or Gases (pCi/l) (pCi/kg) ,)
(pCi/m3) (weg)
Gross beta 4 0.01 H-3 2,000 Mn-54 15 130
)
Fe-59 30 260 1 Co-58,60 15 130 Zn-65 30 260 Zr-95 30
(] Nb-95 15
'd I-131 (NOTE 3) 1 0.07 1 60 l Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-140 60 60 La-140 15 15 NOTE 1: Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13.
NOTE 2: This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
NOTE 3: LLD for the Ground (drinking) Water Samples. The LLD for the surface (non-drinking) water samples is 10 pCi/1.
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l ATTACHMENT 24 O l
(Page 2 of 2)
Detection Capabilities for Surry Environmental Sample Analysis LOWER LIMIT OF DETECTION (LLD)
NOTE 1: For a particular measurement system (which may include radiochemical separation):
4.66 s b l
LLD = (24-1)
E . V e 2.22E+06 e Y = e-(AAt)
Where:
LLD = the "a priori" (before the fact) Lower Limit of Detection as def'med above (as microcuries per unit mass or volume) (see 4.8) sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm)
E = the counting efficiency (as counts per disintegration) l V = the sample size (in units of mass or volume) 2.22E+06 = the number of disintegrations per minute (dpm) per microcurie Y = the fractional radiochemical yield (when applicable)
A = the radioactive decay constant for the particular radionuclide At = the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples)
Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori" (before the fact) limit representing the capability of a measurement system and not a "posteriori" (after the fact) limit for a particular measurement.
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VIRGINIA VPAP-2103 !
POWER REVISION 7 l PAGE 135 0F 156 ;
f%
ATTACHMENT 25 (Page 1 of 2) I Detection Capabilities for North Anna Environmental Sample Analysis I
LOWER LIMIT OF DETECTION (LLD) 1
^' '"'
Fish Sediment Analysis Water Particulate ""
(pCi/kg) (pCi/k8)
(NOTE 2) (pCi/l) or Gases (pCi/1) (pCi/kg)
(pCi/m3) (Wet) (weg) (Wet)
Gross beta 4 0.01 I H-3 2,000 :
Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zn-65 30 260 )
Zr-95 30 l Nb-95 15 I-131 (NOTE 3) 1 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-140 60 60 La-140 15 15 NOTE 1: Required detection capabilities for thermoluminescent dosimeters used for environmental
] measurements are given in Regulatory Guide 4.13.
NOTE 2: This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
NOTE 3: LLD for the ground (drinking) water samples. The LLD for the surface (non-drinking) water samples is 10 pCi/1.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 136 OF 156 ATTACHMENT 25 (Page 2 of 2)
Detection Capabilities for North Anna Environmental Sample Analysis LOWER LIMIT OF DETECTION (LLD) (NOTE 3)
NOTE 3: For a particular measurement system (which may include radiochemical separation):
4.66 s b LLD = (25-1)
E e V e 2.22E+06
- Y
- e-(AAt) 1 1
Where:
1 LLD = the "a priori"(before the fact) Lower Limit of Detection as defined above l (as microcuries per unit mass or volume) (see 4.9) sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute, cpm) l E = the counting efficiency (as counts per disintegration)
V = the sample size (in units of mass or volume) 2.22E+06 = the number of disintegrations per minute (dpm) per microcurie l Y = the fractional radiochemical yield (when applicable) l A = the radioactive decay constant for the particular radionuclide At = the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples)
Typical values of E, V, Y and At should be used in the calculation.
The LLD is an "a priori"(before the fact) limit representing the capability of a measurement system and not a "posteriori" (after the fact) limit for a particular measurement.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 137 OF 156 ATTACHMENT 26 (Page1of1)
Reporting Levels for Radioactivity Concentrations in Environmental Samples at Surry AI' '"'
Water Fish Milk Food Produits A "8II ". (pCill)
"' (pCi/kg, wet) (pCill) (pCi/kg. wet)
G (p i m3)
H-3 30,000 Mn-54 1,000 30,000 Fe 59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 I-131 (NOTE 1) 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 NOTE 1: Reporting level for the ground (drinking) water samples required by Attachment 20. The i l
reporting level for the surface (non-drinking) water samples required by Attachment 20 is 20 pCi/1.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 138 OF 156 9 1 l
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 139 OF 156 0%-
ATTACHMENT 27 (Page1of1)
Reporting Levels for Radioactivity Concentrations in Environmental Samples at North Anna
^I' '"'
Water Fish Milk Food Products
^ "* II '". (pCi/l)
(PCi/kg, wet) (pCi/I) (pCi/kg, wet)
G pi )
(NOTE 1)
N' 20,000 l ,
I Mn-54 1,000 30,000 Fe-59 400 10,000
! Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 I-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 i NOTE 1: For drinking water samples l
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VIRGINIA VPAP-2103 POWER REVISION 7 l , PAGE 141 OF 156 ,
i l ATTACHMEN'T 28 _
j (Page 1 of 8) l Surry Meteorological, Liquid, and Gaseous Pathway Analysis 1.0 METEORLOGICAL ANALYSIS j 1.1 Purpose The purpose of the meteorological analysis was to determine the annual average X/Q and D/Q values at critical locations around the Station for ventilation vent (ground level) and process l vent (mixed mode) releases. The annual average. X/Q and D/Q values were used in a dose l pathway analysis to determine both the maximum exposed individual at site boundary and member of the public. The X/Q and D/Q values resulting in the maximum exposures were incorporated into the dose factors in Attachments 12 and 18.
1.2 Meteorological Data, Parameters, and Methodology 1
Onsite meteorological data for the period January 1,1979, through December 31,1981, were l used in calculations. Th ese data included wind speed, wind direction, and differential l ,. temperatu e for the purpose of determining joint frequency distributions for those releases
! (/ characterized as ground level (i.e., ventilation vent), and those characterized as mixed mode j i (i.e., process vent). The portions of release characterized as ground level were based on
! AT158.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were j based on AT158.9ft-28.2ft and 158.9 ft wind data.
l %/Qs and D/Qs were calculated using the NRC computer code "XOQDOQ - Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations",
September,1977. The code is based upon a straight line airflow model implementing the assumptions outlined in Section C (excluding Cla and Clb) of Regulatory Guide 1.111, j
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors."
The open terrain adjustment factors were applied to the X/Q values as recommended in Regulatory Guide 1.111. The site region is characterized flat terrain such that open terrain correction factors are considered appropriate. The ground level ventilation vent release calculations included a building wake correction based on a 1516 m2 containment minimum cross-sectional area. The effective release height used in mixed mode release calculations was g based on a process vent release height of 131 ft, and plume rise due to momentum for a vent
- V diameter of 3 in with plume exit velocity of 100 ft/sec.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 142 OF 156 ATTACHMENT 28 (Page 2 of 8)
Surry Meteorological, Liquid, and Gaseous Pathway Analysis Ventilation vent, and vent releases other than from the process vent, are considered ground j level as specified in Regulatory Guide 1.111 for release points less than the height of adjacent solid structures. Terrain elevations were obtained from Surry Power Station Units 1 and 2 Virginia Electric and Pcwer Company Updated Final Safety Analysis Report Table 11 A-8.
X/Q and D/Q values were calculated for the nearest site boundary, resident, milk cow, and vegetable garden by sector for process vent and ventilation vent releases. %/Q values were also calculated for the nearest discharge canal bank for process and ventilation vent releases.
According to the definition for short term in NUREG-0133," Preparation of Radiological l Effluent Technical Specifications for Nuclear Power Stations," October,1978, some gaseous releases may fit this category, primarily waste gas decay tank releases and containment purges.
However, these releases are considered long term for dose calculations as past releases were both random in time of day and duration as evidenced by reviewing past release reports.
Therefore, the use of annual average concentrations is appropriate according to NUREG-0133.
1.3 Results The X/Q value that resulted in the maximum total body, skin, and inhalation exposure for ventilation vent releases was 6.0E-05 sec/m3 at a site boundary location 499 meters N sector.
For process vent releases, the site boundary X/Q value was 1.0E-06 sec/m3 at a location 644 meters S sector. The discharge canal bank %/Q value that resulted in the maximum inhalation exposure for ventilation vent releases was 7.8E-05 sec/m3 at a location 290 meters NW sector.
The discharge canal bank X/Q value for process vent was 1.6E-06 sec/m3 at a location 290 '
meters NW sector.
Pathway analysis indicated that the maximum exposure from 1131, 1133, and from all radionuclides in particulate form with half-lives greater than 8 days, was through the grass-cow-milk pathway. The D/Q value from ventilation vent releases resulting in the maximum exposure was 9.0E-10 per m2at a location 5150 meters S sector. For process vent releases, the D/Q value was 4.3E-10 per m2 at a location 5150 meters S sector. For tritium, the X/Q value from ventilation vent releases resulting in the maximum exposure for the milk pathway was 3.0E-07 sec/m3, and 1.3E-07 sec/m3for process vent releases at a location 5150 meters S sector. The inhalation pathway is the only other pathway existing at this location. Therefore, the X/Q values given for tritium also apply for the inhalation pathway.
i l VIRGINIA VPAP-2103 l POWER REVISION 7 i PAGE 143 0F 156 m
ATTACHMENT 28
( (Page 3 of 8) l Surry Meteorological, Liquid, and Gaseous Pathway Analysis 2.0 LIQUID PATHWAY ANALYSIS 2.1 Purpose The purpose of the liquid pathway analysis was to determine the maximum exposed member of the public in unrestricted areas as a result of radioactive liquid effluent releases. The analysis included a determination of most restrictive liquid pathway, most restrictive age group, and critical organ. This analysis is required for Subsection 6.2, Liquid Radioactive Waste Effluents.
2.2 Data, Parameters, and Methodology Radioactive liquid effluent release data for the years 1976,1977,1978,1979,1980, and 1981 were compiled from the Surry Power Station effluent release reports. The data for each year, along with appropriate site specific parameters and default selected parameters, were entered into the NRC computer code LADTAP as described in NUREG-0133.
- n U Liquid radioactive effluents from both units are released to the James River via the discharge l
canal. Possible pathways of exposure for release from the Station include ingestion of fish and invertebrates and shoreline activities. The irrigated food pathway and potable water pathway do not exist at this location. Access to the discharge canal by the general public is gained two ways: bank fishing, controlled by the Station and limited to Virginia Power employees or guests of employees, and by boat as far upstream as the inshore end of the discharge canal l groin. It has been estimated that boat sport fishing would be performed a maximum of 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year, and that bank fishing would be performed a maximum of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per year.
l l For an individual fishing in the discharge canal, no river dilution was assumed for the fish
! pathway. For an individual located beyond the discharge canal groins, a river dilution factor of 5 was assumed as appropriate according to Regulatory Guide 1.109, Rev.1, and the fish, invertebrate, and shoreline pathways were considered to exist. Dose factors, bioaccumulation factors, and shore width factors given in Regulatory Guide 1.109, Rev.1, and in LADTAP were used, as were usage terms for shoreline activities and ingestion of fish and invertebrates. Dose to an individual fishing on the discharge bank was determined by multiplying the annual dose g calculated with LADTAP by the fractional year the individual spent fishing in the canal.
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VIRGINIA VPAP-2103 i POWER REVISION 7 PAGE 144 OF 156 ATTACHMENT 28 (Page 4 of 8) l Surry Meteorological, Liquid, and Gaseous Pathway Analysis i
1 2.3 Results For the years 1976,1977,1979,1980, and 1981, the invertebrate pathway resulted in the largest l dose. In 1978 the fish pathway resulted in the largest dose. The maximum exposed member of the public was determined to utilize the James River. The critical age group was the adult and l the critical organ was either the thyroid or GI-LLI. The ingestion dose factor, Ai, in 6.2.3 includes the fish and invertebrate pathways. Ai dose factors were calculated for the total body, thyroid, and GI-LLI organs.
3.0 GASEOUS PATHWAY ANALYSIS 3.1 Purpose A gaseous effluent pathway analysis was performed to determine the location that would result in the maximum doses due to noble gases, for use in demonstrating compliance with 6.3.1.a.
and 6.3.3.a. The analysis also included a determination of the location, pathway, and critical organ, of the maximum exposed member of the public, as a result of the release of 1131, I133, tritium, and for all radionuclides in particulate form with half-lives greater than eight days for use in demonstrating compliance with 6.3.4.a. In addition, the analysis included a determination of the critical organ, maximum age group, and sector location of an exposed individual through the inhalation pathway from 1131, 1133, tritium, and particulates to demonstrate compliance with 6.3.1.a..
3.2 Data, Parameters, and Methodology Annual average X/Q values were calculated, as described in Section 1 of this attachment, for the nearest site boundary in each directional sector and at other critical locations accessible to the public inside site boundary. The largest X/Q value was determined to be 6.0E-05 sec/m3 at site boundary for ventilation vent releases at a location 499 meters N direction, and 1.0E-06 sec/m3 at site boundary for process vent releases at a location 644 meters S direction. The maximum doses to total body and skin, and air doses for gamma and beta radiation due to noble gases would be at these site boundary locations. The doses from both release points are summed in calculations to calculate total maximum dose.
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VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 145 OF 156 ATTACHMENT 28 (Page 5 cf 8)
Surry Meteorological, Liquid, and Gaseous Pathway Analysis Step 6.3.1.a.2 dose limits apply specifically to the inhalation pathway. Therefore, the locations and %/Q values determined for maximum noble gas doses can be used to determine the maximum dose from 1131, 1133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway.
l The NRC computer code GASPAR," Evaluation of Atmospheric Releases," Revised 8/19/77, l was run using 1976,1977,1978,1979,1980, and 1981 Surry Power Station gaseous effluent I release report data. Doses from I131, g133, tritium, and particulates for the inhalation pathway were calculated using the 6.0E-05 sec/m3 site boundary %/Q. Except for the source term data l and the X/Q value, computer code default parameters were used. Results for each year indicated that the critical age group was the child and the critical organ was the thyroid for the inhalation pathway. In 1979, the teen was the critical age group. However, the dose calculated for the teen was only slightly greater than for the child and the doses could be considered equivalent, i
~ The gamma and beta dose factors K vy, Livy, Mivy, and N ivv in Attachment 12 were obtained by performing a units conversion of the appropriate dose factors from Table B-1, Regulatory Guide 1.109, Rev.1, to mrem /yr per Ci/m3 or mrad /yr per Ci/m3, and multiplying by the ventilation vent site boundary X/Q value of 6.0E-05 sec/m3. The same approach was used to calculate the gamma and beta dose factors Kipv, Lipv, Mipy, and Nipv in . Attachment 12, using l the process vent site boundary %/Q value of 1.0E-06 sec/m3 Inhalation pathway dose factors Pi vy and Pipv in Attachment 12 were calculated using the ;
equation:
P; = K'(BR) DFA;(x/Q)(mrem /yr per Curie /sec) (28-1) where:
K' = a constant of unit conversion,1E+12 pCi/Ci BR = the breathing rate of the child age group,3700 m3 /yr, from Table E-5, Regulatory Guide 1.109, Rev.1 DFAi = the thyroid organ inhalation dose factor for child age group for the ith radionuclide, in mrem /pCi, from Table E-9, Regulatory Guide 1.109, Rev.1 X/Q = the ventilation vent site boundary X/Q,6.0E-5 sec/m3, or the process vent site boundary X/Q,1.0E-06 sec/m3, as appropriate
l VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 146 OF 156 i ATTACHMENT 28 Ol (Page 6 of 8)
Surry Meteorological, Liquid, and Gaseous Pathway Analysis I
Step 6.3.4.a., requires that the dose to the maximum exposed member of the public from I 131 , ,
I133, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days l be less than or equal to the specified limits. Dose calculations were performed for an exposed member of the public within site boundary unrestricted areas, discharge canal bank, and to an l exposed member of the public beyond site boundary at real residences with the largest X/Q values using the NRC computer code GASPAR. Doses to members of the public were also calculated for the vegetable garden, meat animal, and milk-cow pathways with the largest D/Q j values using the NRC computer code GASPAR. j It was determined ti.at the member of the public within site boundary would be using the discharge canal bank for fishing a maximum of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per year. The maximum annual l
%/Q at this location was determined to be 7.8E-05 sec/m3 at 290 meters NW direction. After applying a correction for the fractional part of year an individual would be fishing at this location, the dose was calculated to be less than an individual would receive at site boundary.
The member of the public receiving the largest dose beyond site boundary was determined to j be located 5150 meters S sector. The critical pathway was the grass-cow-milk, the maximum age group was the infant, and the critical organ the thyroid. For each year 1976,1977,1978, 1979,1980 and 1981 the dose to the infant from the grass-cow-milk pathway was greater than the dose to the member of the public within site boundary, nearest residence, vegetable or meat l pathways. Therefore, the maximum exposed member of the public was determined to be the ,
1 infant, exposed through the grass-cow-milk pathway, critical organ thyroid, at a location 5150 l meters S sector. The only other pathway existing at this location for the infant is inhalation.
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i VIRGINIA VPAP-2103 POWER REVISION 7 l l PAGE 147 0F 156 .
! O(/ ATTACHMENT 28
)
(Page 7 of 8) l Surry Meteorological, Liquid, and Gaseous Pathway Analysis l
l The RM ivv and RMipv dose factors, except for tritium,in Attachment 18 were calculated by multiplying the appropriate D/Q value with the following equation: l dtia ff (i _ fPs f)e 4 it' RM; = K O (U,p ) F m F
(f)(N i ) PS 4 e (28-2)
_P s ,
l where:
- K' = a constant of unit conversion, IE+12 pCi/Ci QF = cow's consumption rate,50, in Kg/ day (wet weight)
Uap = infant milk consumption rate,330, liters /yr
- Yp = agricultural productivity by unit area of pasture feed grass,0.7 Kg/m 2 Ys = agricultural productivity by unit area of stored feed,2.0, in Kg/m2 Fm = stable element transfer coefficients, from Table E-1, Regulatory Guide 1.109, ;
I Rev.I r = fraction of deposited activity retained on cow's feed grass,1.0 for radiciodine, and 0.2 for particulates '
DFLi = thyroid ingestion dose factor for the ith radionuclide for the infant, in mrem /pCi, ,
from Table E-14, Regulatory Guide 1.109, Rev.1 Ai = decay constant for the ith radionuclide, in sec-1, from Table of Isotopes, Lederer, l Hollander, and Perlman, sixth Edition.
l Aw = decay constant for removal of activity of leaf and plant surfaces by weathering, 5.73E-07 sec-1 (corresponding to a 14 day half-life) tr = transport time from pasture to cow, to milk, to receptor,1.73+05, in seconds
( th = transport time from pasture, to harvest, to cow, to milk, to receptor,7.78E+06, in
! seconds fp = fraction of year that cow is on pasture,0.67 (dimensionless),7.78E+06 in seconds 1
fs = fraction of cow feed that is pasture grass while cow is on pasture,1.0, dimensionless V Parameters used above were obtained from NUREG-0133 and Regulatory Guide 1.109, Rev.l.
l
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 148 OF 156 ATTACHMENT 28 (Page 8 of 8)
Surry Meteorological, Liquid, and Gaseous Pathway Analysis Since the concentration of tritium in milk is based on the airborne concentration rather than the deposition, the following equation is used:
R H3 = K'K'"F,QpU,pmH3 ) _0.75(0.5/H) %/Q (28-3) where:
K'" = a constant of unit conversion IE+03 gm/kg H = absolute humidity of the atmosphere,8.0, gm/m3 0.75 = the fraction of total feed that is water 0.5 = the ratio of the specific activity of the feed grass to the atmospheric water X/Q = the annual average concentration at a location 5150 meters S sector,3.0E-07 sec/m3 for ventilation vent releases, and 1.3E-07 sec/m3 for the process vent releases Other parameters have been previously defined.
The inhalation pathway dose factors RIivy and Riipv in Attachment 18 were calculated using the following equation:
RI. = K'(BR) DFA (x/Q)(mremi
/yr per Curie /sec) i (28-4) where:
K' = a constant of unit conversion, IE+12 pCi/Ci BR = breathing rate of the infant age group,1400 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev.1 I
DFAi = thyroid organ inhalation dose factor for infant age group for the ith radionuclide, in mrem /pCi, from Table E-10, Regulatory Guide 1.109, Rev.1 X/Q = ventilation vent X/Q,3.0E-07 sec/m3, or the process vent site boundary X/Q, !
1.3E-07 sec/m3, at a location 5150 meters S sector. l l
l 1
O1 l
l
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 149 OF 156 i
ATTACHMENT 29 (Page 1 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis 1.0 METEOROLOGICAL ANALYSIS ,
1.1 Purpose The purpose of the meteorological analysis was to determine the annual average X/Q and D/Q values at critical locations around the Station for ventilation vent (ground level) and process vent (mixed mode) releases. The annual average X/Q and D/Q values were used to perform a dose pathway analysis to determine both the maximum exposed individual at site bour. dry a .d member of the pubhc. The X/Q and D/Q values resulting in the maximum exposures were incorporated into the dose factors in Attachments 13 and 19.
1.2 Meteorological Data, Parameters, and Methodology Onsite meteorological data for the period January 1,1981, through December 31,1981, were used in calculations. These data included wind speed, wind direction, and differential p temperature for the purpose of determining joint frequency distributions for those releases j d characterized as ground level (e.g., ventilation vent), and those characterized as mixed mode l (i.e., process ven0. The portions of release characterized as ground level were based on AT158.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were based on AT158.9ft-28.2ft and 158.9 ft wind data.
l X/Q's and D/Q's were calculated using the NRC computer code "XOQDOQ - Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations,"
September,1977. The code is based upon a straight line airflow model implementing the assumptions outlined in Section C (excluding Cla and Cib) of Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Lignt-Water-Cooled Reactors."
The open terrain adjustment factors were applied to the %/Q values as recommended in Regulatory Guide 1.111. The site region is characterized by gently rolling terrain so open terrain correction factors were considered appropriate. The ground level ventilation vent release calculations included a building wake correction based on a 1516 m2 containment minimum cross-sectional area.
O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 150 OF 156 ATTACHMENT 29 (Page 2 of 8)
North Anna Meteorologhal, Liquid, and Gaseous Pathway Analysis The effective release height used in mixed mode release calculations was based on a process vent release height of 157.5 ft, and plume rise due to momentum for a vent diameter of 3 in.
with plume exit velocity of 100 ft/sec. Ventilation vent, and vent releases other than from the process vent, are considered ground level as specified in Regulatory Guide 1.111 for release points less than the height of adjacent solid structures. Terrain elevations were obtained from North Anna Power Station Units 1 and 2, Virginia Electric and Power Company Final Safety Analysis Report Table 11C.2-8.
X/Q and D/Q values were calculated for the nearest site boundary, resident, milk cow, and vegetable garden by sector for process vent and ventilation vent releases at distances specified from North Anna Power Station Annual Environmental Survey Data for 1981. X/Q values were also calculated for the nearest lake shoreline by sector for the process vent and ventilation vent releases. ,
1 According to the definition for short term in NUREG-0133," Preparation of Radiological l Effluent Technical Specifications for Nuclear Power Stations," October,1978, some gaseous releases may fit this category, primarily waste gas decay tank releases and containment purges.
However, these releases are considered long term for dose calculations as past releases were both random in time of day and duration as evidenced by reviewing past release reports. ,
1 Therefore, the use of annual average concentrations is appropriate according to NUREG-0133.
The X/Q and D/Q values calculated from 1981 meteorological data are comparable to the values presented in the North Anna Power Station UFSAR.
1.3 Results The X/Q value that resulted in the maximum total body, skin and inhalation exposure for l ventilation vent releases was 9.3E-06 sec/m3 at a site boundary location 1416 meters SE sector. l For process vent releases, the site boundary %/Q value was 1.2E-06 sec/m3 at a location 1513 meters S sector. The shoreline %/Q value that resulted in the maximum inhalation exposure for ventilation vent releases was 1.0E-04 sec/m3 at a location 274 meters NNE sector. The shoreline X/Q value for process vent was 2.7E-06 sec/m3 at a location 274 meters NNE sector.
O
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 151 OF 156 l
k ATTACHMENT 29 (Page 3 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis Pathway analysis indicated that the maximum exposure from I-131,1133, and from all i
radionuclides in particulate form with half-lives greater than 8 days was through the grass-cow-l milk pathway. The D/Q value from ventilation vent releases resulting in the maximum exposure was 2.4E-09 per m2 at a location 3250 meters N sector. For process vent releases, the D/Q value was 1.1E-09 per m2 at a location 3250 meters N sector. For tritium, the X/Q value l from ventilation vent releases resulting in the maximum exposure for the milk pathway was 7.2E-07 sec/m3, and 3.9E-07 sec/m3 for process vent releases at a location 3250 meters N sector.
I 2.0 LIQUID PATHWAY ANALYSIS 1
! 2.1 Purpose The purpose of the liquid pathway analysis was to determine the maximum exposed member of the public in unrestricted areas as a result of radioactive liquid effluent releases. The analysis lV includes a determination of most restrictive liquid pathway, most restrictive age group, and critical organ. This analysis is required for Subsection 6.2.
2.2 Data, Parameters, and Methodology l
Radioactive liquid effluent release data for the years 1979,1980, and 1981 were compiled from the North Anna Power Station semi-annual effluent release reports. The data for each year, along with appropriate site specific parameters and default selected parameters, were entered into the NRC computer code LADTAP as described in NUREG-0133.
Re-concentration of effluents using the small lake connected to larger water body model was selected with the appropriate parameters determined from Table 3.5.3.5, Design Data for Reservoir and Waste Heat Treatment Facility from Virginia Electric and Power Company, Applicant's Environmental Report Supplement, North Anna Power Station, Units 1 and 2, March 15,1972. Dilution factors for aquatic foods, shoreline, and drinking water were set to one. Transit time calculations were based on average flow rates. All other parameters were defaults selected by the LADTAP computer code, i
(
i
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 152 OF 156 ATTACHMENT 29 (Page 4 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis 2.3 Results For each year, the fish pathway resulted in the largest dose. The critical organ each year was the liver, and .he adult and teenage age groups received the same organ dose. However, since the adult total body dose was greater than the teen total body dose for each year, the adult was selected as the most restrictive age group. Dose factors in Attachment 7 are for the maximum exposed member of the public, an adult, with the critical organ being the liver.
3.0 GASEOUS PATHWAY ANALYSIS 3.1 Purpose A gaseous effluent pathway analysis was performed to determine the location that would result in the maximum doses due to noble gases for use in demonstrating compliance with 6.3.1.a.
and 6.3.3.a. The analysis also included a determination of the critical pathway, location of maximum exposed member of the public, and the critical organ for the maximum dose due to 131 1 ,1133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for use in demonstrating compliance with requirements in 6.3.4.a.1. In addition, the analysis included a determination of the critical pathway, maximum age group, and sector location of an exposed individual through the inhalation pathway from 1131, Il33, tritium, and particulates with half-lives greater than 8 days to demonstrate compliance with 6.3.1.a..
3.2 Data, Parameters, and Methodology Annual average X/Q values were calculated, as described in Section 1 of this attachment, for the nearest site boundary in each directional sector and at other critical locations beyond the site boundary. The largest X/Q value was determined to be 9.3E-06 sec/m3 at site boundary for ventilation vent releases at a location 1416 meters SE direction, and 1.2E-06 sec/m3 at site boundary for process vent releases at a location 1513 meters S direction. The maximum doses to total body and skin, and air doses for gamma and beta radiation due to noble gases, would be at these site boundary locations. The doses from both release points are summed in calculations to calculate total maximum dose.
O
i VIRGINIA VPAP-2103 POWER REVISION 7 '
PAGE 153 OF 156 g
( ATTACHMENT 29 -
(Page 5 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis l i
i Step 6.3.1.a.2 dose limits apply specifically to the inhalation pathway. Therefore, the locations and WQ values determined for maximum noble gas doses can be used to determine the maximum dose from 1131, 1133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway.
The NRC computer code GASPAR," Evaluation of Atmospheric Releases," Revised 8/19/17, was run using 1979,1980 and 1981 North Anna Power Station Gaseous Effluent Release l Report data. Doses from 1133, Il33, tritium, and particulates for the inhalation pathway were calculated using the 9.3E-06 sec/m3 site boundary WQ. Except for the source term data and the j WQ value, computer code default parameters were used. Results for each year indicated that l the critical age group was the child and the critical organ was the thyroid for the inhalation l pathway, g The gamma and beta dose factors Ki vy, Livy, Mjyy, and Ni vv in Attachment 13 were obtained l d by performing a units conversion of the appropriate dose factors from Table B-1, Regulatory Guide 1.109, Rev.1, to mrem /yr per Ci/m3 or mrad /yr per Ci/m3, and multiplying by the l ventilation vent site boundary WQ value of 9.3E-06 sec/m3. The same approach was used in calculating the gamma and beta dose factors Kip y, Lip y, M ip y, and Nipv in Attachment 13 using the process vent site boundary WQ value of 1.2E-06 sec/m3, The inhalation pathway dose factors P ivy and Pipv in Attachment 13 were calculated using the l following equation:
P; = K'(BR) DFA;(x/Q) (mrem /yr per Curie /sec) (29-1) where:
K' = a constant of unit conversion,1E+12 pCi/Ci BR = the breathing rate of the child age group,3700 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev.1 DFAi = the thyroid organ inhalation dose factor for child age group for the ith radionuclide, in mrem /pCi, from Table E-9, Regulatory Guide 1.109, Rev. I q WQ = the ventilation vent site boundary WQ,5.3E-06 sec/m3, or the process vent site b' boundary UQ,1.2E-06 sec/m3, as appropriate.
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 154 OF 156 ATTACHMENT 29 (Page 6 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis Step 6.3.4.a., requires that the dose to the maximum exposed member of the public from 1131, 1133, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days be less than or equal to the specified limits. Dose calculations were performed for an exposed member of the public within site boundary unrestricted areas, and to an exposed member of the public beyond site boundary at locations identified in the North Anna Power Station Annual Environmental Survey Data for 1981.
It was determined that the member of the public within site boundary would be using Lake Anna for recreational purposes a maximum of 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br /> per year. It is assumed that this member of the public would be located the entire 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br /> at the lake shoreline with the largest annual X/Q of 1.0E-04 at a location 274 meters NNE sector. The NRC computer code G ASPAR was run to calculate the inhalation dose to this individual. The GASPAR results were corrected for the fractional year the member of the public would be using the lake.
Using the NRC computer code GASPAR and annual average %/Q and D/Q values obtained as described in Section 1 of this attachment, the member of the public receiving the largest dose beyond site boundary was determined to be located 3250 meters N sector. The critical pathway was the grass-cow-milk, the maximum age group was the infant, and the critical organ the i
thyroid. For each year 1979,1980, and 1981 the dose to the infant from the grass-cow-milk pathway was greater than the dose to the member of the public within site boundary. Therefore, the maximum exposed member of the public was determined to be the infant, exposed through ,
l the grass-cow-milk pathway, critical organ thyroid, at a location 3250 meters N sector.
Pathway analysis results indicate that existing pathways, including ground and inhalation, ;
within five miles of North Anna Power Station, yield R dose i factors less than those determined l for the cow-milk pathway. Although the cow-milk pathway does not exist within five miles of l l
the Station, NUREG-0133 requires the use of cow-milk R dose i factors since these values result in the most limiting dosec. There is no requirement to include the other pathways.
[ Commitment 3.2.3]
l 9 l
1 VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 155 OF 156 O ATTACHMENT 29 (Page 7 of 8)
North Anna Meteorological, Liquid, and Gaseous Pathway' Analysis 4 The RMivv and RMipv dose factors, except for tri:ium,in attachment 19 were calculated by multiplying the appropriate D/Q value with the following equation:
1 ff ia RM. = K'O F(U,P ) F -A i ,'
i A. + A W 1
m (r)(DFL. )
i yP s + (3 _ f f e) e-A t (29-2)
I ps y
I S
. P _ ,
1 where:
1 l K' = a constant of unit conversion, IE+12 pCi/Ci QF = cow's consumption rate,50, in Kg/ day (wet weight)
Uap = infant milk consumption rate,330 liters /yr ,
Yp = agricultural productivity by unit area of pasture feed grass,0.7 Kg/m2 l l
Ys = agricultural productivity by unit area of stored feed,2.0, in Kg/m2 Fm = stable element transfer coefficients, from Table E-1, Regulatory Guide 1.109, Rev. I r = fraction of deposited activity retained on cow's feed grass,1.0 for radioiodine, and O.2 for particulates DFL=i thyroid ingestion dose factor for the ith radionuclide for the infant, in mrem /pCi, from Table E-14, Regulatory Guide 1.109, Rev.1 Ai = decay constant for the ith radionuclide, in sec 1, from Table of Isotopes, Lederer, Hollander, and Perlman, sixth Edition.
Aw = decay constant for removal of activity ofleaf and plant surfaces by weathering, 5.73E-07 sec-1 (corresponding to a 14 day half-life) tr = transport time from pasture to cow, to milk, to receptor,1.73E+05, in seconds th = transport time from pasture, to hrdvest, to cow, to milk, to receptor,7.78E+06, in seconds fp = fraction of year that cow is on pasture,0.58 (dimensionless),7 months per year from NUREG-0597 fs = fraction of cow feed that is pasture grass while cow is on pasture,1.0, dimensionless Parameters used in the above equation were obtained from NUREG-0133 and Regulatory l Guide 1.109, Rev.l.
VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 156 OF 156 ATTACHMENT 29 (Page 8 of 8)
~
North Anna Meteorological, Liquid, and Gaseous Pathway Analysis Since the concentration of tritium in milk is based on the airborne concentration rather than the deposition, the following equation is used:
(29-3)
MH3 = K'K'"FmOF ap@ H3 ) 0.75(0.5/H) %/Q where:
K "' = a constant of unit conversion 1E+03 gm/kg H = absolute humidity of the atmosphere,8.0, gm/m3 0.75 = the fraction of total feed that is water 0.5 = the ratio of the specific activity of the feed grass to the atmospheric water X/Q = the annual average concentration at a location 3250 meters N sector,7.2E-07 sec/m3 for ventilation vent releues, and 3.9E-07 sec/m3 for the process vent releases Other parameters have been previously defined.
i l
l l
\
l r
l ATTACHMENT 4 MAJOR CHAE ES TO RADIOACTIVE LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS (01/95 -
12/95)
As required by the ODCM, Section 6.7.2.a.4, major changes to radioactive liquid, gaseous and solid waste treatment systems for the time period covered by this report are synopsized in this attachment. Supporting information as to the reason (s) for the change (s) and a summary of the 10 CFR 50.59 evaluations are included, as applicable.
The high capacity blowdown system was installed in Unit 1 in September 1995 and in Unit 2 in December 1995. It is used in conjunction with the chemical feed of the feedwater system to control steam generator chemistry.
Each unit's high capacity blowdown system consists of a flash tank, a blowdown cooler, a radiation monitor and associated piping and valves. The system discharges to the Circulating Water Discharge Tunnel through an existing discharge point.
The high capacity blowdown system is automatically terminated in the event of High-High flash tank level, High flash tank pressure, or high condenser ,
pressure. It is also automatically terminated on a containment isolation signal or a radiation monitor High-High alarm.
In the event of a high capacity steam generator radiation monitor alarm procedures require that a sample be taken and analyzed for principal gamma emitters and tritium. If the presence of any primary-to-secondary leakage is confirmed, the release via the high capacity blowdown will be terminated, and blowdown will be diverted through the normal blowdown pathway.
The high capacity blowdown system is described in NAPS UFSAR Section 10.4.6, Secondary Vent and Drain Systems.
10
~) Safety Evaluation mum u l (G
- WN ay Page 1 of 12 3g T & &, h VPAP 3001 GOV 02
^
)
- 1. Safety Evaluation Number 2. Applicable Station 3. Applicable Unit (X)NorthAnnaPowerStation [] Unit 2 5 b' kOh _ [ ] Surry Power Station (X] Unit 1
[ ] Unit 1 ( ) Unit 2 PART'AW Resolution Summary R@ cit? ~
4 List the governing documents for which this safety evaluation was performed.
OC-94-003. Steam Generator 8 lowdown System Upgrades. North Anna Unit 1
- 5. Sumar12e the cnange. test or expertment evaluated.
See Page 1A and 18 of 12.
- 6. State the purpose for this change. test. or experiment.
This change makes the ex1 sting abandoned hign capacity S/G B0 system operational without the require ent of continuous condeasate :disning for the seccqdary side of the Unit. The existing centrol Instrumentation will be replaced and the crains from tne 60 flasn tark will be routed througn a heat excnanger to recover heat anc cool the effluent prior to discnarge into the circulattng wat.er discharge tunnel. The original cesign routec the flash tant crains to tne cencenser not well. Tne original design concept was to have the 1mourities in the bicwcown removed by the condensate polisning system. Since the condensate polishing system is not continuously coerated at power, the original hign capacity DiowdOwn system, operated in the Unit's current operational configuration without continuous condensate polishing, would result in zero net blowdown.
- 7. List all limiting conditions and special requirements identified or assumed by this safety analysts. For each item. Indicate the formal tracking mechanism that will be used to ensure that these conditions and/or requirements will be met.
The installation of all the upgrades to the S/G B0 equipment and piping. except for the final tie-in of the 80 flash tank drains cooler concensate return line divert to the condenser. Including the removal of the previously installed tie-in stub weld caps, may be performed during any mode of operation with the modification being planned during non-outage time. Installation of the 80 cooler condensate return line divert to the condenser will require a Unit outage, mode 5 or 6. I Existing fire barriers fecm the Turbine Building to the ESGR and from the ESGR to the TSC will be temocrartly breeched for cable pulling during installation. Re-installation of the fire barriers will be completed upon i I
completion of the cable pulls. All work on the fire barriers shall be done in accordance with approved Station procedures. i
- 8. Will the proposed activity /Condittor, result in or constitute an unreviged 5)fety Question.
an unreviewed environmental question, a change to the Fire Protectior(Progran that affects [ ] Yes (X) No i
the ability of the station to achieve and ;raintain safe shutdown ist the event of a fire, or recuire a license amendment or Tecnnical Soecifications change 7 A )
- 9. Preparer Name (Print) 10.PreparerSigna/urel W. A. Thomas, Jr. ] 11. Date I( M.R M _Q, 9./2-95
- 12. Co n1zant Supervisor Name (Print) 13. Cognizant S ure 'V 14 0 e
- 15. Disposition
\
fyQ Approved ( ) Disapproved ( ) Approved As Modified [ ]\ Requires Further Evaluation
- 16. SNSOC Chairman Signature i 1 Date Coments l
Note: : Attach a Copy of Part A. Resolution Sumary Report;ato the Change / Activity Occumentation Package? .
l Send a' Copy- of Part^A to t.tcensing for Submittalgoithe NRC}in1A cordance With1 PAP-2802L Reporting:.
- Requ1rementSc .- . ,. . . a + ;. .w . n. - L;
,,_,,,.,..c.
LSend a copy of the complete 1 Safety Evaluation to the: Independent. Review CoordinatorMfore the MSRCh Send the completed Safety Evaluation'Originabto Records: Management;;.1, J.Z ..n '
D Use ." Safety Evaluationn Supplemental Page" Form No.5730928L ff additional: Space.is NeededJ Key: Mw. Management datety Aeview t.ocmttee rorm No. nuno (Oct 94)
OG Safety Evaluation W ht*CD-5% l Supplemental Page 1 A of 12 WtGIMA POWER VPAP 3001 GOV 02 i PART: A xResolution Sumary:ReportWtem 5. Continuation # ,
- go e W 5 The abandoned high capacity S/G B0 system will be made operational without requiring continuous condensate polishing for the secondary side of the Unit. The existing control instrumentation will be replaced and the drains from the 80 flash tank will be cooled, then rerouted to the circulating water discharge tunnel. Instead of the original design of routing the drains to the condenser hot well. The original design concept was to have the impurities in the blowdown removed by the condensate polishing system. Sincethecondensatepolishingsystem15notcontinuously operated at power, the original high capacity blowdown system, operated in the unit's current operational configuration without continuous condensate polishing, would result in zero net blowdown.
The drains from the 80 flash tank will be cooled via a new shell and tube heat exchanger (the 80 flash tank drains cooler) with condensate as the cooling medtum. Thus, energy from the flash tank drains will be recovered to the steam cycle, while cooling the BD discharge to approximately the sama temperature as the circulating water to which it is discharging. These upgrades to the high capacity 80 system will permit the system to operate at its design capacity of approximately 100.000 lb/hr (200 gpm at 60F), or 67 gpm per S/G. if sufficient makeup water capacity is available. Present makeup water capacity allows a continuous S/G BD rate of approximately 45 gpm per S/G with both Units operating.
i In addition to tre 80 flasn tank drains cooler and its associated piping. this Design (nange will also install a continuous radiation monitor and a sampling system In the cooled blowdown line upstream of tne point of discharge to 1 the circulating water discnarge tunnel. Tne point o.f discharge of the cooled blowdown effluent from the h1gn I capacity S/G B0 system 1s into the 20" - Class 121 piping dtscnarging to the circulating water discharge tunnel, i Tne low level liculd waste clarif ter effluent also discharges to this 20" pipe In the same location. Therefore, the '
extsting low capacity 5/G B0 system effluent (from the clarifier discharge) and the upgraded high capacity S/G B0 l system effluent enter the circulating water discharge tunnel at the same location. ;
The S/G 80 is monitored for radioactivity by existing radiation monitors in sample lines off of the three S/G BD lines in the Auxiliary Building. The primary function of these radiation monitors is to detect radioactivity in the 80 resulting from S/G tube leaks / ruptures. Currently the discharge of the Auxiliary Building S/G 80 line radiation monitors (normal operating flow rate is less than 1 gpm per S/G) is routed threugh Individual sample coolers, then ,
combined into one line (l") for discharge into the low capacity S/G 80 tank. and finally routed to the low level !
liquid waste system (entering the clarifier hold up tank inlet piping). Since the low capacity 80 system will serve 1 as a back-up system and is not planned for continuous operation after implementation of this Design Change, an I alternate discharge path for the 80 radiation monitors / sample coolers is included in the design. The alternate !
discharge path added by this Design Change is a 1" line teed off of the existing 1" line near the low capacity 80 '
tank, routed to bypass the 80 tank the low capacity 80 coolers, and the low capacity 80 pumps. The Unit 2 BD radiation monitors / sample coolers discharge line will also tee into the alternate discharge path when DC 95-015 is Implemented. Thus. when the low capacity system is not in operation. the S/G B0 line radiation monitors / sample coolers can be manually aligned to discharge through the alternate radiation monitors / sample coolers discharge path to the clarifter hold up tank inlet piping downstream of the low capacity BD pumps. Since these new lines are located in the Auxiliary Building in the vicinity of various safety-related equipment, they will be seismically supported non safety related lines. The current 80 radiation monitors / sample coolers discharge path (80 tanks to low capacity 80 coolers to SD pumps to clarifier hold up tank) may still be used with the hign capacity 80 system in operation. If necessary.
The new radiation monitor in the hign capacity S/G 80 system cooled blowdown effluent line (similar to the clarifier effluent radiation monitor 1-LW.RM 111) does not perform a safety function, but is included in the design as added protection against release of radioactivity to the environment. The hign capacity S/G 80 system will be automatically isolated if the Hi H1 trip setpoint of the cooled 80 effluent radiation monitor is exceeded. This radiation monitor will provide local indication as well as local alarms. The trip and Hi H1 alarm setpoints will be SE-7 pC1/cc and the H1 alar 111 setpoint will be lower to preclude a release of radioacitivity to the environment.
The high capacity 80 system as currently designed will be operated only when the Unit's steam generators have no identified primary to secondary tube leakage. Should leakage be identified at a level greater than SE-7 uCi/cc at the effluent radiation monitor, the system will isolate on the high radiation trip. Upon the trip signal all valves
.c.>o f
- tank.+W the 80 cooler outlet LCV will fall in the closed position. thus isolating 80 flow and draining the 80 flash The tank contents at the time of the trip signal will empty.to the circulating water discharge. This release will be considere an unclanned, but monitored release. Upon recet of a H1 alarm signal. Ocerations will contact Health Physics to cbtain analysts of the h1gn cap 3C1ty blowoown effluent and Ste3m Generator Blowdown streams to determine the scurce of *ne leakage. If S/G tube leakage is determined for a given steam generator during routine sampling, tne steam gene-at r shall be isolated from the hign capacity 80 system. Tne low ca:a-1ty 80 system will still be ava11ade for use or steam generators with discernable primary to secondary leakage in accordance with the recuirements of tne ODCM An alarm setootnt for the h1gn capacity BD system effluent raataticn mcnitor will be estaDlished to provide early warning of negative trends. From a maintenance and surveillance perspective, the new radiation monitor will be Consistent with the requirements for the Clarifier discharge monitor RM LW lll since it serves a similar purpose.
In the future, additional sampling will be provided for the high capacity B0 system wnich will allow the use of the system when low levels of primary to secondary leakage exists within the Current technical specification limits, rorm No. NQWe (Nov '>1)
Y O Safety Evaluation k5 l
O WRGIMA POWM SupplementalPage 18 of 12 VPAP 3001 GOV 02 PART AH Resolution Sumary ReportFItera 52Continuationa pu >
Two new divert lines to the condenser will be installed by this Design Change to help control the high capacity S/G BD system during operational periods other than 100% power (steady state). i. e.. start up, transients, or low power operation. One of the divert lines will be a 80 flash tank outlet steam divert to the condenser. and the other will be a 80 flash tank drains cooler condensate return line divert to the condenser. The steam discharge divert line 1 (including a new PCV) will be installed to tee off of the existing BD flash tank steam discharge line for directing' 1 flashed steam from the flash tank to the condenser. if additional capacity is needed or if the normal steam l discharge line to the 3rd point extraction lines is not available for use. The section of the original BD flash I tanx drain line from devstream of the original LCV to the condenser will be re-installed for the majority of the I i
piping run for this new steam divert line. The 80 flash tank drains cooler condensate return divert line will tee '
! off of the normal condensate return line. include a TCV and a check valve. then run to condenser 1-CN SC-18, i
j connecting to spare penetration number 59. The TCV in this line will open on high 80 (tubeside) outlet temperature from the 80 flash tank drains cooler. High cooler 80 outlet tcmperature is indicative of low condensate (cooling)
I flow. in most cases resulting from low main condensate flow which produces low pressure drop (motive force for the l l condensate flow to the cooler) across the condensate side of the cooler. Low main condensate flow occurs during j l times of condensate pumo recirculation operation or low power operation. Thus, these two new divert lines will l l enable the hign cacacity 80 system to be operational at full normal capacity during either condensate system upset cr icw power oce ation conditions.,
The existing UF5M (ca;e 10.4-33) references the operational concern that the Hign Cacacity Blowdown system as or1ginally designec cic not isolatt on Hi-Hi Feedwater Heater level or on Turcine tric. This concern has been corrected via tnis DCP. The 1 ES NRV-103A. 1 ES NRV-1038 Non Return valves are previoing an input to the Process l Control System. wnicn will trip 1-80-PCV 100. With 1-80-PCV-100 closed. pressure control of the system will l provided via 1 BD PCV-101. which diverts the steam from the blowdown tank to the condenser. The Non-return valves l receive trip close signals for a Turbine trip and a Hi H1 Water Level in the associated Feedwater heater. Therefore. i l the isolation of the supplemental heater steam from the high capacity 80 system is accomplished via the new control l l System functions and the original intent of the UFSAR to prevent overspeed of the turDine an inadvertant water l intrusion are accomplisned by this design change, pg l l The control system is ecuipped with a Bypass /Run function. The " Bypass" part of the Bypass /Run function will allow " #"
l l
system start up by disaoling the Lo-Lo inlet flow and level Trip signals and allow small inlet flows to warm up and l fill the system. Once tne system has warmed up and operationally defined blowdown flow rates and operationally '
! defined levels have been established, the Bypass /Run function will be placed in "Run" which will enable the trip funCtlons. Use of this function will be administratively Controlled by Operating procedures.
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form No. /JUN O thov 91) l
. . ~ ,
O Safety Evaluation camcwS%
0 VIRGIMIA POWEM Page 2 of 12 l VPAP-3001 GOV 02 I
l Part A1 Resolution Sumary Report > < m -
e l
- 18. Sumarize from Part D. Unreviewed Safety Question Determination the major issues considered; state the reason the change, test, or experiment should be allowed: and state why an unreviewed safety question does or does not exist (a simple conclusion statement is insufficient),
j This Design Change does not constitute an unreviewed safety question as defined in 10CFR50.59 since:
l 1. This modification does not affect or impact any safety related equipment or systems. Therefore, this Design I
Change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment l 1
important to safety previously evaluated in the UFSAR.
I
- 2. This Design Change is consistent with the affected systems' design bases and existing design basis criteria. The -
l systems affected by this Design Change do not include any systems important to safety or required for accident l
mitigation. Therefore, this modification does not create a possibility for an accident or malfunction of a different !
type than any previously evaluated in the UFSAR.
I
- 3. This Design Change does not impact or change the basis of any of the Technical Specifications, and, therefore. '
the margin of safety as defined in the bases of the Technical Specifications remains unchanged.
The naw steam divert and the new B0 cooler condensate return divert to tne condenser. the new flash tank drains line to tre 80 cooler. and the new B0 cooler concensate supply and return lines are additional hign energy lines that were not present in the original S/G B0 system design. The HELB analysis in the turbine building area will not be affectec. however, for the following reasons: a break in the steam divert line (as with a break in the normal steam I outlet Itne to the 3rd point extraction) is bounded by the 3rd point extraction line break analysis; a break in the l flasn tank drains line to the BD cooler will be quickly mitigated by a shutdown of the high capacity S/G B0 system i upon loss of level in the flash tank due to the break; and breaks in the BD cooler condensate supply. return, and I divert lines are all bounded by the main condensate line break analysis. The routing for the new high energy lines added by this Design Change are not in the vicinity of any safety related, or any Appendix R safe-shutdown equipment, and. thus, could not impact the aforementioned equipment during a postulated HELB.
The existing UFSAR (page 10.4-33) references the operational concern that the High Capacity Blowdown system as originally designed did not isolate on Hi-Hi Feedwater Heater level or on Turbine trip. This concern has been corrected via this DCP. The 1-ES-NRV-103A. 1 ES-NRV-1038 Non Return valves are providing an input to the Process i
Control System. which will trip 1-BD PCV-100. With 1-BD PCV-100 closed, pressure control of the system will provided I
via 1 BD-PCV-101, which diverts the steam from the blowdown tank to the condenser. The Non-return valves receive trip close signals for a Turbine trip and a Hi Hi Water Level in the associated Feedwater heater. Therefore, the isolation of the supplemental heater steam from the high capacity BD system is accomplished v1a the new control system functions and the original Intent of the UFSAR to prevent overspeed of the turbine an inadvertant water Intrusion are accomplished by this design change, g
i Tne control system is equipped with a Bypass /Run function. The " Bypass" part of the Bypass /Run function will allow 8E system start up by disabling the Lo-Lo inlet flow and level Trip signals and allow small inlet flows to warm up and fill the system. Once the system has warmed up and operationally defined blowdown flow rates and operationally definec levels have been established, the Bypass /Run function will be placed in "Run" which will enable the trip functions. Use of this function will be administratively controlled by Operating procedures.
The modifications to the high capacity S/G B0 system (non-safety related system) installed by this Design Change will have no pnysical effect on levels of radiation or airborne radioactivity during installation, since none of the work will be done in the RCA. During operation if the radiation monitors detect radioactivity in the S/G BD. the S/G BD system will be isolated prior to potential for GOC-19 concerns.
.The high capacity 80 system as currently designed will be operated only when the Unit's steam generators have no identified primary to secondary tube leakage. Should leakage be identified at a level greater than SE-7 pC1/cc at the effluent radiation monitor, the system will isolate on the high radiation trip. Upon the trip signal all valves i
w e/M the 80 cooler outlet LCV will fail in the closed position. thus isolating 80 flow and draining the 80 flash tank. The tank contents at the time of the trip signal will empty to the circulating water discharge. This release will be consioered an unplanned, but monitored release. Upon recesot of a Hi alarm signal. Ooerations will contact Health Physics to obtain analysis of the hign capacity blowdM effluent and Steam Generator Blowdown streams to l
' determine the source of the leakage. If S/G tube leakage is determined for a given steam generator during routine samcling. the steam generator shall be isolated from the high capacity BD system. The low Capacity 80 system will still be available for use on steam generat0rs with discernable prtmary to secondary leakage in accordance with the re:Jirements of the 00CM. An alarm setpcInt for the high caoacity BD system effluent radiation monitor will be estacitsned to provide early warning of negative trends. From a maintenance and surveillance perspective, the new radiation monitor will be consistent with tne requirements for the clarif ter discnarge monitor. RM LW-111. since it serves a similar purpose.
In the future. additional sampling will be provided for the high capacity BD system which will allow the use of the system when low levels of primary to secondary leakage exists within the current technical specification limits.
Additionally, the new S/G B0 effluent dischar"e locations resulting from this Design Change are being included in the new Virginia Pollution Discharge Elimination system (VPDES) permit for North Anna Power Station.
rorm No. uuno (uct v4)
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Safety Evaluation MG* MOM
. Page 1 of 12 WMW \
\ l l VPAP-3001 GOV 02
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- 1. Safety Evaluation Nurber 2. Applicable Station 3. Applicable Unit (X) North Anna Power Station [ ] Unit 1 (X) Unit 2 l hb MNM ( ) Surry Power Station [ ] Unit 1 [ ] Unit 2 PARTl A% ResoWeionikumary Report:jgiMMM MGWCW;WWR$ @@ 4 - ^
- sglMQ
- 4. List the governing documents for e lch this safety evaluation was performed. j DC-94-015, steam Generator Blowdown System Upgrades, North Anna Unit 2
- 3. Sunnerize the change, test, or experiment evaluated.
See Page 1A and 15 of 12.
- 6. State the purpose for this change, test, or experiment.
This change makes the existing abandoned high capacity S/G B0 system operational kithout the requirenent of continuous condensate polishing for the secondary side of the Unit. The existint control instrunentation will be replaced and the drains from the 80 flash tank will be routed through a heat exchanger to recover heat ard cool the effluent prior to discharge into the circulating water discharge tunnel. The original design routed the flash tank drains to the condenser hot well. The original design concept was to have the lipurities in the blowdown removed by the condensate polishing system. $1nce the condensate polishing system is not continuously operated at power, the original high capacity blowdown system, operated in the Unit's current operational configuration without continuous condensate' polishing, would result in zero net blowdown.
- 7. List all limiting conditions and soecial requirements identified or assuned by this safety analysis. For each item, Indicate the formal tracking mechanism tnet will be ue=d to ensure that these conditions and/or requirements will be met.
- The installation of all the Lpgrades to the S/G 80 equipment and piping, incitajing the removal of the previously installed tie in stub weld caps, may be performed during any mode of operation with the modification being planned during non-outage time, e
- $ 0 M W J- CM172KS AM h!Gouav 77 66 w R.,4cs. AVoA ; v M.40*% Jb729*r a SS&Kd. WMQ l ili. Will the proposed activity / condition result in or constitute an unreviewed safety question, #'~ '
an unreviewed environmental question, a change to the Fire Protection Program that affects t ] Yes (Xj No l the ability of the station to achieve and maintain safe shutdown in the event of a fire, {
or require a license amendnent or Technical Specifications change? '
- 9. Preparer Name (Print) er r ig R. J. Atkinsom .
_. 11.Cate[yf""
,>b
- 12. Cognizant Supervisor Name (Print) 13. Co,gpire,nt (ser: Si 14. Date A/,T Atd h . M^ A gnature C
/2/S/ff
- 15. Disposition (74 Approved ( ) Disapproved [ ] Approved As Modified [] Requires Further Evaluation
- 16. SNSOC Chairman Signature 17. Da e comments Notet f Attach a. Copy of:.Part A,. Resolution Sunmary Report, to the change / Activity Doctanentation Package.
. Send a Copy of Part; A'to Lic.ensinglforlSutmittaljtn;the,NRC;in Accordance Witft VPAP-2802,LReportinga EtEequiraments@ . . . . , . , , . . ~ . . . ,
...m. ..e.,.,s . , , , . . . . . , . . , , . , . , . .. , , . . , . . . . . . .,
s Send eTcopyfof the~ ccepletedL5efetREvol.;.. .ustion;totthejindependentiReview Coordinator 1for;.;the MSRcn tsend' the. completed:$afety Evaluation'originaltto . Records' Management 4 ., ... ... ....s,...,,.,,
' EUses" Safety Evaluation?SupplementafPege9erts No2:730928Fifiedditional.:;Spacesis' Weeded.1 Kry: MSRC-Management Saf ety Review Cormiittee form No. 730916 (Oct 94)
M s. ,
~\ Safety Evaluation y-psc% l (O
vmamwowa SupplementalPage 1A of 12
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VPAP-3OO1 GOVO2 I
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- PAaTlAMeesolutionfstannary Naport?ItanMCantivastlee WWNMWi&@M YN hM N ,
! l The abandoned high capacity $/G B0 system will be made operational without requiring continuous condensate polishing for the secondary aide of the Unit. The existing control instrumentation will be replaced and the drains from the 80 flash tank will be cooled, then rerouted to the circulating water discharge tumel, instead of the original design of routing the drains to the condenser hot well. The original design concept was to have the igurities in the blowdown removed by the condensate polishing system. $1nce the condensate polishing system is not continuously operated at power, the original high capacity blowdown system, operated in the Unit's current operational configuration without continuous condensate polishing, would result in zero net blowdown.
The drains from the 80 flash tank will be cooled via a new shell and ttbe heat exchanger (the 80 flash tank dra. ins cooler) with condensate as the cooling medius. Thus, energy from the flash tank drains will be recovered to the steam cycle, while cooling the 80 discharge to approximately the same temperature as the circulating water to which it is discharging. These ggrades to the high capacity 80 systet will pensit the system to operate at its design capacity of approximately 100,000 lb/hr (200 gpm at 60F), or 67 gpn per S/G, if sufficient makeup water capacity is avaltable. Present makeup water capacity allows a continuous S/G B0 rate of approximately 45 spn per S/G with both
! Units operating.
! In addition to the 80 flash tank drains cooter and its associated piping, this Design Change will also install a continuous radiation mnitor and a saapling system in the cooled blowdown line upstream of the point of discharge to the circulating water discharge tunnel. The point of discharge of the cooled blowdown effluent from the high capacity S/G BD system is into the 20" Class 171 piping discharging to the circulating water' discharge tunnel.
The low level liquid waste clarifier effluent also discharges to this 20" pipe in the same location. Therefore, the existing low capacity S/G BD system effluent (from the clarifier discharge) and the upgraded high capacity $/G B0 !
system ef fluent enter the circulating water discharge tunnel at the same location. l The S/G BD is monitored for radioactivity by existing radiation monitors in samle lines off of the three S/G B0 lines in the Auxiliary Building. The primary fmetion of these radiation monitors is to detect radioactivity in the 80 resulting from S/G tube Leaks / ruptures. Currently the discharge of the Auxiliary Building S/G B0 line radiation
! monitors (normal operating flow rate is less than 1 spm per S/G) is routed through indivicbal sagte coolers, then conbined into one line (1") for discharge into the low capacity $/G B0 tank, and finally routed to the low level i
' liquid waste syste:n (entering the clarifier hold up tank inlet piping). Since the low capacity 80 system will serve as a back g system and is not planned for continuous operation af ter iglementation of this Design Change, an alternate discharge path for the 80 radiation monitors /sagte coolers is included in the design. The alternate
, discharge path added by this Design Change is a 1" line teed off of the existing 1" line near the low capacity 80 tank, routed to bypass the B0 tank, the low capacity B0 coolers, and the low capacity B0 pwps. The Unit i BD
' radiation monitors /sagte coolers discharge line is also teed into the alternate discharge path (by DC 94 003).
Thus, when the low capacity system is not in operation, the S/G B0 line radiation monitors /samle coolers can be manually aligned to discharge through the alternate radiation monitors /sagte coolers discharge path to the clarifier hold up tank intet piping downstream of the low capacity B0 pmps. Since these new lines are located in the Auxillery Building in the vicinity of various safety related eQJipment, they will be seismically supported non-safety related lines. The current 80 radiation monitors / sample coolers discharge path (80 tanks to low capacity B0 cooters to 80 pwps to clarifier hold up tank) may still be used with the high capacity B0 system in operation, if necessary.
The new radiation unitor in the high capacity S/G B0 system cooled blowdown ef fluent line (similar to the clarifier ef fluent radiation monitor,1-LW RM 111) does not perform a safety fmetion, but is included in the design as added protection against release of radioactivity to the envirornent. The high capacity S/G B0 system will be
- automatically isolated if the Hi-Hi trip setpoint of the cooled 80 effluent radiation monitor is exceeded. This radiation monitor will provide local indication as well as local alarms. The trip and Hi Hi alarm setpoints will be 5E-7 ACl/cc and the Hi alarm setpoint will be lower to preclude a release of radioactivity to the envirorment.
The high capacity 80 system as currently designed will be operated only when the Unit's steam generators have no identified primary to secondary tube leakage. Should leakage be identified at a level greater than 5E 7 uCf/cc at i
the effluent radiation monitor, the system will isolate on the high radiation trip. Upon the trip signal all valves except the 80 cooler outlet LCV will fait in the closed position, thus isolating B0 flow and draining the B0 flash tank. The tank contents at the time of the trip signal will enty to the circulating water discharse. This release will be considered an unplanned, but monitored release. Upon receipt of a Hi alarm signal, Operations will contact Health Physics to obtain analysis of the high capacity blowdown effluent and Steam Generator Blowdown streams to determine the source of the leakage. If S/G t @e leakage is determined for a given steam generator during routine sa@ ling, the steam generator shall be isolated from the high capacity 80 system. The low capacity 80 system will still be available for use on steam generators with discernable primary to secondary leakage in accordance with the requirements of the 00CM. An alarm setpoint for the high capacity 80 system ef fluent radiation monitor will be established to provide early warning of negative trends. From a maintenance and surveillance perspective, the new radiation monitor will be consistent with the requirements for the clarf fler discharge monitor, RM LW 111, since it serves a similar purpose.
In the future, acMitional sagting will be provided for the high capacity 80 system which will allow the use of the l system when low levels of primary to secondary leakage exists within the current technical specification limits.
(* Safety Evaluation p.y
\ Page 2 of 12 w=uwown w Stgw VPAP 3001 GOV 02 i .
PartLAMResolutfors stammary; Reports JE d%p > <
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f l 18. Sumarize from Part D, Unreviewed Safety Question Determination, the major issues considered; state the reason the l change, test, or experiment should be allowed; and state why an unreviewed safety question does or does not exist (a sigle conclusion statement is insuf ficient).
This Design Change does not constitute en mreviewed safety question as defined in 10CFR50.59 since:
- 1. This modification does not affect or impact any safety related equipment or systems. Therefore, this Design l Change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipnent l I g ortant to safety previously evaluated in the UFSAR.
- 2. This Design Change is consistent with the affected systems' design bases and existing design basis criteria. The systems af fected by this Design Change do not incitz$e any systems leportant to safety or required for accident citigation. Therefore, this modification does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR.
l 3. This Design change does not inpact or change the basis of any of the Technical Specifications, and, therefore, the margin of safety as defined in the bases of the Technical Specifications remains mchanged.
l The new steam divert and the new B0 cooler condensate return divert to the condenser, the new flash tank drains line to the 80 cooler, and the new B0 cooler condensate suply ard return lines are additional high energy lines triat were not present in the original S/G BD system design. The HELB analysis in the turbine building area wiLL not be affected, however, for the following reasons: a break in the steam divert line (as with a break in the normal steam outlet line to the 3rd point extraction) is bomded by the 3rd point extraction line break analysis; a break in the flash tank drains line to the 80 cooler will be wickly mitigated by a shutdown of the high capacity S/G BD system won loss of level in the flash tank due to the break; and breaks in the 80 cooler condensate s@ ply, return, and divert lines are all bounded by the main condensate line break analysis. The routing for the new high energy lines added by this Design Change are not in the vicinity of any safety-related, or any Appendix R safe-shutdown equipnent, and, thus, could not impact the aforementioned e wipment & ring a postulated HEiB.
The existing UFSAR (page 10.4-33) references the operational concern that the High Capacity Blowdown system as originally designed did not isolate on Hi HI Feedwater Heater level or on Turbine trip. This concern has been j corrected via this DCP. The 2 ES-NRV 203A, 2 ES NRV 2038 Non Return valves are providing an input to the Process Control System, which will trip 2 00 PCV 200. With 2 80 PCV 200 closed, pressure control of the system will provided vie 2-BO PCV 201, which diverts the steam from the blowdown tank to the condenser. The Non return valves receive trip close signals for a Turbine trip and a Hi-Hi Water Level in the associated Feedwater heater. Therefore, the isolation
, of the supplemental heater steam from the high capacity 80 system is accorrplished via the new control system
( functions and the original intent of the UFSAR to prevent overspeed of the turbine an inadvertent water intrusion are accortplished by this design change.
The control system is equimed with a Bypass /Run fmetion. The " Bypass" part of the Bypass /Rm function will allow system start @ by disabling the Lo-Lo inlet flow and the Lo Lo level Trip signals and allow small flow rates to warm up the system. Once the system has warmed w and operationally defined blowdown flow rates have been established, the l Bypass /Run function will be placed in "Rm" which will enable the trip functions. Use of this function will be administratively controlled by Operating procedures.
The modifications to the high capacity S/G B0 system (non-safety related system) installed by this Design Change will have no physical ef fect on levels of radiation or airborne radioactivity daring installation, since none of the work will be done in the RCA, except for the installation of a short (less than 10' length) section of 1" pipe in the Aux.
Bldg. Work done in the Aux. Bldg. will be performed in the low general dose rate areas. The nutter of man hours expected to cormlete the installation is minimal. Therefore, the accunulated exposure is expected to be minimal and no further calculation for personnel exposure is required. During operation if the radiation monitors detect radioactivity in the S/G B0, the S/G BD system will be isolated prior to potential for GDC-19 concerns.
The high capacity 80 system as currently designed will be operated only when the Unit's steam generators have no identified primary to secondary tube leakage. Should leakage be identified at a level greater than SE 7 uCi/cc at the ef fluent radiation mnitor, the system will isolate on the high radiation trip. Upon the trip signal all valves except the 80 cooler outt tt LCV will f ail in the closed position, thus isolating 80 flow and draining the BD flash tank. The tank contents at the time of the trip signal will ecoty to ths circulating water discharge. This release will be considered an unplanned, but monitored release. Upon receipt of a Hi alarm signal, Operations will contact Health Physics to obtain analysis of the high capacity blowdown effluent and Steam Generator Blowdown streams to determine the source of the leakage. If S/G ttbe leakage is determined for a given steam generator during routine samling, the steam generator shall be isolated frm the high capacity 80 system. The low capacity 80 system will still be available for use on steam generators with discernable primary to secondary leakage in accordance with the requirements of the 00CM. An alarm setpoint for the high capacity 80 system ef fluent radiation monitor will be sstablished to provide early warning of negative trends. From a maintenance and surveillance perspective, the new radiation monitor will be consistent with the requirements for the clarifier discharge monitor, RM LW 111, since it
- serves a similar purpose.
1 Form No. 730916 (Oct 94) l
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ba :ms Safety Evaluation qHvwn SupplementalPage 18 of 12 VPAP-3001 GOV 02
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PART?Aisasso(istiertasanary Repoet% Item 5MCantfaustianW /1MMM . n"# s S N L , & y
~ , ~ c Two new divert lines to the condenser will be installed by this Design Change to help control the high capacity S/G BD system during operational periods other than 100% power (steady state), 1. e., start up, transients, or low power operation. One of the divert lines will be a 30 flash tank outlet steam divert to the condenser, and the other will be a 30 flash tank drains cooler condensate return line divert to the condenser. TFr steam discharge divert line (including a new PCV) will be installad to tee off of the existing 30 flash tank steam discharge line for directing flashed steam from the flash tart to the condensu, if adt tional capacity is needed, or if the normal steam discharge line to the 3rd point extisetion ifnes is not available for use. The section of the original 80 flash tank drain line from downstream cf the original LCV to the condenser will be re installed for the majority of the piping rm for this new steam divert u w The 80 flash tank drains cooler condensato return divert line will tee off of the normal condeosate return line, include a TCV ard a check valve, then run to condenser 2-CN SC 1B, '
connecting to spare pr.wtration nurber 57. The TCV in this line will open on high 80 (tubeside) outlet tanperature from the 30 flash tarx drains cooler. High cooler 30 outlet temperature is indicative of low condensate (cooling) flow, in most cases amulting from low main condensate flow, which prockJces low pressure drop (motive force for the 1 condensate flow to the :ooter) across the condensate side of the cooler. Low main condensate flow occurs during l times of cendensate puns recirculation operation or low power operation. Tcus, these two new divert lines will enable the high capacity B0 system to be operational at full normal capacity during either condensate system upset or low power operation conditions.
U.e existing UFSAR (page 10.4 33) references the operational concern that the High Capacity Blowdown system as originally designed did not isolate on Hi Hi Feedwater Hester level or on Turbine trip. This concern has been corrected via this DCP. The 2-ES NRV 203A, 2-ES-NRV 2038 Non Return valves are providing an input to the Process control System, which will trip 2 BO-PCV-200. With 2 BO PCV-200 closed, pesssure control of the system will provided via 2-BD-PCV-201, which diverts the steam from the blowdown tank to the condenser. The Non-return valves receive trip close signals for a Turbine trip and a Hi Hi Water Level in the associated Feedwater heater. Therefore, the isolation of the swplemental hester steam from the high capacity 50 system is accomplished via the new control system fmettons and the original intent of the UFSAR to prevent overspeed of the turbine an inadvertent water intrusion are acconplished by this design change.
The control system is e4Jipped with a Bypass /Rm function. The " Bypass" part of the Bypass /Run function will allow system start up by dir,abling the Lo Lo intet flow and Lo-Lo level Trip signals and allow small flow rates to warm up t' e system. Once the system has warmed up and operationally defined blowdown flow rates have been established, the Bypass /Rm fmetion will be placed in "Run" which will enable the trip functions. Use of this functicn will be administratively controlled by Operating procedures.
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Form No. 730928 (Nov 91)
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ATTAunsGDR 5 INOPERABILITY OF RADIOACTIVE LIQUID AND GASEOUS l
EFFLUENT MONITORING INSTRIPNMTATIQ/{
i (01/95 -
12/95)
As required by the ODCM, Sections 6.2.2.b.2 and 6.3.2.b.3, a list and explanation for extended inoperability of radioactive liquid and/or gaseous effluent monitoring instrumentation is provided in this attachment.
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l No extended periods of inoperability occurred with any of the Liquid or Gaseous Effluent Monitoring Instrumentation specified in the ODCM, Attachments 2 and 15, for the time period covered by this report.
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- ATTACEMENT 6 UNPLANNED RELEASES 1
4 (01/95 -
12/95) 1 i
As required by the ODCM, Section 6.7.2.a.3, a list of unplanned releases, 4
l from the site to unrestricted areas, of rad,icactive material in gaseous and liquid effluents occurring during the reporting period, is made in this
, attachment.
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. No unplanned releases, ac defined by the criteria presented in the ODCM, Section 6.7.2.a.3, occurred during the time period covered by this report.
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ATTACHMENT 7 LOWER LIMITS OF DETECTION FOR EFFLUENT SAMPLE ANALYSIS (01/95 -
12/95)
Gaseous Effluentq:
i Required L.L.D. Typical L.L.D.
__EAdioisotone (uci/ml) (uci/ml)
Krypton - 87 1.00E-4 4.40E-8 -
3.98E-7 Krypton - 88 1.00E-4 7.09E-8 -
5.70E-7 Xenon - 133 1.00E-4 4.48E-6 -
3.23E-7
, Xenon - 133m 1.00E-4 1.51E-7 -
1.38E-6 Xenon - 135 1.00E-4 1.79E-8 -
1.50E-7 Xenon - 135m 1.00E-4 6.84E-8 -
5.83E-7 Xenon - 138 1.00E-4 1.64E-7 -
1.96E-6 Iodine - 131 1.00E-12 5.52E 7.59E-14 Mancanese - 54 1.00E-11 3.64E 5.32E-14 Cobalt - 58 1.00E-11 4.05E 5.21E-14 Iron - 59 1.00E-11 7.84E 1.01E-13 Cobalt - 60 1.00E-11 5.40E 1.52E-13 Zj.nc - 65 1.00E-11 9.09E 1.19E-13 Strontium - 89 1.00E-11 4.00E 5.00E-15 Strontium - 90 1.00E-11 7.00E 1.00E-15 Molybdenum - 99 1.00E-11 2.72E 3.47E-13 l l
Cesium - 134 1.00E-11 5.32E 1.61E-13 Cesium - 137 1.00E-11 4.64E 5.77E-14 ,
1 Cerium - 141 1.00E-11 4.54E 6.79E-14 Cerium - 144 1.00E-11 2.04E 3.22E-13 Gross Aloha 1.00E-11 6.90E 1.20E-14 Tritium 1.00E-6 1.12E-7 -
1.37E-7 13
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ATTACHMENT 7 LOWER LIMITS OF DETECTION FOR EFFLUENT SAMPLE ANALYSIS (01/05 -
12/95) )
Liouid Effluents:
Required L.L.D. Typical L.L.D.
Radioisotone (uci/ml) (uci/ml) 1 Kryoton - 87 1.00E-5 5.93E-8 -
7.29E-8 Krvoton - 88 1.00E-5 9.38E-8 -
1.16E-7 Xenon - 133 1.00E-5 6.08E-8 -
8.83E-8 Xenon - 133m 1.00E-5 2.04E-7 -
2.69E-7 Xenon - 135 1.00E-5 2.81E-8 -
3.14E-8 Xenon - 135m 1.00E-5 8.92E-8 -
1.19E-7 Xenon - 138 1.00E-5 2.17E-7 -
3.41E-7 ___
Iodine - 131 1.00E-6 2.63E-8 -
3.43E-8 Manganese - 54 5.00E-7 2.45E-8 -
3.39E-8 h
Iron - 55 1.00E-6 9.70E-9 -
8.00E-7 l
l Cobalt - 58 5.00E-7 2.61E-8 -
3.25E-8 Iron - 59 5.00E-7 5.04E-8 -
5.86E-8 l Cobalt - 60 5.00E-7 2.86E-8 -
8.43E-8 l Zinc - 65 5.00E-7 5.81E-8 -
6.83E-8 Strontium - 89 5.00E-8 3.00E-8 -
5.00E-8 l Strontium - 90 5.00E-8 5.00E-9 -
1.00E-8 Molybdenum - 99 5.00E-7 1.82E-7 -
2.03E-7 l Cesium - 134 5.00E-7 3.59E-8 -
1.00E-7 Cesium - 137 5.00E-7 3.32E-8 -
3.79E-8 Cerium - 141 5.00E-7 4.07E-8 -
5.47E-8 l
l Cerium - 144 5.00E-7 1.81E-7 -
2.63E-7 1 Gross Alpha 1.00E-7 2.15E-8 -
3.72E-8 Tritium 1.00E-5 2.78E-6 -
3.32E-6 14 l