Text

Annual Radioactive Effluent Release Rept Surry Power Station (Jan-Dec 1995). W/960422 Ltr
	ML18151A299
Person / Time
Site:	Surry
Issue date:	12/31/1995
From:	Blount P, Bowling M; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	96-184, NUDOCS 9604300410
	Download: ML18151A299 (188)
	v • d • e

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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 22, 1996 United States Nuclear Regulatory Commission Serial.No.96-184 Attention: Document Control Desk NURPC Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Enclosed is the Surry Power Station Annual Radioactive Effluent Release Report for January 1, 1995 through December 31, 1995. The report, submitted pursuant to Surry Power Station Technical Specification 6.6.B.3, includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released during the 1995 calendar year, as outlined in Regulatory Guide 1.21, Revision 1, June 1974.

Very truly yours,

$~~

M. L. Bowling, Manager Nuclear Licensing and Operations Support Attachment cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station 300125

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ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT FOR JANUARY 1, 1995 THRU DECEMBER 31, 1995 REC'D W/LTR DTD 4/22/96 ... 9604300410

- NOTICE -

THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE INFORMATION &

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- NOTICE -

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SURRY POWER STATION (January 1, 1995 Through December 31, 1995)

Prepared By:l~-1-. ?ZL-r Staff Health Physicist Reviewed By: )>IC L0ilL---

Supervisor

/

RaJdiolo/C:~IFing

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Reviewed By: _/;:---;-'.~ ~ / ~ * -

Supervisor Health Physics Technical Services Approved By: t ~v Id-Superintendent Radiological Protection

1-FORWARD This report is submitted as required by Appendix A to Operating License No. 's DPR-32 and DPR-37, Technical Specifications for Surry Power Station, Units 1 and 2, Virginia Electric and Power Company, Docket No.'s 50-280, 50-281, Section 6.6.B.3 .

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT FOR THE SURRY POWER STATION (January 1, 1995 Through December 31, 1995)

Index Section No. Subject Page 1 Executive Summary 1 2 Purpose and Scope 2 3 Discussion 3 & 4 4 Supplemental Information 5 Attachment 1 Effluent Release Data Attachment 2 Annual and Quarterly Doses Attachment 3 Revisions to Offsite Dose Calculation Manual {ODCM)

Attachment 4 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems Attachment 5 Inoperability of Radioactive Liquid and Gaseous Effluent Instrumentation Attachment 6 Unplanned Releases Attachment 7 Lower Limit of Detection {LLD) for Effluent Analysis

ATTACHMENT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SURRY POWER STATION UNITS 1 AND 2 LICENSE NOS. DPR-32 AND DPR-37 L

EXECUTIVE

SUMMARY

ANNUAL RADIOLOGICAL EFFLUENT RELEASE REPORT The Annual Radiological Effluent Release Report describes the radiological effluent control program conducted at the Surry Power station during the 1995 calendar year. This document summarizes the quantities of radioactive liquid and *gaseous effluents and solid waste released from the Surry Power Station in accordance with R. G. 1.21 and includes an assessment of radiation doses to the maximum exposed member of the public due to the radioactive liquid and gaseous effluents.

There were no unplanned liquid or gaseous effluent releases classified according to the criteria in the Offsite Dose Calculation Manual during this reporting period.

Based on the 1995 effluent release data, Appendix I dose calculations were performed in accordance with *the Offsite Dose Calculation Manual. The dose calculations are as follows:

1. The total body dose due to liquid effluents was 3.85E-04 mrem, which is 6.42E-03 % of. the dose limit. The critical organ doses due to liquid effluents, GI-LLI and Thyroid respectively, were 4.17E-04 mrem and 1.81E-04 mrem. These doses are 2.09E-03 % and 9.05E-04 % of the respective dose limit.

2. The air dose due to noble gase 9 was 1.13E-Ol mrad gamma, which is 5.65E-01% of the annual gamma dose limit, and 3.38E-01 mrad beta, which is 8.45E-1 % of the annual beta dose limit.

3. The critical organ dose for I-131, I-133, H-3, and particulates with half-lives greater than 8 days is 4.41E-02 mrem, which is 1.47E-01 % of the annual dose limit.

There were no major changes to radioactive liquid, gaseous, and solid waste treatment systems during this reporting period.

There were changes to the Offsite Dose Calculation Manual, VPAP-2103, during this reporting period which were implemented in Revision 7 on October 31, 1995. Attachment 3 provides the changes to VPAP-2103.

Based on the levels of radioactivity measured during this reporting period and the dose calculations performed, the operations of the Surry Nuclear Power Station, Units 1 and 2, have resulted in negligible radiation dose consequences to the maximum exposed member of the public in unrestricted areas.

1

Purpose and scope The Radioactive Effluent Release Report, Attachment 1, includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste as outlined in 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", Revision 1, June 1974, with data summarized on a quarterly basis following the format of Tables 1, 2 and 3 of Appendix B thereof. Attachment 2 of this report 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. Additionally a list of unplanned releases during the reporting period is included in Attachment 6.

0 As required by Technical Specification 6.8.B, chai'lges to the Offsite Dose Calculation Manual (ODCM) for the time period covered by this report are inclu~ed in Attachment 3.

Major changes to the radioactive liquid, gaseous and solid waste treatment systems are reported in Attachment 4, as required by the ODCM, Section 6.7.2. If changes are made to these systems, the report shall include information to support the reason for the change and a summary of the 10CFR50.59 evaluation. 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 explanation for the inoperability of radioactive liquid and/or gaseous effluent monitoring instrumentation is provided in Attachment 5 of this report.

2

Discussion The basis for the calculation of the percent of technical specification for the critical organ in Table lA of Attachment 1 is the ODCM section, 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-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, inhalation pathway.

The basis for the calculation of the percent of technical specification for the total body and skin in Table lA 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 mR.em/yr to the skin.

The basis for the calculation of the percent of technical specification in Table 2A of Attachment 1, is the ODCM, section 6.2.1, which states that the concentration of radioactive material releases in liquid effluents to unrestricted areas shall not exceed ten times the concentrations specified in 10CFR20, 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 2.00E-04 microcuries/ml.

Percent of technical specification calculations are based on the total gaseous or liquid effluents released for that respective quarter. '

The annual and quarterly doses, as reported in Attachment 2, were 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 the site boundary. The maximum exposed member of the public from the release of airborne iodine-131, iodine-133, tritium and all radionuclides in particulate form with half lives greater than 8 days, is defined as an infant, exposed through the grass-cow-milk pathway, 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 adult, exposed by either the invertebrate or fish pathway, with the critical organ being either the thyroid gland or gastrointestinal-lower large intestine. The total body dose was also determined for this individual.

3

1-~-

Presented in Attachment 6 is a list of unplanned gaseous and liquid releases as required by the ODCM, Section 6.7.2.

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 volumes 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 the radioisotope was reported as Not Detected (N/D) on Attachment 1 of this report. When all isotopes listed on Attachment 1 for a particular quarter and release mode are less than the lower level of detection, then the totals for this period will be designated as Not Applicable (N/A).

4 L

supplemental Information Section 6.6.1.b.4 of-the ODCM requires the identification of the cause(s) for the unavailability of milk or if required, leafy vegetation samples, and' the identification for obtaining replacement samples. As all milk samples were available for collection during this reporting period, leafy vegetation sampling was not required.

As required by the ODCM, section 6.6.2, evaluation of the Land Use Census is made to determine if new sample location(s) must be added to the Radiological Environmental Monitoring Program. Evaluation of the Land Use Census conducted for this reporting period identified no change in sample locations for the radiological environmental monitoring program

5

Attachment 1 EFFLUENT RELEASE DATA (January 1, 1995 Through December 31, 1995)

This attachment includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste as outlined in Regulatory Guide 1.21, Appendix B

Attachment 1 TABLE 1A Page 1 of 12

EFFLUENT AND W.ASTE DISPOSAL .ANNUAL REPORT PERIOD 1/1/9S TO 12/31/9S GASEOUS EFFLUENT-suMMATION OF .ALL RELEASES SURRY POWER STATION UNITS 1&2 UNIT FIRST SECOND %EST.ERROR QUARTER QUARTER A. FISSION & ACTIVATION GASES

1. TOTAL RELEASE Ci 1.36E+OO 2.41E+OO 1.80E+Ol

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 1.75E-01 3.07E-01 B. IODINE

1. TOTALI-131 Ci 1.42E-04 7.74E-05 2.80E+Ol

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 1.82E-05 9.84E-06 C. PARTICULATE

1. HALF-LIFE >8 DAYS Ci 5.34E-05 3.SSE-05 2.80E+Ol

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 6.87E-06 4.89E-06

3. GROSS ALPHA RADIOACTIVITY Ci NiD N/D D. TRITIUM

1. TOTAL RELEASE Ci i 6.03E+OO 2.13E+OO 3.lOE+Ol

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 7.76E-01 2.71E-01 PERCENTAGE OF T.S. LIMITS CRITICAL ORGAN DOSE RATE* % 4.70E-03 1.92E-03 TOTAL BODY DOSE RATE % 5.30E-05 2.66E-03 SKIN DOSE RATE % 1.53E-05 9.65E-04

Attachment 1 TABLE 1A Page 2 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

SURRY POWER STATION UNITS 1&2 PERIOD 1/1/95 TO 12/31/95 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES UNIT THIRD FOURTH %EST.ERROR QUARTER QUARTER A. FISSION & ACTIVATION GASES

1. TOTAL RELEASE Ci 1.87E+02 3.64E+01 1.SOE+Ol

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 2.3SE+01 4.58E+OO B. IODINE

1. TOTAL 1-131 Ci 1.54E--03 4.41E--04 2.80E+01

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 1.93E--04 5.54E--05 C. PARTICULATE

1. HALF-LIFE > 8 DAYS Ci 5.66E--05 1.4SE--05 2.80E+01

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 7.12E--06 1.83E--06

3. GROSS ALPHA RADIOACTIVITY Ci N/D N/D D. TRITIUM

1. TOTAL RELEASE Ci 3.86E+OO 4.64E+OO 3.10E+01

2. AVE RELEASE RATE FOR PERIOD µ.Ci/sec 4.86E--01 5.83E--01 PERCENTAGE OF T.S. LIMITS CRITICAL ORGAN DOSE RATE % 1.49E--02 6.34E--03 TOTAL BODY DOSE RATE % 7.30E--02 4.42E--04 SKIN DOSE RATE % 2.93E--02 2.SlE--04

Attachment 1 TABLE1B Page 3 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

PERIOD: 111/95 TO 12131/95 GASEOUS EFFLUENTS-MIXED MODE RELEASES

CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT FIRST SECOND FIRST SECOND QUARTER QUARTER QUARTER QUARTER

1. FISSION & ACTIVATION GASES Kr-85 Ci N/D N/D N/D N/D Kr-85m Ci N/D N/D 2.60E-04 N/D Kr-87 Ci N/D N/D N/D N/D Kr-88 Ci N/D NID N/D N/D Xe-133 Ci 1.70E-02 1.61E-02 1.26E+OO 1.38E+OO Xe-135 Ci 1.79E-02 5.20E-02 5.24E-02 9.40E-03 Xe-135m Ci N/D N/D N/D N/D Xe-138 Ci N/D N/D N/D N/D Xe-131m Ci N/D N/D N/D 2.70E-03 Xe-133m Ci N/D N/D 1.0SE-02 5.60E-03 Ar-41 Ci N/D N/D N/D 3.90E-04 TOTAL FOR PERIOD Ci 3.49E-02 6.81E-02 1.32E+OO 1.40E+OO

2. IODINES 1-131 1-133 1-135 Ci Ci Ci 1.SlE-06 1.60E-07 N/D 3.SOE-07 2.70E-08 N/D N/D N/D N/D N/D N/D N/D TOTAL FOR PERIOD Ci 1.67E-06 3.77E-07 NIA NIA

3. PARTICULATES Sr-89 Ci N/D N/D N/D N/D Sr-90 Ci N/D N/D N/D N/D Cs-134 Ci N/D N/D N/D N/D Cs-137 Ci 2.34E-08 2.93E-08 N/D N/D Ba-140 Ci N/D N/D N/D N/D La-140 Ci .N/D N/D N/D N/D Co-60 Ci 1.70E-08 4.68E-08 N/D N/D TOTAL FOR PERIOD Ci 4.04E-08 7.61E-08 NIA NIA I_

Attachment 1 TABLElB Page 4 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

PERIOD: 111/95 TO 12131/95 GASEOUS EFFLUENTS-MIXED MODE RELEASES CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT THIRD FOURTH THIRD FOURTH QUARTER QUARTER QUARTER QUARTER

1. FISSION & ACTIVATION GASES Kr-85 Ci N/D N/D 4.54E+OO 2.02E+Ol Kr-85m Ci N/D N/D 7.00E-03 2.00E-03 Kr-87 Ci N/D N/D N/D N/D Kr-88 Ci N/D N/D N/D N/D Xe-133 Ci 7.43E-01 7.97E-02 1.39E+Ol 1.40E+Ol Xe-135m Ci N/D N/D N/D N/D Xe-135 Ci 4.60E-02 1.13E-02 6.lOE-01 1.19E-Ol Xe-138 Ci N/D NID N/D N/D Xe-131m Ci N/D NID 5.SOE-01 1.30E+OO Xe-133m Ci N/D N/D 2.03E-Ol 9.70E-02 Ar-41 Ci N/D N/D NID 1.36E-03 TOTAL FOR PERIOD Ci 7.89E-Ol 9.lOE-02 1.98E+Ol 3.57E+Ol

2. IODINES 1-131 1-133 1-135 Ci Ci Ci N/D N/D I

3.30E-08 3.90E-09 N/D N/D N/D N/D N/D N/D N/D N/D TOTAL FOR PERIOD Ci 3.30E-08 3.90E-09 NIA NIA

3. PARTICULATES Sr-89 Ci N/D N/D N/D N/D Sr-90 Ci J'i/D N/D N/D N/D Cs-134 Ci N/D N/D N/D N/D Cs-137 Ci 3.SOE-08 7.30E-08 NID N/D Ba-140 ti N/D N/D NID N/D La-140 Ci N/D N/D N/D N/D Co-58 Ci N/D N/D N/D N/D Co-60 Ci 1.74E-08 5.47E-08 N/D N/D TOTAL FOR PERIOD Ci 5.24E-08 1.28E-07 NIA NIA L

Attachment 1 TABLElC PageS of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

PERIOD:1/1195 TO 12131195 GASEOUS EFFLUENTS-GROUND LEVEL RELEASES CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT FIRST SECOND FIRST SECOND QUARTER QUARTER QUARTER QUARTER

1. FISSION & ACTNATION GASES Kr-85 Ci NID N/D N/D N/D Kr-85m Ci N/D 4.40E-05 N/D N/D Kr-87 Ci NID NID N/D N/D Kr-88 Ci N/D N/D N/D N/D Xe-133 Ci 1.3SE-03 1.93E-04 N/D N/D Xe-135 Ci 1.lSE-03 9.41E-01 N/D N/D Xe-135m Ci 7.88E-04 1.41E-03 NID N/D Xe-138 Ci N/D N/D N/D N/D Ar-41 Ci 2.37E-03 2.12E-03 NID NID Xe-133m Ci N/D N/D N/D NID Xe-131m Ci NID NID N/D N/D TOTAL FOR PERIOD Ci 5.68E-03 9.45E-01 NIA NIA

2. IODINES 1-131 Ci 1.39E-04 7.70E-05 1.SOE-06 N/D 1-133 Ci 2.44E-05 4.12E-OS N/D N/D 1-135 Ci N/D N/D N/D N/D TOTAL FOR PERIOD Ci 1.63Ei°4 1.lSE-04 1.SOE-06 NIA

3. PARTICULATES Sr-89 Ci N/D N/D NID N/D Sr-90 Ci NID N/D N/D N/D Cs-134 Ci N/D N/D N/D N/D Cs-137 Ci 2.62E--05 1.90E-OS N/D N/D Ba-140 Ci N/D N/D N/D N/D La-140 Ci N/D N/D N/D N/D Se-75 Ci 1.lOE-06 NID N/D N/D Co-58 Ci 1.0lE-05 8.20E--06 N/D N/D Co-60 Ci 1.60E-OS 1.12E-OS N/D N/D TOTAL FOR PERIOD Ci S.34E-05 3.84E-OS NIA NIA

Attachment 1 TABLElC Page6 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

PERIOD:1/1/95 TO 12/31/95 GASEOUS EFFLUENTS-GROUND LEVEL RELEASES CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT THIRD FOURTH THIRD FOURTH QUARTER QUARTER QUARTER QUARTER

1. FISSION & ACTIVATION GASES Kr-85 Ci 1.77E--02 N/D 8.13E--Ol N/D Kr-85m Ci 2.70E--05 1.0SE--03 NID N/D Kr-87 Ci 6.30E--05 9.60E--04 N/D N/D Kr-88 Ci N/D 1.00E--03 3.00E--07 N/D Xe-133 Ci 2.90E+OO 5.32E--Ol 1.60E+02 N/D Xe-135m Ci 7.20E--04 1.40E--03 N/D N/D Xe-135 Ci 1.99E--03 1.04E--02 7.SOE--02 N/D Xe-138 Ci N/D N/D N/D N/D Ar-41 Ci 1.27E--03 1.lSE--03 N/D N/D Xe-131m Ci 7.19E-03 N/D l.82E+OO N/D Xe-133m Ci 1.74E--03 N/D 1.87E--02 N/D TOTAL FOR PERIOD Ci 2.93E+OO 5.48E-01 1.63E+02 NIA e 2. IODINES 1-131 Ci 6.97E--04 3.9SE--04 8.40E--04 4.60E--05 1-132 Ci 7.SOE--07 N/D 1.00E--04 N/D 1-133 Ci 6.28E--05 6.19E--05 1.70E--05 N/D TOTAL FOR PERIOD Ci 7.61E--04 4.56E-04 9.57E-04 4.60E-05

3. PARTICULATES Sr-89 Ci N/D N/D N/D N/D Sr-90 Ci .N/D N/D N/D N/D Cs-134 Ci NID N/D N/D N/D Cs-137 Ci 1.73E-05 1.'28E--05 2.lOE--05 3.20E-09 Ba-140 Ci N/D N/D N/D N/D La-140 Ci NID N/D N/D N/D Se-75 Ci N/D N/D N/D N/D Co-58 Ci 1.04E--05 N/D 2.lOE-06 N/D Co-60 Ci 5.SOE-06 1.60E-06 N/D N/D TOTAL FOR PERIOD Ci 3.34E--05 1.44E--05 2.31E-05 3.20E-09 I_

Attachment 1 TABLE2A Page 7 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

SURRY POWER STATION UNITS 1&2 PERIOD: 1/1/95 TO 12/31/95 LIQUID EFFLUENTS-SmWATION OF ALL RELEASES UNIT FIRST SECOND %EST.ERROR QUARTER QUARTER A. FISSION AND ACTIVATION PRODUCTS.

1. TOTAL RELEASE (NOT INCLUDING TRITIUM, GASES, ALPHA) Ci 1.33E--02 1.67E-02 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml 2.71E-ll 2.36E-11

3. PERCENT OF APPLICABLE LIMIT % 2.20E--04 2.21E-04 B.TRITIUM

1. TOTAL RELEASE Ci 1.71E+02 2.28E+02 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml 3.47E--07 3.22E--07

3. PERCENT OF APPLICABLE LIMIT % 3.47E--03 3.22E--03 C. DISSOLVED AND ENTRAINED GASES

1. TOTAL RELEASE Ci 6.40E--08 NID 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml l.30E-16 NIA

3. PERCENT OF APPLICABLE LIMIT % 6.SOE-11 NIA

D. GROSS ALPHA RADIOACTIVITY

1. TOTAL RELEASE Ci 'NID NID 2.00E+Ol E. VOLUME OF WASTE RELEASED (PRIOR TO DILUTION) LITERS l.69E+08 1.68E+08 3.00E+OO F. VOLUME OF DILUTION WATER USED DURING PERIOD LITERS 4.92E+ll 7.07E+ll 3.00E+OO

Attachment 1 TABLE2A Page 8 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

SURRY POWER STATION UNITS 1&2 PERIOD: 1/1/95 TO 12/31/95 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES UNIT THIRD FOURTH % EST. ERROR QUARTER QUARTER A. FISSION AND ACTNATION PRODUCTS

1. TOTAL RELEASE (NOT INCLUDING TRITIUM,GASES,ALPHA) Ci 1.97E-02 5.92E-03 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml 2.96E-ll 1.0SE-11

3. PERCENT OF APPLICABLE LIMIT % 8.34E-05 6.30E-05 B.TRITIUM

1. TOTAL RELEASE Ci 2.52E+02 1.80E+02 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml 3.78E-07 3.21E-07

3. PERCENT OF APPLICABLE LIMIT % 3.78E-03 3.21E-03 C. DISSOLVED AND ENTRAINED GASES

1. TOTAL RELEASE Ci 8.21E-05 N/D 2.00E+Ol

2. AVE DIL. CONC. DURING PERIOD µCi/ml 1.23E-13 NIA

3. PERCENT OF APPLICABLE LIMIT % 6.17E-08 NIA

D. GROSS ALPHA RADIOACTNITY

1. TOTAL RELEASE Ci N/D N/D 2.00E+Ol E. VOLUME OF WASTE RELEASED (PRIOR TO DILUTION) LITERS 1.06E+08 9.86E+07 3.00E+OO F. VOLUME OF DILUTION WATER USED DURING PERIOD LITERS 6.66E+ll 5.62E+ll 3.00E+OO

[_

Attachment 1 Page9 of 12

TABLE2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD: 111195 TO 12/31/95 LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT FIRST SECOND FIRST SECOND QUARTER QUARTER QUARTER QUARTER Sr-89 Ci N/D N/D N/D N/D Sr-90 Ci N/D N/D N/D N/D Cs-134 Ci 1.lOE-OS 1.19E-04 N/D N/D Cs-137 Ci 9.60E-03 1.49E-02 S.90E-07 6.60E-06 1-131 Ci N/D N/D N/D N/D Co-58 Ci N/D N/D 1.12E-OS N/D Co-60 Ci 3.70E-03 1.61E-03 1.28E-OS 9.40E-06 Fe-59 Ci N/D N/D N/D N/D Zn-65 Ci N/D N/D N/D N/D Mn-54 Ci N/D N/D N/D N/D Cr-Sl Ci N/D N/D N/D N/D Zr-95 Ci N/D N/D N/D N/D Nb-95 Ci N/D N/D N/D N/D Mo..:99 Ci N/D N/D N/D N/D Tc-99m Ci N/D N/D N/D N/D

,.., Ba-140 La-140 Ci Ci N/D N/D N/D N/D N/D N/D N/D N/D Ce-141 Ci N/D N/D N/D N/D Sb-125 Ci N/D N/D 6.90E-07 N/D TOTAL FOR PERIOD Ci 1.33E-02 1.67E-02 2.S2E-OS 1.60E-OS Xe-133 Ci N/D N/D N/D N/D Xe-135 Ci N/D N/D 6.40E-08 N/D TOTAL FOR PERIOD NIA NIA 6.40E-08 NIA

Attachment 1 TABLE2B Page 10 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD: 1/1/95 TO 12131/95 LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE SURRY POWER STATION UNITS 1&2 UNIT TIIlRD FOURTH THIRD FOURTH QUARTER QUARTER QUARTER QUARTER Sr-89 Ci N/D N/D N/D N/D Sr-90 Ci N/D N/D N/D N/D Cs-134 Ci 3.00E-05 1.20E-05 1.90E-06 2.90E-06 Cs-137 Ci 4.79E-03 2.90E-03 5.00E-06 1.26E-05 I-131 Ci N/D N/D 8.94E-OS 1.44E-OS Co-58 Ci 2.70E-OS 1.70E-04 N/D 1.UE-04 Co-60 Ci 9.74E-04 4.39E-04 9.90E-OS 1.24E-03 Fe-59 Ci N/D N/D N/D N/D Zn-65 Ci N/D N/D N/D N/D Mn-54 Ci N/D N/D N/D 2.30E-06 Cr-51 Ci N/D N/D N/D 4.30E-07 Zr-95 Ci N/D N/D N/D 1.60E-07 Nb-95 Ci N/D N/D N/D 3.90E-06 Mo-99 Ci N/D N/D N/D N/D Tc-99m Ci N/D N/D N/D N/D Ba-140 Ci N/D N/D N/D N/D La-140 Ci N/D N/D N/D N/D Ce-141 Ci N/D N/D N/D N/D Co-57 Ci N~ N/D N/D 2.30E-08 Sb-122 Ci N/D N/D NID N/D Sb-125 Ci N/D N/D N/D 2.90E-07 I-133 Ci N/D N/D 7.SOE-07 N/D Na-24 Ci N/D N/D 1.37E-02 1.0lE-03 TOTAL FOR PERIOD Ci 5.82E-03 3.52E-03 1.39E-02 2.40E-03 1

Xe-133 Ci N/D N~ 8.21E-05 N/D Xe-135 Ci N/D N/D N/D N/D TOTAL FOR PERIOD Ci NIA NIA 8.21E-05 NIA

TABLE3 Attachment 1 Page 11 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS PERIOD: 1/1/9S - 12/31/9S SURRY POWER STATION A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

1. Type of waste 12month Est. Total Period Error, %

a. Spent resins, filter sludges, evaporator m3 4.69E+02* 1.00E+Ol bottoms, etc. Ci 4.40E+Ol 3.00E+Ol

b. Dry compressible waste, contaminated m3 2.39E+02** 1.00E+Ol equip., etc. Ci 7.47E+OO 3.00E+Ol

c. Irradiated components, control m3 O.OOE+OO 1.00E+Ol rods, etc. Ci O.OOE+OO 3.00E+Ol

d. Other (Waste Oil) m3 1.00E+Ol *** 1.00E+Ol Ci 5.lSE-02 3.00E+Ol

2. Estimate of major nuclide composition (by type of waste)

a. Co-58 % 5.63E+Ol Cs-137 % 1.13E+Ol Cs-134 % 1.02E+Ol Co-60 % 1.02E+Ol Ni-63 % 7.59E+OO C-14 % 1.69E+OO

b. Co-58 % 3.17E+Ol Co-60 % 2.68E+Ol Ni-63 % 1.34E+Ol Fe-55 % 1.32E+Ol Cs-137 % 9.46E+OO H-3 % 2.12E+OO Cs-134 % 1.07E+OO Mn-54 % 1.02E+OO

TABLEJ Attachment 1 Page 12 of 12 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SURRY POWER STATION SOLID WASTE AND IRRADIATED FUEL SHIPMENTS PERIOD: 1/1/9S - 12/31/9S CONT'D A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

2. Estimate of major nuclide composition (by type of waste) c.

d. Pu-241 % 3.69E+Ol Sr-90 % 2.S9E+Ol Ni-63 % 2.SSE+Ol Fe-55 % 4.SSE+OO Cm-243/244 % 2.S3E+OO Pu-238 % 2.27E+OO Am-241 % 1.29E+OO

3. Solid Waste Disposition Number of Shipments :M:odeofTransportation Destination 28 Truck Barnwell, SC 20 Truck Oak Ridge, TN B. IRRADIATED FUEL SHIPMENT (Disposition)

Number of Shipments :M:ode of Transportation Destination 0

NOTE 1: 3.11E+02 m3 of the resin/sludge/filters was shipped to a licensed waste processor for processing and/or volume reduction. Therefore, this volume is not representative of the actual volume buried.

The total volume buried for this reporting period is 1.61E+02 m3.

- NOTE 2: 2.36E+02 m3 of the DAW was shipped to licensed waste processors for volume reduction.

Therefore, this volume is not representative of the actual volume buried. The total volume buried for this reporting period is 1.87E+Ol m3.

- - NOTE 3: Waste oil was shipped to a licensed waste processor for incineration. Therefore, 1.00E+Ol m3 is not representative of the actual volume buried. The total volume buried for this reporting period is 2.27E-02 m3 .

I Attachment 2

. Page 1 of 1

ANNUAL AND QUARTERLY DOSES An assessment of radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents released from the site for each calendar quarter for the calendar year of this report, along with an annual total of each effluent pathway is made pursuant to the* ODCM, section 6; 7 .2 requirement.

LIQUID GASEOUS Total Body Thyroid GI-LLI Gamma Beta Thyroid (mrem) (mrem) (mrem) (mrad) (mrad) (mrem) 1st Quarter 1.30E-04 4.30E-05 1.79E-04 7.12E-05 7.25E-05 3.00E-03 2nd Quarter 1.16E-04 3.85E-05 7.34E-05 3.50E-03 4.46E-03 1.59E-03 3rd Quarter 7.51E-05 5.79E-05 7.24E-05 1.09E-01 3.31E-01 3.07E-02 4th Quarter 6.40E-05 4.18E-05 9.23E-05 6.50E-04 2.93E-03 8.89E-03 Annual 3.85E-04 1.81E-04 4.17E-04 1.13E-Ol 3.38E-01 4.41E-02

' 1 .

Attachment 3 Page 1 of 1

REVISIONS TO OFFSITE DOSE CALCULATION MANUAL CODCM}

As required by Technical Specification 6.8.B, revisions to the ODCM, effective for the time period covered by this report, are summarized in this attachment.

There were 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 the ODCM corresponding to Surry' s procedure changes. Revision 7, effective October 31, 1995, incorporates the two procedure changes and additional changes described in the revision summary.

The changes which relate to Surry Power Station are the following:

1. 6-PS 1 - a replacement to one of the oyster sampling locations included as part of the Radiological Environmental Monitoring Program, which was implemented April 4, 1995.

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.

3. Revision 7 - update of Rated ThermatPower definition to reflect Surry's Core Uprate to 2546 MWt, which was implemented October 31, 1995.

FOR INFORl~ATIOi~ ONLY

Station Administrative VIRGINIA POWER

Title:

Offsite Dose Calculation Manual Procedure

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.

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 miles.

Incorporate PN&S 1. Change equipment mark numbers to new fonnat.

Changed Attachments 14 and 16 to include mark numbers for VG-RM-104 and ventilation flow rate monitors.

E-PAR PS1 Change Att #2, replace Wallace Point oyster sample with Kingsmill oyster sample Surry Power Station North Anna Power Station Approved by: Approved by:

J. A. Price 04-03-95 NIA NIA SNSOC Chairman Date SNSOC Chairman Date Approved by: Approved by:

J. H. McCarthy 04-04-95 NIA NIA Station Manager Date Station Manager Date Approved by: NIA NIA Vice President-Nuclear Operations Date CONTROLLED DOCUMENT: Use within 24 hrs. after 3/8/96 @ 0913 or verify rev/change#.

,* , , * , \ t *1 : 0 ~

VIRGINIA ~ * ,_ . . .

1

.: .,. *~ J./ r., :. . ... ' .l .... ,_,* '. ::e:* i l*J .' .~ i' :..*i VPAP-2103 POWER REVISION 6-PSl PAGE2 OF 156 Intentionally Blank CONTROLLED DOCUMENT: Use within 24 hrs. after 3/8/96 @ 0913 or verify rev/change#.

_j

Station Administrative VIRGINIA POWER

Title:

Offsite Dose Calculation Manual Procedure

© 1995 by Virginia Power. All Rights Reserved Lead Department: Radiological Protection Procedure Number Revision Number Effective Date VPAP-2103 7 10-31-95 Revision Summary 0 Incorporates PS 1: Changes ATIACHMENT 22, Surry Environmental Sampling Locations to replace Walnut Point oyster sampling location with Kingsmill oyster sampling location on page 128

Incorporates PNl: Changes ATIACHMENT 23, North Anna Environmental Sampling Locations to correct Emergency TLD locations on pages 129 and 130, adds reference 3.1.22, Deviation Report N94-l 137, Improper Placement of Emergency TLDs to

page9

Corrects units of I 131 concentration to µCi/cc at 4.5, Dose Equivalent I-131 on page 10

Reflects Surry Core Uprate to 2546 MWt at 4.13, Rated Thermal Power on page 12 Surry Power Station North Anna Power Station Approved by: Approved by:

~o~ q[2 t/'i~ DA~\~ \ O*S~'S'

~~mby:

SNOChairman Station Manager Date

---,:;--z r-"'I~

Date SNSOC Chairman

7§: ii_)

Approved by:

Station Manager 1ol'otl5 Date Date J/;ti-

Approved by:

Vice President-Nuclear Operations Date

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-*-~j

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Section TABLE OF CONTENTS 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 6.2 Liquid Radioactive Waste Effluents 15 6.2.1 Liquid Effluent Concentration Limitations 15

6.2.2 Liquid Monitoring Instrumentation 6.2.3 Liquid Effluent Dose Limit 6.2.4 Liquid Radwaste Treatment 6.2.5 Liquid Sampling 16 20 23 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 6.3.5 Gaseous Radwaste Treatment 36 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

6.4.5 Miscellaneous Gaseous Release Permit 40

VIRGINIA VPAP-2103 POWER REVISION 7 PAGE40F 156 TABLE 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 ATTACHMENTS 1 Surry Radioactive Liquid Effluent Monitoring Instrumentation 51 53 2 North Anna 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 1 and 2 63 7 North Anna Liquid Pathway Dose Commitment Factors for Adults 69 8 Surry Radioactive Liquid Waste Sampling and Analysis Program 71 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 13 Gaseous Effluent Dose Factors for North Anna 89 93

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Section TABLE OF CONTENTS (continued)

ATTACHMENTS (continued)

Page

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 19 Critical Organ Dose Factors for North Anna 113 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 Locations 129

24 Detection Capabilities for Surry Environmental Sample Analysis 25 Detection Capabilities for North Anna Environmental Sample Analysis 26 Reporting Levels for Radioactivity Concentrations in Environmental Samples at Surry 133 135 137 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

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1.0 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 µCi/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

10 CFR 50, Appendix I

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

- I 131 , I 133 , and H 3 , 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 I, and

Exposure to a real individual will not exceed 40 CFR 190 dose limits 2.0 SCOPE This procedure applies to the Radioactive Effluent and Environmental Monitoring Programs at Surry and North Anna Stations.

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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 Releases 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 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 Assessment 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 *

3.1.14 NUREG-0543, February 1980, Methods for Demonstrating LWR.Compliance With 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 Radiological Effluent Technical Specifications (RETS) in the Administrative Controls

Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program

. 3.1.18 UFSAR (Surry and North Anna)

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3.1.19 Nuclear Reactor Environmental Radiation Monitoring Quality Control Manual, IWL-0032-361 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 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 Adjustment, as necessary, of the channel output 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 Chann~I 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 sensor as practicable, to

verify Operability, including alarm and/or trip functipns .

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PAGE 10 OF 156 4.3 Channel Functional Test (continued) 4.3.2 Bistable Channel Injection of a simulated signal into 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 1-131 That concentration of 1131 (µCi/cc) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131 , I 132, I 133 , I 134, and I 135 actually present. Thyroid dose conversion factors for 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 NOTE: Frequencies are allowed a maximum extension of 25 percent.

NOTATION FREQUENCY D - Daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W -Weekly At least once per 7 days M- Monthly At least once per 31 days Q - Quarterly At least once per 92 days SA - Semi-annually At least once per 184 days R - Refueling At least once per 18 months SIU - 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

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4.7 Gaseous Radwaste Treatment System A system that reduces radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing delay or holdup to reduce total radioactivity prior to-release tothe 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).

4.8 General Nomenclature X = Chi: concentration at a point at a given instant (curies per cubic meter)

D = Deposition: quantity of deposited radioactive material per unit area (curies per square meter)

Q = Source strength (instantaneous; grams, curies)

= Emission rate (continuous; grams per second, curies per second)

= Emission rate (continuous line source; grams per second per meter) 4.9 Lower Limit of Detection (LLD)

The smallest concentration of radioactive material in a sample that will yield a net count (above system background) that can be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

'II 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 materials.

4.11 Operable - Operability A system, subsystem, train, component, or device is operable or has operability when it is 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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PAGE 12 OF 156 4.12 Purge* Purging Controlled discharge of air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, so that replacement air or gas is required to purify tlie confinement.

4.13 Rated Thermal Power Total reactor core heat transfer rate to reactor coolant.

Surry - 2546 Megawatts Thermal (MWt)

North Anna - 2893 MWt 4.14 Site Boundary The line beyond which Virginia Power does not own, lease, or otherwise control the land.

4.15 Source Check A qualitative assessment of channel response when a channel sensor is exposed to radiation.

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 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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4.19 Ventilation Exhaust Treatment System A system that reduces gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases throu~h charcoal adsorbers and High 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 effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not Ventilation Exhaust Treatment System components.

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 radioactive effluents and the environment.

5.1.2 Surveying, sampling, and analyzing plant effluents and environmental monitoring, and documenting these activities.

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 authorization to release effluents.

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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 Surveys, 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 radiation or parameter monitored.

6.1.3 A sufficient number of survey points shall be used or samples taken fo 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

documented, and sufficiently detailed that the meaning and intent of the records are clear.

6.1.6 Surveys, analyses, and monitoring records shall be reviewed for trends, completeness, and accuracy.

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:

1. For radionuclides (other than dissolved or entrained noble gases), liquid effluent concentrations released to unrestricted areas shall not exceed ten times the effluent concentration 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 µCi/ml.

b. If the concentration of liquid effluent exceeds the limits in 6.2.1.a., promptly reduce

concentrations to within limits.

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PAGE 16 OF 156 6.2.1 Liquid Effluent Concentration Limitations (continued)

. c. Daily concentrations of radioactive materials in liquid waste released to unrestricted areas shall meet the following:

Volume of Waste Discharged+ Volume of Dilution Water> (l) 1

. . µCi/ml. -

Volume of Waste Discharged x L ACW.

1

. i I where:

µCi/mli. = the concentration of nuclide i in the liquid effluent discharge

. ACWi =ten times the effluent concentration value in unrestricted areas of nuclide i, expressed as µCi/ml from 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than noble gases, and 2E-4 µCi/ml for dissolved or entrained noble gases 6.2.2 Liquid Monitoring Instrumentation

a. Radioactive Liquid Effluent Monitoring Instrumentation Radioactive liquid effluent monitoring instrumentation channels shown on

.Attachments 1 and 2, Radioactive Liquid Effluent Monitoring Instrumentation, 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

Declare the channel inoperable

Change the setpoint to an acceptable, conservative value

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b. Radioactive Liquid Effluent 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 atthe-frequencies shown in Attachments 3 and 4, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

1. If the number of operable channels is less than the minimum required by the tables in Attachment 1 or 2, perform the action shown in those tables.

2. Attempt to return 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 Liquid Radwaste Effluent Line 1-LW-RM-111 NIA Service Water System Effluent Line 1-SW-RM-108 1-SW-RM-107 A, B,C,D Condenser Circulating Water Line l-SW-RM-130 1-SW-RM.:120 2-SW-RM-230 2-SW-RM-220 Radwaste Facility Effluent Line NIA 1-RRM-RITS-131

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PAGE 18 OF 156 6.2.2 Liquid Monitoring Instrumentation (continued)

d. Setpoint Calculation NOTE:. This methodology does not preclude use of more conservative setpofuts.

1. Maximum setpoint values shall be calculated by:

s- (2) where:

S = the setpoint, in µCi/ml, of the radioactivity monitor measuring the 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 µCi/ml FE = maximum design pathway effluent flow rate F0 = dilution water flow rate calculated as:

(Surry) D =FE+ (200,000 gpm x number of circ. pumps in service)

(N. Anna) D =FE+ (218,000 gpni x number of circ. 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

. circulating water discharge tunnel beyond the last point of possible radioactive material addition. No dilution is_assumed for this pathway. Therefore, Equation (2) becomes:

S=C (3)

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 rad waste 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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4.

For the liquid radwaste effluent line, Equation (2) becomes: .

s= (4) where:

KLw = 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:

s = (5) where:

Ksw = The fraction of the effluent concentration limit, used to implement 10 CFR 20 for the Station, attributable to the service water effluent line pathway

6. For the Radwaste Facility effluent line, Equation (2) becomes:

s = (6) where:

KRw = The fraction of the effluent concentration limit, used to implement 10 CFR 20 attributable to the Radwaste Facility effluent line pathway

7. The sum KLw + Ksw + KRw shall not be greater than 1.0.

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PAGE 20 OF 156 6.2.3 Liquid Effluent Dose Limit

a. Requirement At least once pet 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:

1. During any calendar quarter:

Less than or equal to 1.5 mrem to the total body

Less than or equal to 5 mrem to the critical organ

2. During any calendar year:

Less than or equal to 3 mrem to the total body

Less than or equal to 10 mrem to the critical organ

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, special report in accordance with VPAP-2802, Notifications and Reports, that identifies causes for exceeding limits and defines corrective actions taken to reduce releases of radioactive materials in liquid effluents to ensure that subsequent a

releases will be in compliance with the above limits.

c. Surry Dose Contribution Calculations NOTE: Thyroid and GI-LLI organ doses must be calculated to determine which is the critical organ for the period being considered.

Dose contributions shall be calculated for all radionuclides identified in liquid effluents released to unrestricted areas based on the equation:

D = tFM~C.A. L,J I I (7) i where:

Subscripts = i, refers to individual radionuclide

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PAGE 21 OF 156

D t

=

the cumulative dose commitment to the total body or critical organ from the liquid effluents for the period t, in mrem the period for which Ci and F are averaged for all liquid releases, in hours M = the mixing ratio (reciprocal of the dilution factor) at the point of exposure, dimensionless, 0.2 from Appendix 1 lA, Surry UFSAR F = the near field average dilution factor for q 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 .

Ci = the average concentration of radionuclide, i, in undiluted liquid effluent during the period t, from all liquid releases, in µCi/ml Ai = 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 in mrem-ml per hr-µCi. Values for Ai are given in Attachment 5, Liquid Ingestion Pathway Dose Factors For Surry Power Station.

A.1 = 1.14 E+05 (21BF.1 + 5BI.) 1 DF.1 (8)

where:

1.14 E+05 = 1 E+o6 pCi/µCi 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 Bli = 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. 1 OF1 = 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

VIRGINIA VPAP-2103 POWER REV1SION7 PAGE 22 OF 156 6.2.3 Liquid Effluent Dose Limit (continued)

d. North Anna Dose Contribution Calculations NOTE: . Attachment 6,.North.Anna Liquid Ingestion Pathway,Dose Factor*Calculation provides the derivation for Equation (9).

Dose contribution shall be calculated for all radionuclides identified in liquid effluents released to unrestricted areas. based on:

(9)

D = ~Q.xB.

£- 1 1 1

Where:

Subscripts = i, refers to individual radionuclide 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 (mrern/Ci) for adults. Values for Bi are provided

in Attachment 7, North Anna Liquid Ingestion Pathway Dose Commitment Factors for Adults Qj = Total released activity for the considered period and the ith nuclide Q.1 = txC.xWasteFlow 1

(10)

Where:

t = the period for which Ci and F are averaged for all liquid releases, in hours q = the average concentration of radionuclide, i, in undiluted liquid effluent during the period, t, from any liquid releases, in µCi/ml

e. Quarterly Composite Analyses 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 composite analyses. However, for reporting purposes, calculated dose contribution shall be based on the actual composite analyses .

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6.2.4 Liquid Radwaste Treatment

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 above limits prepare and submit to the NRC, within 30 days, a special report in accordance with VPAP-2802, Notifications and Reports, that includes the following:

1. An explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-system, and the reason for the

inoperability.

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 OTB, 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 1, the ratio of the estimated volume of liquid effluent releases in the present 31-day period to the volume released in the previous 31-day period.

3. Estimate F 1, 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 (µCi/ml).

4. Determine PDTB, the projected total body dose in a 31-~ay period.

(11)

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PAGE 24 OF 156 6.2.4 Liquid Radwaste Treatment (continued)

d. Projected Critical Organ Dose Calculation

Bistorical data pertaining to the volumes and radioactivity of liquid effluents released in connection with specific Station functions, such as maintenance or refueling outages, shall be used in projections as appropriate.

1. Determine D0 , the critical organ dose from liquid effluents in the previous 31-day period, per Equation (7) or Equation (9) (Surry and North Anna, respectively).

2. Estimate R 1 as in 6.2.4.c.2.

3. Estimate F1 as in 6.2.4.c.3.

4. Determine PD0 = projected critical organ dose in a 31-day period.

(12)

. 6.2.5 Liquid Sampling .

Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis requirements in Attachments 8 and 9, Radioactive Liquid Waste Sa111pling and Analysis Progra~*(Surry and North *Anna;respectively).

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,::;; 500 mrern/year to the total body and ,::;; 3000 mrern/year to the skin.

2. The dose rate limit for I 131 , I 133 , for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days shall be,::;; 1500 mrern/year

to the critical organ.

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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2. Dose rates due to noble gases in gaseous effluents shall be determined, continuously, to be within 6.3.1.a. limits.

3. Dose rates due to I 131 , I 133 , tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents shall be determined to be within the above limits by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified on Attachments 10 and 11, Radioactive Gaseous Waste Sampling and Analysis Program.

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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:

4'1

[KivvQivv + KipvQipvl ~ 500mrem/yr to the total body (13)

OR

~

[(L. + l.lM. ) Q1*vv + (L. + l.lM. ) Q1*pvl :::; 3000mrem/yr to the skin lVV lVV 1pv 1pv .

i . (14) where:

Subscripts = vv, refers to vent releases from the building ventilation vent, including Radwaste Facility Ventilation Vent; pv, refers to the vent releases from the process vent; i, refers to individual radionuclide I<.ivv* Kipv. = 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

I and North Anna, respectively)

Livv* 4pv = 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 Mivv* Mipv = 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

'2ivv'2ipv = 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

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 27 OF 156

2. The dose rate limit for I 131 , I 133 , 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:

L[PivvQivv + PipvQipv]::;; 1500mrem/yr to the critical organ (15) i where:

= The critical orEan dose factor for ventilation vents or process vent for I 131 , I 33 , H 3, and all radionuclides in particulate form 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

= The release rate for ventilation vents or process vent of I 131 ,

1133 , H 3, 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

and shall be included in the determination of '21vv 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 that 6.3.1.a. noble gas limits are not exceeded. Alarm/trip setpoints of these channels 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 Instrumentation Surveillance Requirements.

VIRGINIA VPAP-2103 POWER REVISION7 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.l, 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 Radioactive gaseous effluent monitors for which alarm/trip setpoints shall be determined are:

Release Point Instrument Number NorthAnna Surry Process Vent 1-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 NIA l-VG-RM-179-1 Ventilation Vent B 1-VG-RM-113 NIA 1-VG-RM-180-1 Ventilation Vent No. 1 NIA 1-VG-RM-104 Ventilation Vent No. 2 NIA 1-VG-RM-110 1-VG-RM-131-1 Radwaste Facility Vent NIA RRM-101

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d. Setpoint Calculations

1. Setpoint calculations for each monitor listed in 6.3.2.c. shall maintain this relationship:

D~D pv +D cae +D vv (16) where:

D = Step 6.3.1.a. dose limits that implement 10 CFR 20 for the Station, mrem/yr Dpv = The noble gas site boundary dose rate from process vent gaseous effluent releases, mrem/yr Dcae = The noble gas site boundary dose rate from condenser air ejector gaseous effluent releases, mrem/yr Dvv = The noble gas site boundary dose rate from:

Surry: Summation 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

effluent releases, mrem/yr

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2. Setpoint values shall be determined by:

R m x 2.12 E-03 cm= (17) where:

m = The release pathway, process vent (pv), ventilation vent (vv) condenser air ejector (cae), or Rad waste Facility (rv)

Cm = The effluent concentration limit implementing 6.3.1.a. for the Station, µCi/ml Rm = The release rate limit for pathway m determined from methodology in 6.3.1.c., using Xe 133 as nuclide to be released,

µCi/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 radioiodines, radioactive materials in particulate form, and radionuclides other 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 gamma radiation and ~ 10 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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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 report in-accordance withVPAP-2802, Notifications and Reports, that identifies the 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 subsequent releases will be in compliance with the limits in 6.3.3.a.

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 Qvv*

The air dose to areas at or beyond the site boundary due to noble gases shall be determined by the following:

For gamma radiation:

Dg = 3.17E-08i [MivvQivv+MipvQipvJ 1 (18)

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PAGE 32 OF 156 For beta radiation:

Db = 3.17E-08L [NivvQivv +NipvQipvJ (19) 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 i, refers to individual radionuclide Dg = the air dose for gamma radiation, in mrad Db = the air dose for beta radiation, in mrad Mivv, 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 arid 13 Nivv, Nipv = the air dose factor for ventilation vents or process vent release due to beta emissions for eachidentified noble gas

radionuclide i, in mrad/yr per Curie/sec. Factors are listed in Attachments 12 and 13

= 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)

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6.3.4 1-131, 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,! 131 , I 133 , tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents released from the site to unrestricted areas from each reactor unit shall be:

During any calendar quarter: ~ 7 .5 mrem to the critical organ

During any calendar year: ~ 15 mrem to the critical organ

2. Cumulative dose contributions to a member of the public from I 131 , I 133 ,

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 ff the calculated dose from the release of I 131 , I 133 , tritium, and radionuclides in

particulate form, with half-lives greater than 8 days, in gaseous 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 Qivv. 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 .

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1. The dose to the maximum exposed member of the public, attributable to gaseous effluents at and beyond the site boundary that contain I 131 , I 133 , tritium, arid particulate-form radionuclides with half-lives greater than 8 days, shall be determinedby: *

Dr = 3.l?E-OS L[(RMivvQivv + RMipvQipv) + (RlivvQivv + RiipvQipv)] (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 frorri the process vent Dr = the dose to the critical organ of the maximum exposed member of the public in mrem RMivv, RMipv= the cow-milk pathway dose factor for ventilation vents or process vent release due to I 131 , I 133 , 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 Rlivv, Rlipv = the inhalation pathway dose factor for ventilation vents or process vent release due to I 131 , I 133 , 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

= the release for ventilation vents or process vent of I 131 , I 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

VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 35 OF 156 6.3.4 1-131, 1-133, H-3, and Radionuclides In Particulate Form Effluent Dose Limit (continued)

d. North Anna Dose Calculations Gaseous releases, not through the process vent; are considered ground level and shall be included in the determination of Q;vv* 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 I 131 , I 133 ,

tritium, and particulate-form radionuclides with half-lives greater than 8 days, shall be determined by:

)

r 3.17E-084 [RMivvQivv + RMipvQipv: (21) 1

Where:

Subscripts Dr

= vv, refers to vent releases from the building ventilation vents; pv, refers to the vent releases from the process vent

= the dose to the critical organ of the maximum exposed member of the public, in mrem RMivv, RMipv =the cow-milk dose factor for ventilation vents or process vent release due to I 131 , I 133 , tritium, and from all particulate-form radionuclides with half-lives greater than 8 days, in rnrem/yr per Curie/sec. Factors are listed in Attachment 19, Critical Organ Dose Factors for North Anna Qiw Q ipv = the release for ventilation vents or process vent of I 131 , I 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

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PAGE 36 OF 156 6.3.5 Gaseous Radwaste Treatment Historical data pertaining to the volumes and radioactive concentrations of gaseous effluents released in connection with specific Station functions, such as containment purges, shall be used to calculate projected doses, as appropriate.

a. Requirement

1. The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be used to reduce radioactive material in 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 exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation, averaged 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)

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:

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.

3. Summary description of actions taken to prevent recurrence.

VIRGINIA VPAP-2103 POWER REVISION 7 PAGE 37 OF 156 6.3.5 Gaseous Radwaste Treatment (continued)

c. Projected Gamma Dose

1. Determine Dg, the 31-day gamma air dose for the previous 3'1-day period; per Equation (18).

2. Estimate Rg, 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 Fg, the 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 (µCi/ml).

4. Determine PDg, the projected 31-day gamma air dose.

PD g = D g (R g xF g ) (22)

d. Projected Beta Dose

1. Determine Db, the 31-day beta air dose in the previous 31 days,

per Equation (19).

2. Estimate Rg and Fg as in 6.3.5.c.2. and 6.3.5.c.3.

3. Determine PDb, the projected 31-day beta air dose.

(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=

Dmax*

2. Estimate Fi, the ratio of the estimated activity from I 131 , I 133 , radioactive materials in particulate form with half-lives greater than 8 days, and tritium in the current 31-day period to the activity of 1131 , 1133 , radioactive materials in particulate form with half-lives greater than 8 days, and tritium in the previous 31-day period (µCi/ml) .

VIRGINIA* VPAP-2103 POWER REVISI0N7

. PAGE 38 OF 156

3. Detennine PDmax, the projected 31-day maximum exposed member of the public dose.

PD max =-D max (R g xF.-} 1 (24) 6.4 Radioactive Liquid and Gaseous Release Permits RP shalfmaintain 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 RP 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 shall have 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 (LWTT)

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 the presence of contamination from primary-to-secondary leakage _

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 39 OF 156

6.4.1 Liquid Waste Batch Release Permits (continued)

b. North Anna Batch Releases NOTE: H the clarifier is in seivice, releases from tanks processed through the clarifier* are considered continuous releases.

A Batch Release Pennit is required for a release from any tanks/sumps which contain (or potentially contain) radioactive liquid. Tanks/sumps include:

BRTT

LLWDT

HLWDT

Turbine Building Sumps when secondary coolant activity exceeds 1.0 E-5 µCi/ml

CDT*

6.4.2 Continuous Release Permit Operations shall obtain RP authorization before initiating continuous releases of

radioactive liquids ..

a. Surry Continuous Releases A Continuous release pennit is required at Surry for:

Steam generator blowdown

Component Cooling Water (CCW) heat exchanger to service water leakage, if applicable

Turbine Building sumps and/or subsurface drains if source activity concentrations are sufficiently low to allow continuous release

b. North Anna Continuous Releases A Continuous Release Pennit is required at North Anna for:

Clarifier, unless being bypassed

Steam generator blowdown when clarifier is bypassed

Containment mat sumps and seivice 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.

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 40 OF 156 6.4.4 Reactor Containment Release Permits Operations shall-obtain authorization from RP before initiating containment purges or containment hogging. Reactor Containment Release Permits shall be valid from start of purge/hog until:

Routine termination

Terminated for cause by RP

Receipt of Radiation Monitoring System (RMS) Containment Gas Monitor high alarm 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 Proc~ss 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

a. Operations shall notify RP of pending releases and request RP to initiate the 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 Primary 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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PAGE 41 OF 156

6.4.6 Radioactive Liquid and Gaseous Release Controls (continued)

e. RP shall notify the Shift Supervisor if it 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 correct 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.

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 shall include:

Date and time release was stopped

Tank/sump ending level

Release flow rate just prior to termination

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 exceed 25 mrem to the total body or the critical organ (except the thyroid, which shall not exceed 75 mrem).

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 42 OF 156 6.5.2 Action

a. If the calculated 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 directradiation contribution from the units and from outside storage tanks) whether limits in 6.5.1 have been exceeded.

b. If the limits in 6.5.1 have been exceeded, prepare and submit to the NRC within 30 days, a special report in accordance with VPAP-2802, Notifications and Reports, that defines the corrective action to be taken to reduce subsequent releases and to prevent recurrence, and includes a schedule for achieving conformance with the limits. Special reports, as defined in 10 CFR 20.2203(a)(4), shall include:

1. An analysis that estimates the radiation exposure (dose) to a real individual from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the releases covered by the report.

2. A description of the levels of radiation and conc.entrations of radioactive material involved, and the cause of the exposure levels or concentrations.

3. If the estimated dose exceeds the limits in 6.5.1, and if the 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

-1... .

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.

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PAGE 43 OF 156

6.6 Radiological Environmental Monitoring 6.6.1 Monitoring Program

a. Requirement

1. The Radiological Environmental Monitoring Program shall be conducted as 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. 1

b. Action

1. 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, in the Annual Radiological Environmental Operating Report required by Technical Specification (Surry Technical Specification 6.6.B.2 and North Anna Technical 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 calendar year limits of 6.2.3, 6.3.3, and 6.3.4

VIRGINIA VPAP-2103 POWER REV1SION7 PAGE 44 OF 156 When more than one of the radionuclides listed in Attachment 26 or 27 are detected in the sampling medium, the report shall be submitted if:

concentration (1). concentration.(2) >10 ( )

25 reporting level (1) + reporting level (2) + *** - *

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 ihe condition in the Annual Radiological Environmental Operating Report in accordance with VPAP-2802, Notifications and Reports.

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 replacement samples and add them to the radiological environmental

.monitoring program within 30 days. The specific locations from which samples 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 samples in 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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PAGE 45 OF 156

6.6.2 Land Use Census

a. Requirement A land use census shall be conducted and shall identify within a distance of 8 km (5 miles),thelocation in each of the 16 meteorological sectors of the following:

Nearest milk animal

Nearest residence

Nearest garden greater than 50 m2 (500 ft2) that produces broad leaf vegetation

1. The land use census shall be conducted during the growing season, at least once 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 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, identify the new locations in the next Annual Radioactive Effluent Release Report in accordance with VP AP-2802, Notifications and Reports.

2. If a land use census identifies locations that yield a calculated dose or dose 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 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]

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 46 OF 156 6.6.3 Interlaboratory Comparison Program

a. Requirement Radioactive materials (which contain nuclides produced at the Stations), supplied as part of arr Interlaboratory Comparison Program that has been approved by the NRC, shall be analyzed.

b. Action

1. Analyses shall be performed as part of the Environmental Protection Agency's Environmental Radioactivity Laboratory Intercomparison Studies (Cross Check) Program and include:

Program Cross-Check of Milk I 131 , Gamma, K, Sr89 and Sr90 Water Gross Beta, Gamma, 1131 , H 3 (Tritium), Sr89 and Sr90 (blind-any combinations of above radionuclides)

Air Filter Gross Beta, Gamma, Sr90

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, Notifications and Reports, the corrective actions taken to prevent recurrence .

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PAGE 47 OF 156

6.6.3 lnterlaboratory 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 I of each year.

A single submittal may be made for the Station. Radiological Environmental Operating Reports shall include:

a. Summaries, interpretations, and analysis of trends of results of radiological environmental surveillance activities for the report period, including:

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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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 format of Regulatory Guide 1.21, Appendix B.

2. An assessment of radiation doses to the maximum exposed members of the public due to the radioactive liquid and gaseous effluents released froin 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

Unplanned releases which require a Deviation Report and involve the 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 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, which may or may not require a Deviation Report

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PAGE 49 OF 156

6.7.2 Annual Radioactive Effluent 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. l, 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

a. Meteorological data collected during the previous year shall be in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.

b. Meteorological data shall be retained in a file on site and shall be made available to NRC upon request.

L

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 50 OF 156 6.7.4 Changes to the ODCM 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 th.e 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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PAGE 51 OF 156

7.0 7.1 RECORDS 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

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 maintenance, 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

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PAGE 53 OF 156

ATTACHMENT 1 (Page 1 of 1)

Surry Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum Action Operable Channels

1. GROSS RADIOACTNITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE (a) Radwaste Facility Liquid Effluent Line RM-RRM-131 1 1

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (a) Circulating Water Discharge Line Unit 1: 1-SW-RM-120 2 2 Unit 2: 2-SW-RM-220

3.

(b) Component Cooling Service Water Effluent Line 1-SW-RM-107A 1-SW-RM-107B 1-SW-RM-107C 1-SW-RM-107D FLOW RATE MEASUREMENT DEVICES 4 2 Rad waste 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.

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for principal gamma emitters, as defined in Attachment 8, ~urry Radioactive Liquid Waste Sampling 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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ATTACHMENT2 (Page 1 of 2)

North Anna Radioactive Liquid Effluent Monitoring Instrumentation Minimum Instrument Operable Action Channels

1. Liquid Radwaste Effluent (a) 1-RM-LW-111, Liquid Radwaste Effluent Monitor 1 1 (b) 1-LW-Fr-104, Liquid Radwaste Effluent Total Flow Measuring 1 2 Device (c) 1-LW-SOV-121, Clarifier Effluent Line Continuous Composite 1 1 Sampler and Sampler Flow Monitor (d) 1-LW-TK-20, Liquid Waste Effluent Sample Vessel 1 1 (e) 1-LW-1130, Liquid Waste Effluent Proportional Sample Valve 1 1 (t) 1-RM-SW-108, Service Water Effluent Monitor 1 1 (g) 1-RM-SW-130, Unit 1 Circulating Water System Effluent Line 1 4 Monitor (h) 2-RM-SW-230, Unit 2 Circulating Water System Effluent Line 1 4 Monitor

2. Tank Level Indicating Devices (Note 1)

(a) Refueling Water Storage Tanks Unit 1 1-QS-LT-lOOA 1 3 1-QS-LT-lOOB 1-QS-LT-lOOC 1-QS-LT-lOOD I Unit2 2-QS-LT-200A 1 3 2-QS-LT-200B 2-QS-LT-200C 2-QS-LT-200D (b) Casing Cooling Storage Tanks Unit 1 1-RS-LT-103A 1 3 1-RS-LT-103B Unit2 2-RS-LT-203A 1 3 2-RS-LT-203B (c) PG Water Storage Tanks (Note 2) 1-BR-LT-116A (1-PG-TK-lA) 1 3 1-BR-LT-116B (1-PG-TK-lB) 1 3 (d) Boron Recovery Test Tanks (Note 2)

1-BR-LT-112A (1-BR-TK-2A) 1-BR-LT-112B (l-BR-TK-2B) 1 1

3 3

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 56 OF 156 ATTACHMENT 2 (Page 2 of 2)

North Anna Radioactive Liquid Effluent 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity (beta and gamma) at an LLD of at least lxl0-7 µCi/g or an isotopic radioactivity at an LLD of at least 5x10-7 µCi/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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 operable channels is less than required, make repairs as soon as possible. Grab samples cannot be obtained via this pathway.

NOTE 1: 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 2: This is a shared system between Unit 1 and Unit 2.

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ATTACHMENT 3 (Page 1 of 1)

Surry Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Channel Description Channel Source Channel Channel Check Check Calibration Functional Test

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC 1ERMINATION OF RELEASE (a) Radwaste Facility Liquid Effluent Line RM-RRM-131 D p R Q

2. GROSS BETA OR GAMMA RADIOACTIV-ITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC 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 entLine l-SW-RM-107A D M R Q l-SW-RM-107B 1-SW-RM-107C l-SW-RM-107D

3. FLOW RATE MEASUREMENT DEVICES Radwaste Facility Liquid Effluent Line Instrument Loop RLW-153 DR NIA R NIA

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ATTACHMENT4 (Page 1 of 2)

North Anna Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Channel Description Channel Source Channel Channel Check Check Calibration Functional Test

1. Liquid Radwaste Effluent (a) 1-RM-LW-111,LiquidRadwasteEffluent D D R Q (NOTE 1)

Monitor (b) 1-LW-FT-104, Liquid Radwaste Effluent D (NOTE3) NIA 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) l-LW-TK-20, Liquid Waste Effluent D (NOTE9) 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 (t) 1-RM-SW-108, Service Water System D M R Q(NOTE2)

Effluent Monitor (g) 1-RM-SW-130, Unit 1 Circulating Water D M R Q(NOTE2).

System Effluent Line Monitor (h) 2-RM-SW-230, Unit 2 Circulating Water D M R Q(NOTE2)

System Effluent Line Monitor

2. 1 ank Level 10d1catmg uevice (NOTE6)

(a) Refueling Water Storage Tanks Unit 1 1-QS-LT-lOOA D (NOTE4) N/A R Q (NOTE 7) 1-QS-LT-lOOB 1-QS-LT-lOOC 1-QS-LT-lOOD Unit2 2-QS-LT-200A D(NOTE4) N/A R Q (NOTE 7) 2-QS-LT-200B 2-QS-LT-200C 2-QS-LT-200D (b) Casing Cooling Storage Tanks Unit 1 1-RS-LT-103A D (NOTE4) N/A R Q (NOTE 7) 1-RS-LT-103B Unit2 2-RS-LT-203A D(NOTE4) N/A R Q (NOTE 7) 2-RS-LT-203B (c) PG Water Storage Tanks (NOTE 5) 1-BR-LT-116A (1-PG-TK-lA) D(NOTE4) N/A R Q(NOTE8) 1-BR-LT-116B (1-PG-TK-lB) D (NOTE4) N/A R Q(NOTE8)

(d) Boron Recovery Test Tanks (NOTE 5)

1-BR-LT-112A (1-BR-TK-2A) l-BR-LT-112B (1-BR-TK-2B)

D (NOTE4)

N/A N/A R

R Q (NOTE 8)

Q (NOTE 8)

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 60 OF 156 ATTACHMENT 4 (Page 2 of 2)

North Anna Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements NOTE 1: The Channel Functional Test shall demonstrate:

a. Automatic isolation of this pathway and Control Room alarm annunciation occur if the instrument indicates measured levels above 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

NOTE 4: During liquid additions to the tank, verify indication of level change.

NQTE 5: This is a shared system between Unit 1 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.

VIRGINIA VPAP-2103 POWER REVISION?

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ATTACHMENT 5 (Page 1 of 1)

Liquid Ingestion Pathway Dose Factors for Surry Station Units 1 and 2 Total Body Ai Thyroid Ai GI-LLI Radionuclide mrem/hr mrem/hr mrem/hr

µCi/ml µCi/ml µCi/ml H-3 :./..~2E-01 2.82E-01 2.82E-01 Na-24 4.57E-01 4.57E-01 4.57E-01 Cr-51 5.58E+oo 3.34E-01 1.40E+03 Mn-54 1.35E+03 - 2.16E+04 Fe-55 8.23E+03 - 2.03E+04 Fe-59 7.27E+04 - 6.32E+05 Co-58 l.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+OO - 4.89E+04 Zr-95 3.46E+OO - 1.62E+04 Zr-97 8.13E-02 - 5.51E+04 Nb-95 l.34E+02 - 1.51E+06 Mo-99 2.43E+Ol - 2.96E+02 Ru-103 4.60E+Ol - 1.25E+04 Ru-106 2.01E+02 - 1.03E+05 Ag-llOm 8.60E+02 - 5.97E+05 Sb-124 1.09E+02 6.70E-01 7.84E+03 Sb-125 4.20E+Ol 1.79E-01 1.94E+03 Te-125m 2.91E+Ol 6.52E+Ol 8.66E+02 Te-127m 6.68E+Ol 1.40E+02 1.84E+03 Te-129m l.47E+02 3.20E+02 4.69E+03 Te-131m 5.71E+Ol 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+Ol I-132 9.96E+OO 9.96E+02 5.35E+OO I-133 3.95E+Ol l.90E+04 1.16E+02 I-134 5.40E+OO 2.62E+02 1.32E-02 I-135 2.24E+Ol 4.01E+03 6.87E+Ol Cs-134 l.33E+04 - 2.85E+02 Cs-136 2.04E+03 - 3.21E+02 Cs-137 7.85E+03 - 2.32E+02 Cs-138 5.94E+OO - 5.12E-05 Ba-140 l.08E+02 - 3.38E+03 La-140 2.lOE-01 - 5.83E+04 Ce-141 2.63E-01 - 8.86E+03 Ce-143 4.94E-02 1.67E+04 Ce-144 9.59E+OO - 6.04E+04 Np-239 1.91E-03 - 7.11E+02

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ATTACHMENT 6 (Page 1 of 5)

North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 1.0 Equation (6)

D tF~f.C.A. (6-1)

.£.J1 I I 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 Ci and F are averaged for all liquid releases, in hours F =the near field average dilution factor for Ci 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 q = average concentration of radionuclide i, in undiluted liquid effluent during the period t, from any liquid releases, in µCi/ml Ai =the site-related ingestion dose commitment factor tothe 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 µCi/ml A.I = 1.14 E+05 (730/Dw+ 21 BF./D I a

) DF.I (6-2) where:

1.14 E+o5 = (1 E+06 pCi/µCi x 1 E+o3 ml/kg)/8760 hr/yr~ units conversion factor 730 = adult water consumption, kg/yr, from NUREG-0133

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 64 OF 156 ATTACHMENT 6 (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.

Dw includes the dilution contributions from the Lake Anna Dam to Doswell (0. 73), the WH1F (Cr)Cc), and Lake Anna (CR/'CL). The potable water mixing ratio is calculated as:

where Cc, CL and CR are the respective concentrations for the considered nuclide in the discharge channel, WH1F (Lagoon) and the Lake.

Calculation is per expressions 11.2-5, 11.2-6, and 11.2-8 of the North Anna UFSAR 21 BFi

=

adult fish consumption rate, kg/yr, from NUREG 0133 the bioaccumulation factor for nuclide i, in fish, pCi/kg per pCi/1, from 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

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ATTACHMENT 6 (Page 3 of 5)

North Anna Liquid Ingestion Pathway Dose Factor CalculationUnits 1 and 2 2.0 Equation (9)

Equation (6) is simplified for actual dose calculations by introducing:

WASTE FLOW WASTE FLOW (6-4)

F = CIRC. (WATER) FLOW+ WASTE FLOW - CIR.C. FLOW and CIRC. FLOW

f. = (6-5) 1 EFFECTIVE DIL. FLOWi 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:

Q.1 = txC.xWASTEFLOW 1

(6-6) and Equation (6) reduces to:

A.

~ l D = f QiEFF. DIL. FLOWi (6-7)

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 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 of low 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 .

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 66 OF 156 ATTACHMENT 6 (Page4 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 UFSAR, 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 134 19 percent, Cs 137 15 percent).

By defining Bi= Ai./ EFF. DIL. FLOWi, the dose calculation is reduced to a two factor formula:

D = £.J ~Q.xB.

1 1 1 Values for Bi (mrem/Ci) and EFF. DIL. FLOWi are listed in Attachment 7.

(6-8)

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ATTACHMENT6 (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 (fi) µCi/ml µCi/ml H-3 14.9 6.18E+OO 6.18E+OO Na-24 1.0 3.71E+Ol 3.71E+Ol Cr-51 1.7 l.lOE+OO -

Mn-54 7.0 8.62E+02 4.52E+03 Fe-55 11.3 l.30E+02 5.56E+02 Fe-59 2.2 9.47E+02 2.47E+03 Co-58 2.8 2.49E+02 l.11E+02 Co-60 13.3 8.27E+02 3.75E+02 Zn-65 6.1 3.28E+04 . 7.25E+04 Rb-86 1.5 3.53E+04 7.59E+04 Sr-89 2.3 8.70E+02 -

Sr-90 15.8 2.39E+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 Nb-95 1.9 l.13E+02 2.10E+02 Mo-99 1.0 7.48E+OO 3.93E+Ol Ru-103 2.0 4.lOE+OO -

Ru-106 7.6 2.65E+Ol -

Ag-llOm 6.2 4.94E+OO 8.32E+OO Sb-124 2.6 4.37E+Ol 2.08E+OO Sb-125 11.4 2.46E+Ol 1.16E+OO Te-125m 2.5 3.23E+02 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 l.12E+02 1.35E+02 Te-132 1.0 5.04E+02 5.37E+02 I-131 1.2 9.66E+Ol 1.69E+02 I-132 1.0 1.03E-01 2.95E-01 I-133 1.0 3.47E+OO 1.14E+Ol I-134 1.0 2.15E-02 6.00E-02 I-135 1.0 6.58E-01 1.78E+OO Cs-134 10.3 5.80E+05 7.09E+05 Cs-136 1.3 6.01E+04 8.35E+04 Cs-137 15.8 3.45E+05 5.26E+05 Cs-138 1.0 9.18E-01 1.85E+OO Ba-140 1.3 2.65E+Ol 5.08E-01 La-140 1.0 4.47E-03 l.69E-02 Ce-141 1.8 2.14E-02 l.89E-01 Ce-143 1.0 1.35E-04 l.22E+OO Ce-144 6.6 l.41E+OO l.lOE+Ol Np-239 1.0 5.13E-04 9.31E-04 L

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ATTACHMENT7 (Page 1 of 1)

North Anna Liquid Pathway Dose Commitment Factors for Adults Bi= Ai Fi/CIRC FLOW= (A/Effective Dilution Flowi) X 9.81E-3 hr ft3 µCi/ 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-04 2.66E-04 Na-24 3.39E+03 1.0?E-04 1.0?E-04 Cr-51 1.99E+03 5.44E-06 NIA 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 l.20E+03 2.04E-03 9.lOE-04 Co-60 2.55E+02 3.18E-02 1.44E-02 Zn-65 5.60E+02 5.74E-01 1.27E+OO Rb-86 2.34E+03 l.48E-01 3.18E-01 Sr-89 1.46E+03 5.84E-03 NIA Sr-90 2.16E+02 l.09E+Ol NIA Y-91 l.34E+03 2.50B-06 NIA Zr-95 1.27B+03 2.30B-06 3.40B-06 Zr-97 3.39E+03 4.33B-10 9.46B-10 Nb-95 1.78B+03 6.24E-04 l.16B-03 Mo-99 3.30B+03 2.22B-05 l.17B-04 Ru-103 1.68B+03 2.40B-05 NIA Ru-106 4.48B+02 5.80E-04 NIA Ag-llOm 5.52B+02 8.78B-05 1.48B-04 Sb-124 1.32B+03 3.25B-04 1.55B-05 Sb-125 2.98B+02 8.lOE-04 3.80B-05 Te-125m l.35B+03 2.35B-03 6.35B-03 Te-127m 9.16B+02 8.37E-03 2.46B-02 Te-129m 1.82B+03 8.19E-03 1.93B-02 Te-131m 3.38B+03 3.27E-04 3.92B-04 Te-132 3.27E+03 1.51B-03 l.61E-03 1-131 2.94B+03 3.22B-04 5.62B-04 1-132 3.40E+03 2.98B-07 8.51B-07 1-133 3.39E+03 1.00E-05 3.29B-05 1-134 3.40B+03 6.19B-08 1.73B-07 1-135 3.40B+03 1.90B-06 5.15E-06 Cs-134 3.29B+02 1.73B+Ol 2.llE+Ol Cs-136 2.62B+03 2.25B-01 3.12B-01 Cs-137 2.15B+02 1.57B+Ol 2.40B+Ol Cs-138 3.40B+03 2.65B-06 5.34B-06 Ba-140 2.65B+03 9.83B-05 1.88B-06 La-140 3.36B+03 l.31E-08 4.94E-08 Ce-141 1.85B+03 l.14E-07 l.OOE-06 3.37E+03 Ce-143 3.93E-10 3.55E-06 Ce-144 5.14E+02 2.70E-05 2.lOE-04 Np-239 3.32E+03 1.51E-09 2.75E-09

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VIRGINIA VPAP-2103 POWER REVISION?

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ATTACHMENTS (Page 1 of 3)

Surry Radioactive Liquid Waste Sampling and Analysis Program Lower Limit of Liquid Release Sampling Minimum Analy- Type of Activity Detection (LLD)

Type Frequency sis Frequency Analysis (uCi/ml), (Note 1) p p Principle Gamma 5 X 10-7 Emitters (Note 3) 1131 1 X 10-6 (Each Batch) (Each Batch) p Dissolved and Batch Releases M Entrained Gases 1 X 10-5 (One Batch/M)

(Gamma Emitters)

(Note 2) p M Composite H3 1 X 10-5 (Each Batch) (Note 4) Gross Alpha 1 X 10-1

p (Each Batch)

Continuous Q Composite (Note 4)

W Composite Sr89 and Sr90 Fe55 Principal Gamma 5 X 10-8 1 X 10-6 5 X 10-7 Emitters (Note 6)

(Note 6) (Note 6) 1131 1 X 10-6 Dissolved and Continuous M 1 X 10-5 M Entrained Gases Releases Grab Sample (Gamma Emitters)

(Note 5) Continuous M Composite H3 1 X 10-5 (Note 6) (Note 6) Gross Alpha 1 X 10-?

Continuous Q Composite Sr89 and Sr 90 5 X 10-8 (Note 6) (Note 6) Fe55 1 X 10-6

VIRGINIA VPAP-2103 POWER REVISION?

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 Sb LLD - (8-1)

E

V

2.22E+06

Y

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 2.22E+06

=

the sample size (in units of mass* or volume) 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 appropriate methods will be used to obtain a representative sample for analysis.

VIRGINIA VPAP-2103 POWER REVISION?

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ATTACHMENT 8 (Page 3 of 3)

Surry Radioactive Liquid Waste Sampling and Analysis Program NOTE 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn54, Fe59 , Co 58 , Co 60, Zn 65 , Mo 99 , Cs 134, Cs 137, Ce 141 , and Ce 144 . 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 4: A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid 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.

NOTE 6: To be representative of the quantities and concentrations of radioactive materials in liquid

effluents, composite sampling shall employ appropriate methods which will result in a specimen representative of the effluent release.

. I

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l VIRGINIA VPAP-2103 POWER REVISION?

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ATTACHMENT 9 (Page 1 of 3)

North Anna Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sampling Minimum Analy- Type of Activity Lower Limit of Type Frequency sis Frequency Analysis Detection (LLD)

(µCi/ml), (Note 1) p p Principle Gamma 5 X 10-1 Emitters (Note 3)

(Each Batch) (Each Batch) 1131 i 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 10-1 p

(Each Batch)

Q Composite (Note 4)

Sr89 and Sr90 Fe55 Principal Gamma 5 X 10-8 1 X 10-6 5 X 10-7 Emitters (Note 6)

Continuous W Composite 1131 1 X 10-6 Continuous (Note 6) (Note 6) U1sso1vect anct 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 10-1 Continuous Q Composite Sr89 and Sr90 5 X 10-8 (Note 6) (Note 6) . Fe55 1 X 10-6

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 76 OF 156 ATTACHMENT 9 (Page 2 of3)

North Anna Radioactive Liquid Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):

4.66 Sb LLD= (9-1)

E

V

2.22E+06

Y

e- (11.At)

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.

VIRGINIA VPAP-2103 POWER REVISION?

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ATTACHMENT 9 (Page 3 of 3)

North Anna Radioactive Liquid Waste Sampling and Analysis Program NOTE 3: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn54 , Fe59 , Co58 , Co 60 , Zn65 , Mo 99 , Cs 134 , Cs 137 , Ce 141 , and Ce 144* 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 4: A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid 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.

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 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-s µCi/ml, the turbine building su~p 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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ATTACHMENT 10 (Page 1 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 (µCi/ml), (Note 1)

Prior to Release A. Waste Gas Prior to Release Principal Gamma (Each Tank) 1 X 10-4 Storage Tank (Each Tank) Emitters (Note 2)

(Grab Sample)

Principle Gamma Prior to Release Prior to Release 1 X 10-4 B. Containment Emitters (Note 2)

(Each PURGE) H3 Purge (Each PURGE) 1 X 10-6 (Grab Sample)

C. Ventilation Weekly Principle Gamma Weekly 1 X 10-4

(!)Process Vent (Grab Sample) Emitters (Note 2)

(2)Vent Vent#l (3)Vent Vent#2 (Note 3) (Note 3) H3 1 X 10-6 (4)SRF Vent 1131 1 X 10- 12 Continuous Weekly (Note 5)

(Note 4) (Charcoal Sample) 1133 1 X 10-lO Continuous Weekly (Note 5) Principal Gamma 1 X 10-ll All Release (Note 4) Particulate Sample Emitter (Note 2)

Weekly Continuous Types as listed Composite Gross Alpha 1 X 10-ll (Note 4)

Particulate Sample in A, B, and C Quarterly Continuous Composite Sr89 and Sr90 1 X 10-ll (Note 4)

Particulate Continuous Noble Gas Noble Gases Gross 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) H3 Ejector (Note 3) 1 X 10-6

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 80 OF 156 ATTACHMENT 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 (µCi/ml), (Note 1)

Principle Gamma Prior to Release Prior to Release 1 X 10-4 Emitters (Grab Sample) (Each Release) H3 1 X 10-6 1131 1 X 10-ll Continuous Charcoal Sample (Note 4) (Note 6) 1133 1 X 10-lO Containment Continuous Particulate Principal Gamma 1 X 10-lO HogDepres-(Note 4) Sample (Note 6) Emitter (Nate 2) surization Composite Continuous 1 X 10-lO Particulate Gross Alpha (Note 4)

Sample (Note 6)

Composite Continuous Particulate Sr89 and Sr90 1 X 10-lO (Note 4)

Sample (Note 6)

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ATTACHMENT 10 (Page 3 of 5)

Surry Radioactive Gaseous Waste Sampling and Analysis Program NOTE 1: For a particular measurement system (which may include radiochemical separation):

4.66 Sb LLD=~~~~~~~~~~- (10-1)

E

V

2.22E+06 ** Y

e- (A.At)

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 2.22E+o6 y

=

the sample size (in units of mass or volume).

the number of disintegrations per minute (dpm) per microcurie.

the fractional radiochemical yield (when applicable).

A = the radioactive decay constant for the particular radionuclide.

L\t = 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.

VIRGINIA VPAP-2103

POWER REVISION?

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 following radionuclides: Kr87 , Kr88 , Xe 133 , Xe 133 m, Xe 135 , Xe 135 m, and Xe 138 for gaseous emissions and Mn54, Fe59 , Co58 , Co60, Zn 65 , Mo99 , Cs 134, Cs 137, Ce 141 and Ce 144 for 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 analysis shall also be performed follqwing 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 1131 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 ratiq 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 .hours of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if:

a. Analysis shows that the dose equivalent I 131 concentration in the primary coolant has increased by a factor of 3; and

b. Noble gas monitor shows that effluent activity has increased more than a factor of 3 .

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ATTACHMENT 10 (Page 5 of 5)

Surry Radioactive Gaseous Waste Sampling and Analysis Program NOTE 6: To be representative of the quantities and concentrations of radioactive materials in gaseous effluents, composite sampling shall employ appropriate methods that will result in a specimen representative of the effluent release .

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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 (µCi/ml), (Note 1)

Prior to Release A. Waste Gas Prior to Release Principal Gamma (Each Tank 1 X 10-4 Storage Tank (Each Tank) Emitters (Note 2)

Grab Sample)

Principle Gamma Prior to Release Prior to Release 1 X 10-4 B. Containment Emitters (Note 2)

(Each PURGE Purge (Each PURGE) H3 1 X 10-6 Grab Sample)

C. Ventilation Monthly Principle Gamma Monthly 1 X 10-4 (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) 1131 1 X 10- 12 Continuous Weekly (Note 4) (Charcoal Sample) 1133 1 X 10-lO Continuous Weekly Principal Gamma 1 X 10-ll All Release (Note 4) Particulate Sample Emitter (Note 2)

Monthly Continuous 1 X 10-ll Types as listed Composite Gross Alpha (Note 4)

Particulate Sample in A, B, and C Quarterly Continuous Composite Sr89 and Sr90 1 X 10-ll (Note 4)

Particulate Continuous Noble Gas Noble Gases Gross 1 X 10-6 (Note4) Monitor Beta or Gamma Condenser Air Principle Gamma Weekly Weekly 1 X 10-4 Ejector/Steam Emitters (Note 7)

Generator Grab Sample H3 1 X 10-6 Blowdown Vent (Note 6)

Containment Principle Gamma Prior to Release Prior to each 1 X 10-4 Vacuum Steam Emitters (Note 2)

Ejector (Hogger) (Grab Sample)

Release H3 1 X 10-6 (Note 8)

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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 Sb LLD= (11-1)

E

V

2.22E+06

Y

e- (lAt)

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)

A = the radioactiv.e 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 as "posteriori" (after the fact) limit for a particular measurement.

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ATTACHMENT 11 (Page 3 of 4)

North Anna Radioactive Gaseous Waste Sampling and Analysis Program NOTE 2: The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr87 , Kr88 , Xe 133 , Xe 133 m, Xe 135 , Xe 135 m, and Xe 138 for gaseous emissions and Mn54, Fe59 , Co58 , Co 60, Zn65 , Mo 99 , Cs 134, Cs 137 , Ce 141 and Ce 144 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 I 131 concentration in the primary coolant is greater than 1.0 µCi/gm; 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 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if:

a. Analysis shows that the dose equivalent I 131 concentration in the primary coolant is greater than 1.0 µCi/gm and;

b. Noble gas monitor shows that effluent activity has increased more than a factor of 3 .

VIRGINIA VPAP-2103

POWER REVISION?

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-s µCi/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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

NOTE 7: The principal gamma emitters for which the LLD specification applies exclusively are the following radiomiclides: Kr87 , Kr88 , Xe 133 , Xe 133m, Xe 135 , Xe 135m, and Xe 138 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 l.OE-5 µCi/ml, Steam Generator samples shall be obtained and analyzed prior to release.

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ATTACHMENT 12 (Page 1 of 3)

Gaseous Effluent Dose Factors for Surry (Gamma and Beta Dose Factors) xJQ = 6.0E-05 sec/m3 at 499 meters N Direction Dose Factors for Ventilation Vent Noble Gas Kivv ~vv Mivv Nivv Radionuclide Total Body Skin Gamma Air Beta Air mrem/yr mrem/yr mrad/yr mrad/yr Curie/sec Curie/sec Curie/sec Curie/sec Kr-83m 4.54E+OO - 1.16E+03 1.73E+04 Kr-85m 7.02E+04 8.76E+04 7.38E+04 1.18E+05 Kr-85 9.66E+02 8.04E+04 1.03E+03 l.17E+05

- Kr-87 Kr-88 Kr-89 3.55E+05 8.82E+05 9.96E+05 5.84E+05 l.42E+05 6.06E+05 3.70E+05 9.12E+05 1.04E+06 6.18E+05 1.76E+05 6.36E+05 Kr-90 9.36E+05 4.37E+05 9.78E+05 4.70E+05 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 Xe-135 l.09E+05 1.12E+05 1.15E+05 1.48E+05 Xe-137 8.52E+04 7.32E+05 9.06E+04 7.62E+05 Xe-138 5.30E+05 2.48E+05 5.53E+05 2.85E+05 Ar-41 5.30E+05 l.61E+05

5.58E+05 1.97E+05

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 90 OF 156 ATTACHMENT 12 (Page 2 of 3)

Gaseous Effluent Dose Factors for Surry (Gamma and Beta Dose Factors) xJQ = 1.0E-06 sec/m3 at 644 meters N Direction Dose Factors for Process Vent Noble Gas Kipv Ljpv Mipv Nipv Radionuclide Total Body Skin GammaAir Beta Air mrem/yr mrem/yr mrad/yr mrad/yr Curie/sec Curie/sec Curie/sec . Curie/sec Kr-83m. 7.56E-02 - 1.93E+Ol 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+Ol 1.34E+03 1.72E+Ol 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 Kr-89 Kr-90 1.47E+04 1.66E+04 1.56E+04 2.37E+03 1.01E+04 7.29E+03 1.52E+04 1.73E+04 1.63E+04 2.93E+03 1.06E+04 7.83E+03 Xe-131m 9.15E+Ol 4.76E+02. 1.56E+02 1.11E+o3 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 l.05E+03 Xe-135m 3.12E+03 7.11E+o2 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

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ATTACHMENT 12 (Page 3 of 3)

Gaseous Effluent Dose Factors for Surry (Inhalation Pathway Dose Factors)

Ventilation Vent xJQ = 6.0E-05 sec/m3 at 499 meters N Direction Process Vent xJQ = 1.0E-06 sec/m3 at 644 meters S Direction Radionuclide Pivv Pipv mrem/yr mrem/yr Curie/sec Curie/sec H-3 6.75E+04 1.12E+03 Cr-51 5.13E+03 8.SSE+Ol Mn-54 ND ND.

Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND

Rb-86 Sr-90 Y-91 Zr-95 ND ND ND ND ND ND ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm 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 1-131 9.75E+08 1.62E+07

1-133 2.31E+08 3.85E+06 ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1

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PAGE 93 OF 156

ATTACHMENT 13 (Page 1 of 3)

Gaseous Effluent Dose Factors for North Anna (Gamma and Beta Dose Factors) xJQ = 9.3E-06 sec/m3 at 1416 meters SE Direction Dose Factors for Ventilation Vent Noble Gas Kivv ~vv Mivv Nivv Radionuclide Total Body Skin Gamma Air Beta Air mrem/yr mrem/yr mrad/yr mrad/yr Curie/sec Curie/sec Curie/sec Curie/sec Kr-83m 7.03E-01 - l.79E+02 2.68E+o3 Kr-85m 1.09E+04 1.36E+04 l.14E+04 l.83E+04 Kr-85 l.50E+02 l.25E+04 l.60E+02 l.81E+04 Kr-87 5.51E+04 9.05E+04 5.74E+04 9.58E+04

Kr-88 Kr-89 Kr-90 1.37E+05 1.54E+05 l.45E+05 2.20E+04

9.39E+04 6.78E+04 l.41E+05 l.61E+05 l.52E+05 2.72E+04 9.86E+04 7.28E+04 Xe-131m 8.51E+02 4.43E+03 l.45E+03 1.03E+04 Xe-133m 2.33E+03 9.24E+03 3.04E+03 l.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 Xe-135 l.68E+04 l.73E+04 l.79E+04 2.29E+04 Xe-137 l.32E+04 l.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 I_

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 94 OF 156 ATTACHMENT 13 (Page 2 of 3)

Gaseous Effluent Dose Factors for North Anna (Gamma and Beta Dose Factors)

'X}Q = 1.2E-06 sec/m3 at 1513 meters S Direction**

Dose Factors for Process Vent Noble Gas Kipv Lipv. Mipv Nipv Radionuclide Total Body Skin Gamma Air Beta Air mrem/yr mrem/yr mrad/yr mrad/yr Curie/sec Curie/sec Curie/sec Curie/sec Kr-83m 9.07E-02 - 2.32E+Ol 3.46E+o2**

Kr-85m 1.40E+03 1.75E+03 1.48E+03 2.36E+03 Kr-85 1.93E+Ol 1.61E+03 2.06E+Ol 2.34E+03

.Kr-87 7.10E+03 1.17E+04 7.40E+03 1.24E+04

Kr-88 Kr-89 Kr-90 1.76E+04 1.99E+04 1.87E+04 2.84E+03 1.21E+04 8.75E+03 1.82E+04

. 2.08E+04 1.96E+04 3.52E+03 1.27E+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 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 l.70E+03 l.46E+04 .. .

1.81E+03 1.52E+04*

Xe-138 1.06E+04 4.96E+03 . 1.11E+04 5.70E+03 Ar-41 1.06E+04 3.23E+03 1.12E+04 3.94E+03

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 95 OF 156

ATTACHMENT 13 (Page 3 of 3)

Gaseous Effluent Dose Factors for North Anna (Inhalation Pathway Dose Factors)

Ventilation Vent x}Q = 9.3E-06 sec/m3 at 1416 meters SE Direction Process Vent x}Q = 1.2E-06 sec/m3 at 1513 meters S Direction Radionuclide Pivv Pipv mrem/yr mrem/yr 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 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-llOm 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 Ce-144 ND ND 1-131 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

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ATTACHMENT 14 (Page 1 of 3)

Surry Radioactive Gaseous Effluent 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 l-GW-RM-130-1 (b) Iodine Sampler Process Vent Continuous HP Sampler, or l-GW-RM-130-1 1 2 (c) Particulate Sampler Process Vent Continuous HP Sampler, or 1 2 l-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 1-SV-RM-111 2 (one per unit) 1 2-SV-RM-211 (b) Air Ejector Flow Rate Measuring Device Unit 1: 1-VP-FI-lA 1-VP-FI-lB 2 (one per unit) 3 Unit 2: 2-VP-FI-lA 2-VP-FI-lB

3. VENTILATION VENT SYSTEM (a) Noble Gas Activity Monitor SRF: RRM-101 1 1 SPS: Vent #l l-VG-RM-104 1 1 Vent #2, 1-VG-RM-110, or .1 1 1-VG-RM-131-1

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 98 OF 156 ATTACHMENT 14 (Page 2 of3)

Surry Radioactive Gaseous Effluent Monitoring Instrumentation MINIMUM INSTRUMENT OPERABLE ACTION CHANNELS (b) Iodine Sampler SRF:RRM-101 1 2 SPS: Vent#l, 1-VG-RM-104 1 2 Vent #2, Continuous HP Sampler, or 1 2 1-VG-RM-131-1 (c) Particulate Sampler SRF: RRM-101 1 2 SPS: Vent#l, VG-RM-104 1 2 Vent #2, HP Continuous Sampler, or 1 2 l-VG-RM-131-1 (d) Ventilation Vent Flow Rate Monitor SRF: 01-RHV-FT-156 1 3 (e)

SPS: Vent#l, 1-VS-FT-119 Vent#2, 1-VS-FT-116 Sampler Flow Rate Measuring Device SRF: RRM-101 SPS: Vent#l, 1-VG-RM-104 1

1 1

1 3

3 3

3 Vent #2, KAMAN Flow Rate Measuring Device 1 3 (Parameter #19), or HP Sampler Rotometer

__ J

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ATTACHMENT 14 (Page 3 of 3)

Surry Radioactive Gaseous Effluent 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 2: If the number of operable channels is less than required, effluent releases via the 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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ATTACHMENT 15 (Page 1 of 3)

North Anna Radioactive Gaseous Effluent Monitoring Instrumentation INSTRUMENT MINIMUM ACTION 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 1-RM-GW-178-1 1 2, 5 Process Vent Continuous HP Sampler (d) Total Flow Monitor 1-GW-Ff-108 1 1 (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring Device (Parameter 19)

HP Sampler Rotameter 1 1

2. CONDENSER AIR EJECTOR SYSTEM (a) Gross Activity Monitor Unit 1 1-SV-RM-121 Unit 2 2-SV-RM-221 1 3 (b) Flow Rate Measuring Device Unit 1 1-SV-FI-lOOA 1-SV-FI-lOlA 1-SV-FI-lOOB 1 (NOTE 1) 1 1-SV-FI-lOlB Unit2 2-SV-FI-200A 2-SV-FI-201A 2-SV-FI-200B 1 (NOTE 2) 1

2-SV-FI-201B

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PAGE 102 OF 156 ATTACHMENT 15 (Page 2 of 3)

North Anna Radioactive Gaseous Effluent Monitoring Instrumentation MINIMUM INSTRUMENT OPERABLE ACTION CHANNELS

3. VENTILATION VENT A (a) Noble Gas Activity Monitor 1-~~\f(}-104 1-~-\f<3-179-1 1 2 (b) Iodine Sampler 1-~-\f<3-179-1

\Tent \Tent A Continuous HP Sampler 1 2 (c) Particulate Sampler 1-RM-\1'<3-179-1 1 2

\Tent \Tent A Continuous HP Sampler (d) . Total Flow Monitor 1-H\T-Ff-1212A 1 1 (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring Device (Parameter. 19) 1 1 HP Sampler Rotameter

4. VENTILATION VENT B (a) Noble Gas Activity Monitor 1-RM-\1'<3-113 1-RM-VCi-180-1 1 2 (b) Iodine Sampler .

.1-~-VCi-180-1

\Tent \Tent B Continuous HP Sampler 1 2 (c) Particulate Sampler 1-~-\f(}-180-1 1 2

\Tent \Tent B Continuous HP Sampler (d)

Total Flow Monitor 1-H\T-Ff-1212B. 1 1 (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring Device (Parameter 19) 1 1 HP Sampler Rotameter

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ATTACHMENT 15 (Page 3 of 3)

North Anna Radioactive Gaseous Effluent Monitoring Instrumentation ACTION 1: If the number of operable channels is less than required, effluent releases, via this path, may continue if the flow rate is estimated at least once per four hours.

ACTION 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity or gamma isotopic activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 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.

ACTION 4: If the number of operable channels is less than required, the contents of the Waste Gas Decay Tanks may be released 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 staff independently 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-lOOA, 101A), or Train B (1-SV-FI-lOOB, 101B) or both.

b. Flow instruments 101A and 101B provide low range measurement Flow instruments 1OOA and 1OOB 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, 201A), or Train B (2-SV-FI-200B, 201B) or both .

b. Flow instruments 201A and 201B provide low range measurement. Flow instruments 200A and 200B provide high range measurement.

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_j

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ATTACHMENT 16 (Page 1 of 2)

Surry Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL SOURCE CHANNEL CHANNEL DESCRIPTION . CHECK CHECK CALIBRATION FUNCTIONAL TEST

1. PK' II .... ~~ VbNl :Sl'.:SfhM (a) Noble Gas Activity Monitor -

Providing Alann and Automatic Termination of Release 1-GW-RM-102 D M* R Q l-GW-RM-130-1 '

(b) Iodine Sampler Process Vent Continuous HP Sampler, or 1-GW-RM- 130-1 w NIA NIA NIA (c) Particulate Sampler Process Vent Continuous HP Sampler, or 1-GW-RM- 130-1 w NIA NIA NIA

(d) Process Vent Flow Rate Monitor 1-GW-Ff-100 (e) Sampler Flow Rate Measuring Device HP Sampler Rotometer, or KAMAN Flow Rate Measuring Device (Parameter#19)

D

D D

NIA NIA NIA R

SA R

NIA NIA NIA

2. CONDENSER AIR EJECTOR SYSTEN.

(a) Gross Activity Monitor Unit 1: 1-SV-RM-111 D M R Q Unit 2: 2-SV-RM-211 (b) Air Ejector Flow Rate Measuring

Device Unit 1: 1-VP-FI-lA 1-VP-FI-lB Unit 2: 2-VP-FI-lA D NIA R NIA 2-VP-FI-lB

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

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PAGE 106 OF 156 ATTACHMENT 16 (Page 2 of 2)

Surry Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL DESCRIPTION CHECK CHECK CALIBRATION TEST (b) Iodine Sampler SRF: RRM-101 SPS: Vent#l, 1-VG-RM-104 .

Vent#2,ContinuousHP Sampler, or 1-VG-RM-131-w NIA NIA NIA 1

(c) Particulate Sampler

.SRF: RRM-101 SPS: Vent #1, 1-VG-RM-104 -

Vent#2,ContinuousHP Sampler, or 1-VG-RM-131-w NIA .NIA NIA 1

(d) Ventilation Vent Flow Rate Monitor SRF:. 01-RHV-Fr-156 SPS: Vent#l, 1-VS-Fr-119 Vent#2, 1-VS-Fr-116 (e) Sampler Flo.w Rate Measuring Device D NIA R NIA

SRF: RRM-101 D NIA R NIA SPS: Vent#l, 1-VG-RM-104 D NIA :R NIA Vent#2, KAMAN Flow Rate D NIA R NIA Measuring Device

(Parameter#19), or HP D NIA NIA Sampler Rotometer SIA

... rnor to eacn waste uas Uecay *rank release

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ATTACHMENT 17 (Page 1 of 3)

North Anna Radioactive Gaseous Effluent 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(NOTE5) R Q (NOTE 1) 1-RM-GW-178-1 D M(NOTE5) R Q (NOTE 1)

(b) Iodine Sampler l,-RM-GW-178-1 w NIA NIA NIA Process Vent Continuous HP Sampler D(NOTE3) NIA NIA NIA (c) Particulate Sampler 1-RM-GW-178-1 w NIA .NIA NIA

(d)

(e)

Process Vent Continuous HP Sampler Total Flow Monitor l-GW-Ff-108 Sampler Flow R~1te Measuring D(NOTE3)

D NIA NIA NIA R

NIA Q

Device KAMANS Flow Rate Measuring D(NOTE3) NIA R . NIA Device (Parameter 19)

HP Sampler Rotameter D(NOTE3) NIA SA NIA

2. CONDENSER AIR EJECTOR SYSTEM (a) Noble Gas Activity Monitor Unit l 1-SV-RM-121 D M R Q (NOTE 1)

Unit 2 2-SV-RM-221 (b) Flow Rate Measuring Device Unit 1 1-SV-FI-lOOA 1-SV-FI-lOlA 1-SV-FI-lOOB D NIA R NIA 1-SV-FI-lOlB Unit2 2-SV-FI-200A

2-SV-FI-201A 2-SV-FI-200B 2-SV-FI-201B D NIA R NIA

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North Amfa Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL CHANNEL SOURCE CHANNEL DESCRIPTION CHECK CHECK FUNCTIONAL CALIBRATION TEST

3. VENTILATION VENT A (a) Noble Gas Activity Monitor 1-RM-VG-104 D M R Q(NOTE2) 1-RM-VG-179-1 D M(NOTE5) R Q(NOTE2)

(b) Iodine Sampler 1-RM-VG-179-1 w NIA NIA NIA Vent Vent A Continuous HP Sampler D(NOTE3) NIA NIA NIA (c) Particulate Sampler 1-RM-VG-179-1 w NIA NIA NIA Vent Vent A Continuous ijP Sampler D(NOTE3) NIA NIA NIA (d) Total Flow Monitor 1-HV-Ff-1212A D NIA R Q (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring D (NOTE 3) NIA R NIA Device (Parameter 19)

HP Sampler Rotameter D(NOTE3) NIA SA NIA

4. VENTILATION VENT B I (a) Noble Gas Activity Monitor i'

i l-RM-VG'-113 D M R Q (NOTE4)

I l-RM-VG-180-1 D M(NOTE5) R Q (NOTE2)

(b) Iodine Sampler 1-RM-VG-180-1 w NIA NIA NIA Vent Vent B Continuous HP Sampler D(NOTE3) NIA NIA NIA (c) Particulate Sampler 1-RM-VG-180-1 w NIA NIA NIA Vent Vent B Continuous HP Sampler D(NOTE3) NIA NIA NIA (d) Total Flow Monitor 1-HV-Ff-1212B D NIA R Q (e) Sampler Flow Rate Measuring Device KAMANS Flow Rate Measuring D(NOTE3) NIA R NIA Device (Parameter 19)

HP Sampler Rotameter D(NOTE3) NIA SA NIA

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ATTACHMENT 17 (Page 3 of 3)

North Anna Radioactive Gaseous Effluent 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 alann/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 alann/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.

Channel Checks shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 l-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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. _J

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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 RMivv RMipv 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+Ol 3.08E+Ol 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-llOm 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 Ce-141 ND ND Ce-144 ND ND ND - No data for dose factor accordmg to Regulatory Guide 1.109, Rev1s10n 1

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Critical Organ and Inhalation Dose Factors for Surry (Inhalation Pathway Dose Factors)

Ventilation Vent xfQ = 3.0E-07 sec/m3 at 5150 meters S Direction Process Vent x/Q = 1.3E-07 sec/m3 at 5150 meters S Direction Radionuclide RMivv RMipv mrem/yr mrem/yr Curie/sec Curie/sec H-3 1.94E+02 8.41E+Ol Cr-51 1.73E+Ol 7.48E+OO 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-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-llOm 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 1-133. 1.07E+06 4.63E+05 Cs-134 ND ND Cs-136 ND ND Cs-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND ND - No data for dose factor according to Regulatory Guide 1.109, Revision 1

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ATTACHMENT 19 (Page 1 of 1)

Critical Organ Dose Factors for North Anna (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 RMivv RMipv mrem/yr mrem/yr Curie/sec Curie/sec H-3 1.73E+03 9.36E+02 Mn-54 ND ND Fe-59 ND ND Cr-51 1.50E+02 6.89E+Ol Co-58 ND ND Co-60 ND ND Zn-65 ND ND

Rb-86 Sr-89 Sr-90 Y-91 ND ND ND ND ND ND ND ND Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-llOm ND ND Te-127m l.97E+05 9.04E+04 Te-129m 2.95E+05 1.35E+05 1-131 1.45E+09 6.72E+08 l:-133 1.33E+07 6.12E+06 Cs-134 ND ND Cs;.136 ND ND -

Cs-137 ND ND*

Ba-140 ND ND

Ce-141 Ce-144 ND

-ND ND - No data for dose factor accordmg to Regulatory Gmde 1.109, Rev1s1on 1 ND ND l

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ATTACHMENT 20 (Page 1 of 3)

Surry Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis

1. DIRECT RADIATION About 40 Routine Monitoring Stations to be placed as follows:

1) Inner Ring in general 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

3) The balance of the 8 dosimeters should be placed in special interest areas such as population centers, nearby residents, schools, and in 2 or 3 areas to serve as controls

2. AIRBORNE Samples from 7 locations:

a) 1 sample from close to the site boundary location of the highest Radioiodine Canister calculated annual 1131 Analysis Weekly average ground level Continuous Radioiodines and D/Q Sampler Particulates b) 5 sample locations 6-8 operation with Particulate Sampler km distance located in a sample collection Gross beta radioactivity concentric ring around weekly analysis following filter the Station change; c) 1 sample from a control location 15-30 km Gamma isotopic analysis distant, providing valid of composite (by background data location) quarterly

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PAGE 116 OF 156 ATTACHMENT20 (Page 2 of 3)

Surry Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis

3. WATERBORNE Gamma isotopic analysis a) 1 sample upstream monthly; a) Surface Monthly Sample b) 1 sample downstream Composite for tritium analysis quarterly Sample from 1 or 2 sources Gamma isotopic and tritium b) Ground Quarterly analysis quarterly c) Sediment 1 sample from downstream area with existing or Gamma isotopic analysis from Semi-Annually potential recreational value semi-annually shoreline 5 samples from vicinity of .. ,.

Gamma isotopic analysis d) Silt Semi-Annually the Station semi-annually

4. INGESTION a) Mille (NOTE 1) a) 3 samples from milking animals in the vicinity of the Station b) 1 sample from milking animals at a control Monthly Gamma isotopic and analysis monthly I 131 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) 1 sampling of crabs from the vicinity of the Annually Gamma isotopic on edibles Station d) 2 samples of fish from the vicinity of the Semi-Annually Gamma isotopic on edibles Station (catfish, white perch, eel)

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ATTACHMENT 20 (Page 3 of 3)

Surry Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Collection Type and Frequency of and/or Sample Sample Location Frequency Analysis

4. INGESTION (Continued) a) 1 sample com Gamma isotopic on edible b) 1 sample soybeans Annually portion c) 1 sample peanuts d) 1 sample of a broadleaf vegetation of two different available offsite locations with highest annual average c) Food ground level D/Q, if one Products or more milk samples Monthly, if Gamma isotopic and I 131 are unavailable available, or at analysis e) 1 sample of a broadleaf harvest vegetation grown 15-30 km in the available, least prevalent wind direction, if one or more milk samples are unavailable 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 (N01E 2) Frequency Analysis

1. DIRECT RADIATION (N01E 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 meteorological sector within the site boundary

2) An outer ring of stations, one in each GAMMA DOSE emergency meteorological sector Quarterly Quarterly 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

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PAGE 120 OF 156 ATTACHMENT 21 (Page 2 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

2. AIRBORNE Samples from 5 locations:

a) 3 samples from close to

. the 3 site boundary locations (in different sectors) of the highest Radioiodine Canister 1131 Analysis, weekly calculated historical annual average ground Continuous Radioiodines and levelD/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) 1 sample from a control Gamma isotopic analysis location 15-40 km of composite (by distant and in the least location) quarterly prevalent wind directim1 (NOTES)

3. WATERBORNE Samples from 3 locations:

Gamma isotopic analysis a) 1 sample upstream monthly; (NOTE 5) a) Surface b) 1 sample downstream Grab Monthly Composite for tritium c) 1 sample from cooling analysis quarterly lagoon Sample from 1 or 2 sources Gamma isotopic and tritium b) Ground only if likely to be affected Grab Quarterly analysis quarterly (NOTE 5) 1 sample from downstream Gamma isotopic analysis c) Sediment area with existing or Semi-Annually semi-annually (NOTE 5) potential recreational value

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ATTACHMENT 21 (Page 3 of 5)

North Anna Radiological Environmental Monitoring Program Exposure Pathway Number of Sample and Sample Collection Type and Frequency of and/or Sample Location (NOTE 2) Frequency Analysis

4. INGESTION a) Samples from milking animals in 3 locations within 5 km that have the highest potential. If there are none, then 1 sample from milking animals in each of 3 areas a) Mille between 5 to 8 km where Monthly at all Gamma isotopic. (NOTE 5)

(NOTE 7) doses are calculated to be times and 1131 analysis monthly greater than 1 mrem per yr (NOTE6) b) 1 sample from milking

animals at a control location

( 15-30 km in the least prevalent wind direction) a) 1 sample of commercially and recreationally important species(bass,sunfish,catfish) b) Fish and in vicinity of plant discharge Gamma isotopic on edible Semiannually Invertebrates area portions b) 1 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 Gamma isotopic (NOTE 5) available, or Products sampling _is not performed and 1131 analysis at harvest b) 1 sample of broad leaf vegetation grown 15-30 km in the least prevalent wind direction if milk sampling is not performed

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PAGE 122 OF 156 ATTACHMENT 21 (Page 4 of 5)

North Anna Radiological Environmenta, 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 conditions, 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 shall be documented in the Annual Radiological Environmental Operating Report pursuant to 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 alternative media and locations may be chosen for the particular pathway 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 reflecting the new locations in the report.

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ATTACHMENT 21 (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 dose rate continuously may be used in place of, or in addition to, integrating dosimeters.

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 are not an absolute number. The 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 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.

NOTE 4: Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

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)

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ATTACHMENT 22 (Page 1 of 4)

Surry Environmental Sampling Locations SAMPLE LOCATION DISTANCE DIRECTION REMARKS 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 Onsite **

Environmental Control (00)

TLDs West North West (02) 0.17 WNW Site Boundary Surry Station Discharge (03) 0.6 NW Site Boundary North North West (04) 0.4 NNW Site Boundary North (05) 0.29 N Site Boundary North North East (06) 0.28 NNE Site Boundary North East (07) 0.31 NE Site Boundary East North East (08) 0.43 ENE Site Boundary 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 South South East (15) 0.74 SSE Site Boundary South East (16) 1.00 SE Site Boundary East (17) 0.57 E Site Boundary Station Intake (18) 1.23 ESE Site Boundary

Hog Island Reserve (19) 1.94 NNE Near Resident

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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 (21) 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 Kingsmill North Budweiser Water Plant (30)

(31)

(32)

(33) 7.8 5.6 5.7 4.8 NNE NNE N

NE Population Center Approx. 5 miles Population Center 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 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 Mille Epp's 4.8 SSW Colonial Parkway 3.7 NNW Judkin's 6.2 SSW Wtlliam's 22.5 s Control Location

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ATTACHMENT 22 (Page 3 of 4)

Surry Environmental Sampling Locations SAMPLE LOCATION DISTANCE DIRECTION REMARKS MEDIA (MILES)

Well Water Surry Station Onsite***

Hog Island Reserve 2.0 NNE 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 Control Location Lucas's Garden (Chester, Va.)

River Water Surry Discharge 0.17 NW

(Monthly)

Sediment (Silt)

Scotland Wharf Chickahominy River Surry Station Intake Surry Station Discharge 5.0 11.2 1.9 1.0 WNW WNW ESE NNW Control Location Control Location Hog Island Point 2.4 NE Point of Shoals 6.4 SSE

VIRGINIA VPAP-2103 POWER REVISI0N7 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 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

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ATTACHMENT 23 (Page I of 4)

North Anna Environmental Sampling Locations Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Environment NAPS Sewage Treannent Quarterly &

01 0.20 NE On-Site al Plant Annually TLDs Quarterly &

Frederick's Hall 02 5.30 SSW Annually Mineral,VA 03 7.10 WSW Quarterly &

Annually Wares Crossroads 04 5.10 WSW Quarterly &

Annually Route 752 05 4.20 NNE Quarterly &

Annually Sturgeon's Creek Marina 05A 3.20 N Quarterly &

Annually Levy, VA 06 4.70 ESE Quarterly &

Annually 07 7.30 SSE Quarterly &

B~pass, VA Annually End of Route 685 21 1.00 WNW Quarterly & Site Boundary Annually Route 700 22 1.00 WSW Quarterly & Site Boundary Annually "Aspen Hills" 23 SSE Quarterly &

0.93 Site Boundary Annually Orange, VA 24 22.00 NW Quarterly & Control Annually Bearing Cooling Tower N-1/33 0.06 N Quarterly On-Site Sturgeon's Creek Marina N-2/34 3.20 N Quarterly Parking Lot "C" NNE-3/35 0.24 NNE Quarterly On-Site Good Hope Church NNE-4/36 4.96 NNE Quarterly Parking Lot "B" NE-5/37 0.20 NE Quarterly On-Site Bogg's Drive NE-6/38 1.46 NE Quarterly Weather Tower Fence ENE-7/39 0.36 ENE Quarterly On-Site Route 689 ENE-8/40 2.43 ENE Quarterly Near Training Facility E-9/41 0.30 E Quarterly On-Site "Morning Glory Hill" E-10/42 2.85 E. Quarterly Island Dike ESE-11/43 0.12 ESE Quarterly On-Site Route 622 ESE-12/44 4.70 ESE Quarterly

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North Anna Environmental Sampling Locations Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Environment Biology Lab SE-13/45 0.75 SE Quarterly On-Site al TLDs Route 701 (Dam Entrance SE-14/46 5.88 SE Quarterly "Aspen Hills" SSE-15/47 0.93 SSE Quarterly. Site Boundal)

Elk Creek. SSE-16/48 2.33 SSE Quarterly NAPS Access Road S-17/49 0.47 s Quarterly On-Site Elk Creek Church . S-18/50 1.55 s Quarterly NAPS Access Road SSW-19/51 0.42 SSW Quarterly On-Site Route 618 SSW-20/52 5.30 SSW Quarterly 500KVTower SW-21/53 0.60 SW Quarterly On-Site Route 700' SW-22/54 4.36 SW Quarterly NAPS Radio Tower Route 700 South Gate of Switchyard Route 685 End of Route 685 WSW-23/55 WSW-24/56 W-25/57 W-26/58 WNW-27/59 0.38 1.00 0.32 1.55 1.00 wsw**

WSW WNW w

w Quarterly On-Site Quarterly SiteBo~dW)

Quarterly On-Site Quarterly Quarterly Site Boundal)

Route 685 WNW-28/60 1.40 WNW Quarterly I Laydown Area North Gate NW-29/61 0.45 NW Quarterly On-Site I Lake Anna Campground NW-30/62 2.54 NW Quarterly

1/#2 Intake NNW-31/63 0.07 NNW Quarterly On-Site Route208 NNW-32/64 3.43 NNW Quarterly Bumpass Post Office C-1/2 7.30 SSE Quarterly Control Orange, VA C-3/4 22.00 NW Quarterly Control Mineral, VA C-5/6 7.10 WSW Quarterly Control Louisa, VA C-7/8 11.54 WSW Quarterly Control

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ATTACHMENT23 (Page 3 of 4)

North Anna Environmental Sampling Locations Distance and Direction From Unit No. 1 Sample Location Station Distance Direction Collection REMARKS Media No. (Miles) Frequency Airborne NAPS Sewage Treatment 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 Radioiodine Wares Crossroads 04 5.10 WNW Weekly 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 Route 700 22 1.00 WSW Weekly Site Boundary "Aspen Hills" 23 0.93 SSE Weekly Site Boundary Orange, VA 24 22.00 NW Weekly Control Surface Water Waste Heat Treatment

[Commitment Facility (Second Cooling 08 1.10 SSE Monthly 3.2.2) Lagoon)

North Anna River (upstream) Rt 669 Bridge 09A 12.9 WNW Monthly Control (Brook's Bridge)

North Anna River 11 5.80 SE Monthly (downstream)

Ground Water OlA 0.75 SE Quarterly Biology Lab (well water)

. Aquatic Waste Heat Treatment Sediment Facility (Second Cooling 08 1.10 SSE Semi-Annually Lagoon)

North Anna River (upstream) Rt 669 Bridge 09A 12.9 WNW Semi-Annually Control (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 Sewage Treatment 01 0.20 NE Once per 3 yrs On-Site Plant

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North Anna Environmental Sampling Locations Distance and Direction From Unit No. 1 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 Creek 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 Route700 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. Goodwin)

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) '

Semi-

Lake Orange 25 16.50 NW Control Annually Food Products Route 713 14 varies NE (Broad Leaf Route 614 15 varies SE Monthly vegetation) Route 629/522 16 varies NW if available, or Control Route 685 21 varies WNW at harvest "Aspen Hills" Area 23 varies SSE

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ATTACHMENT 24 (Page 1 of 2)

Detection Capabilities for Surry Environmental Sample Analysis

.LOWER LIMIT OF DETECTION (LLD)

Airborne Food Fish Sediment Analysis Water Particulate Milk Products (pCi/kg) (pCi/kg)

(NOTE2) (pCi/1) or Gases (pCi/1) (pCi/kg)

(wet) (wet)

(pCi/m3) (wet)

Gross beta 4 0.01 H-3 2,000 Mn-54 15 . 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30

Nb-95 I-131 Cs-134 Cs-137 15 (NOTE3) 1 15 18 0.07 0.05 0.06 130 150 1

15 18 60 60 80 150 180 Ba-140 60 60 La-140 15 15 NOTE 1: Required detection capabilities for thermo luminescent 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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PAGE 134 OF 156 ATTACHMENT24 (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):

LLD (24-1)

E

V

2.22E+06

Y

e- (lAt)

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 V

2.22E+06

=

the counting efficiency (as counts per disintegration) the sample size (in units of mass or volume) 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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ATTACHMENT 25 (Page 1 of 2)

Detection Capabilities for North Anna Environmental Sample Analysis LOWER LIMIT OF DETECTION (LLD)

Airborne Food Fish Sediment Analysis Water Particulate Milk Products (pCi/kg) (pCi/kg)

(NOTE2) (pCi/1) or Gases (pCi/1) (pCi/kg)

(wet) (wet)

(pCi/m3) (wet)

Gross beta 4 0.01 H-3 2,000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30

Nb-95 1-131 Cs-134 Cs-137 15 (NOTE3) 1 15 18 0.07 0.05 0.06 130 150 1

15 18 60 60 80 150 180 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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PAGE 136 OF 156 ATTACHMENT25 (Page 2 of2)

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 Sb LLD= (25-1)

E

V

2.22E+06

Y

e- (A.At)

Where:

LLD = the "a priori" (before the fact) Lower Limit of Detection as defined above (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)

E V

2.22E+06

= the counting efficiency (as counts per disintegration)

=

the sample size (in units of mass or volume) 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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ATTACHMENT 26 (Page 1 of 1)

Reporting Levels for Radioactivity Concentrations in Environmental Samples at Surry Airborne Water Fish Milk Food Products Analysis Particulate or (pCi/1) (pCi/kg, wet) (pCi/1) (pCi/kg, wet)

Gases (pCi/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 reporting level for the surface (non-drinking) water samples required by Attachment 20 is 20 pCi/1.

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ATTACHMENT 27 (Page 1 of 1)

Reporting Levels for Radioactivity Concentrations in Environmental Samples at North Anna Airborne Water Fish Milk Food Products Analysis Particulate or (pCi/1) (pCi/kg, wet) (pCi/1) (pCi/kg, wet)

Gases (pCi/m3)

(NOTE 1)

H-3 20,000 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 1-131 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 NOTE 1: For drinking water samples

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VIRGINIA

VPAP-2103 POWER REVISI0N7 PAGE 141 OF 156

ATTACHMENT 28 (Page 1 of 8)

Surry Meteorological, Liquid, and Gaseous Pathway Analysis 1.0 METEORLOGICAL ANALYSIS 1.1 Purpose The purpose of the meteorological analysis was to determine the annual average XIQ and D/Q values at critical locations around the Station for ventilation vent (ground level) and process vent (mixed mode) releases. The annual average XIQ and D/Q values were used in a dose pathway analysis to determine both the maximum exposed individual at site boundary and member of the public. The XIQ 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 Onsite meteorological data for the period January 1, 1979, through December 31, 1981, were used in calculations. These data included wind speed, wind direction, and differential temperature 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 (i.e., process vent). The portions of release characterized as ground level were based on AT1s8.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were based on AT1s8.9ft-28.2ft and 158.9 ft wind data.

X/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, "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 XIQ 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 reiease calculations was based on a process vent release height of 131 ft, and plume rise due to momentum for a vent diameter of 3 in. with plume exit velocity of 100 ft/sec.

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PAGE 142 OF 156 ATTACHMENT28 (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 level as specified in Regulat<;>ry 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 Power Company Updated Final Safety Analysis Report Table 1 lA-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. X!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 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 lorig 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 metersN 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 X/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 I 131 , I 133 , 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 m2 at a location 5150 meters S sect~r. 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/m3 for process vent releases at a location 5150 meters S sector. The inhalation pathway is the only other pathway existing at this location. Therefore, the XIQ values given for tritium also apply for the inhalation pathway.

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ATTACHMENT 28 (Page 3 of 8)

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 .

Liquid radioactive effluents from both units are released to the James River via the discharge 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 groin. It has been estimated that boat sport fishing would be performed a maximum of 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months per year, and that bank fishing would be performed a maximum of 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months per year.

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

calculated with LADTAP by the fractional year the individual spent fishing in the canal.

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PAGE 144 OF 156 ATTACHMENT 28 (Page 4 of 8)

Surry Meteorological, Liquid, and Gaseous Pathway Analysis 2.3 Results For the years 1976, 1977, 1979, 1980, and 1981, the invertebrate pathway resulted in the largest 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 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

131 133 organ, of the maximum exposed member of the public, as a result of the release of I ,I ,

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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ATTACHMENT 28 (Page 5 of 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 'X/Q values determined for maximum noble gas doses can be used to determine the maximum dose from I 131 , I 133 , 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/I9n7, was run using 1976, 1977, 1978, 1979, 1980, and 1981 Surry Power Station gaseous effluent release report data. Doses from I 131 , I 133 , tritium, and particulates for the inhalation pathway were calculated using the 6.0E-05 sec/m3 site boundary 'X/Q. Except for the source term data and the 'XIQ value, computer code default p_arameters 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.

The gamma and beta dose factors Kivv, 4vv, Mivv, and Nivv 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 Cifm3 or mrad/yr per Cifm3, 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, Mipv, and Nipv in Attachment 12, using the process vent site boundary X/Q value of 1.0E"-06 sec/m3.

Inhalation pathway dose factors Pivv and Pipv in Attachment 12 were calculated using the equation:

Pi = K (BR) DFA/x/Q)(mrem/yr per Curie/sec) (28-1) where:

K' =a constant of unit conversion, IE+ 12 pCi/Ci BR =the breathing rate of the child age group, 3700 m 3/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

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PAGE 146 OF 156 ATTACHMENT 28 (Page 6 of 8)

Surry Meteorological, Liquid, and Gaseous Pathway Analysis Step 6.3.4.a., requires that the dose to the maximum exposed member of the public from I 131 ,

I 133 , 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, discharge canal bank, and to an 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 values using the NRC computer code GASPAR. * .

It was determined that the member of the public within site boundary would be using the discharge canal bank for fishing a maximum of 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months per year. The maximum annual X/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 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 pathways. Therefore, the maximum exposed member of the public was determined to be the infant, exposed through the grass-cow-milk pathway, critical orgari thyroid, at a location 5150 meters S sector. The only other pathway existing at this location for the infant is inhalation.

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ATTACHMENT 28 (Page 7 of 8)

Surry Meteorological, Liquid, and Gaseous Pathway Analysis The RMivv and RMipv dose factors, except for tritium, in Attachment 18 were calculated by multiplying the appropriate D/Q value with the following equation:

RM.1 - K' QF(U ap ) F * (r) (DFL.

"A,.+'A 1 W m

1

) [fy+

f ff )

p s p

(1 - p s e -A.;

y S

j e-Aitf (28-2) where:

K' = a constant of unit conversion, IE+ 12 pCi/Ci Qp = 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 Ys =

agricultural productivity by unit area of stored feed, 2.0, in Kg!m2 I

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 radioiodine; and 0.2 for particulates DF'Lj = thyroid ingestion dose factor for the ith radionuclide for the infant, in mrem/pCi, from Table E-14, Regulatory Guide 1.109, Rev.I 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 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+o5, in seconds th = transport time from pasture, to harvest, to cow, to milk, to receptor, 7.78E+o6, in seconds fp = fraction of year that cow is on pasture, 0.67 (dimensionless), 7.78E+06 in seconds

=

fs fraction of cow feed that is pasture grass while cow is on pasture, 1.0, dimensionless Parameters used above were obtained fromNUREG-0133 and Regulatory Guide 1.109, Rev.I.

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PAGE 148 OF 156 ATTACHMENT28 (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:

(28-3) where:

K"' = a constant of unit conversion 1E+03 gm/kg H = absolute humidity of the atmosphere, 8.0, gmfm3 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 XIQ = 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 Riivv and Riipv in Attachment 18 were calculated using the following equation:

RI.

1

= K' (BR) DFA.(x/Q) 1 (mrem/yr per Curie/sec)

(28-4)

where:

K' =

a constant of unit conversion, lE+ 12 pCi/Ci BR = breathing rate of the infant age group, 1400 m3/yr, from Table E-5, Regulatory Guide 1.109, Rev.1 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 XIQ = ventilation vent X/Q, 3.0E-07 sec/m3, or the process vent site boundary X/Q, l.3E-07 sec/m3, at a location 5150 meters S sector.

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ATTACHMENT 29 (Page I 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 boundary and member of the public. The XIQ 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 temperature for the purpose of determining joint frequency distributions for those releases characterized as ground level (e.g., ventilation vent), and those characterized as mixed mode (i.e., process vent). The portions of release characterized as ground level were based on AT1s8.9ft-28.2ft and 28.2 foot wind data, and the portions characterized as mixed mode were based on AT1s8.9ft-28.2ft and 158.9 ft wind data.

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 CI a and Cl b) of Regulatory Guide 1.111, "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 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.

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PAGE 150 OF 156 ATTACHMENT29 (Page 2 of 8)

North Anna Meteorological, 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 llC.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.

According to the definition for short term in NUREG-0133, "Preparation of Radiological 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 ..

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 XJQ value that resulted in the maximum total body, skin and inhalation exposure for ventilation vent releases was 9.3E-06 sec/m3 at a site boundary location 1416 meters SE.sector.

For process vent releases, the site boundary X/Q value was 1.2E-06 sec/m3 at a location 1513 meters S sector. The shoreline X/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.

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ATTACHMENT 29 (Page 3 of 8)

North Anna Meteorological, Liquid, and Gaseous Pathway Analysis Pathway analysis indicated that the maximum exposure from 1-131, 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 2.4E-09 per m2 at a location 3250 meters N sector. For process vent releases, the D/Q value was 1.lE-09 per m2 at a location 3250 meters N sector. For tritium, the XIQ value 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.

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 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 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.

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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 the 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 I 131 , 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 I 131 , I 133 , 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.

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 153 OF 156

ATTACHMENT 29 (Page 5 of 8)

North Anna Meteorological, Liquid, and Gaseous Pathway Analysis Step 6.3.1.a.2 dose limits apply specifically to the inhalation pathway. Therefore, the locations and XIQ values determined for maximum noble gas doses can be used to determine the maximum dose from I 131 , I 133 , 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 8/19/77, was run using 1979, 1980 and 1981 North Anna Power Station Gaseous Effluent Release Report data. Doses from I 131 , I 133 , tritium, and particulates for the inhalation pathway were calculated using the 9.3E-06 sec/m3 site boundary X/Q. Except for the source term data 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.

The gamma and beta dose factors Kivv, Livv, Mivv, and Nivv in Attachment 13 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 CiJm3 or mrad/yr per Cifm3, and multiplying by the ventilation vent site boundary XIQ value of 9.3E-06 sec/m3. The same approach was used in calculating the gamma and beta dose factors Kipv, L:ipv, Mipv, and Nipv in Attachment 13 using the process vent site boundary XIQ value of 1.2E-06 sec/m3.

The inhalation pathway dose factors Pivv and Pipv in Attachment 13 were calculated using the following equation:

Pi - K'(BR) DFAi (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. 1

X/Q = the ventilation vent site boundary X/Q, 9.3E-06 sec/m3, or the process vent site .

boundary X/Q, l .2E-06 sec/m3, as appropriate.

VIRGINIA VPAP-2103 POWER REVISION?

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 I 131 ,

I 133 , 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 aetermined that the member of the public within site boundary would be using Lake Anna for recreational purposes a maximum of 2232 hours0.0258 days 0.62 hours 0.00369 weeks 8.49276e-4 months per year. It is assumed that this member of the public would be located the entire 2232 hours0.0258 days 0.62 hours 0.00369 weeks 8.49276e-4 months 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 GASP AR 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 X/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 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 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 Ri dose factors less than those determined for the cow-milk pathway. Although the cow-milk pathway does not exist within five miles of the Station, NUREG-0133 requires the use of cow-milk Ri dose factors since these values result in the most limiting doses. There is no requirement to include the other pathways.

[Commitment 3.2.3]

VIRGINIA VPAP-2103 POWER REVISI0N7 PAGE 155 OF 156 ATTACHMENT 29 (Page 7 of 8)

North Anna Meteorological, Liquid, and Gaseous Pathway*Analysis The RMivv and RMipv dose factors, except for tritium, in Attachment 19 were calculated by multiplying the appropriate D/Q value with the following equation:

(29-2) where:

K' = a constant of unit conversion, IE+ 12 pCi/Ci Qp = cow's consumption rate, 50, in Kg/day (wet weight)

Uap = infant milk consumption rate, 330 liters/yr

Yp Ys Fm r

=

agricultural productivity by unit area of pasture feed grass, 0. 7 Kg!m2 agricultural productivity by unit area of stored feed, 2.0, in Kg!m2 stable element transfer coefficients, from Table E-1, Regulatory Guide 1.109, Rev. 1 fraction of deposited activity retained on cow's feed grass, 1.0 for radioiodine, and 0.2 for particulates DF'Li=: 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 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, 1.73E+o5, in seconds th = transport time from pasture, to harvest, to cow, to milk, to receptor, 7.78E+o6, 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 Guide 1.109, Rev .1.

VIRGINIA VPAP-2103 POWER REVISION?

PAGE 156 OF 156 ATTACHMENT 29 (Page 8 of 8)

North Anna Meteorological, Liquid, and Gaseous*Patbway Analysis Since the concentration of tritium in milk is based on the airborne concentration rather than the deposition, the_ following equation is used:

where:

K"' = a constant of unit conversion 1E+o3 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 XIQ = the annual average concentration at a: location 3250 meters N sector, 7.2E-07 sec/m3

for ventilation vent releases, and 3.9E-07 sec/m3 for the process vent releases Other parameters have been previously defined.

Attachment 4 Page 1 of 1

-* MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS There were no major changes to Surry's Radioactive Liquid, Gaseous or Solid Waste Treatment Systems during this reporting period.

Attachment S Page 1 of 1

INOPERABILITY OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENT MONITORING INSTRUMENTATION There was one effluent monitor, as identified in the Offsite Dose Calculation Manual, that was out of service for more than thirty days in 1995. The monitor is:

1. 02-SW-RM-220, Unit #2 Discharge Tunnel Radiation Monitor. This monitor was removed from service to replace the coaxial and high voltage cables to the monitor per DCP-94-11. The monitor was removed from service on May 17, 1995 and returned to service on June 20, 1995. During this period, action consisted of once per twelve hour grab sampling of the discharge pathway in accordance with the Offsite Dose Calculation Manual surveillance requirements.

Attachment 6 Page 1 of 1

UNPLANNED RELEASES There were no unplanned liquid or gaseous releases during this reporting period .

Attachment 7 Page 1 ofl LOWER LEVEL OF DETECTION FOR EFFLUENT SAMPLE ANALYSIS GASEOUS: Isotope Reqµired LLD Tu!JicalLLD Kr-87 1.00E--04 l.47E 2.22E-07 Kr-88 1.00E-04 2.36E 2.SOE-07 Xe-133 1.00E--04 1.84E 2.68E-07 Xe-l33m 1.00E--04 5.38E 8.21E-07 Xe-135 1.00E--04 6.52E 1.03E-07 Xe-135m 1.00E--04 3.06E 4.31E-07 Xe-138 1.00E-12 5. 8lE 1.03E-06 1-131 1.00E-12 8.llE 1.75E-13 1-133 1.00E-10 7.94E-13 2.13E-13 Sr-89 1.00E-11 1.00E 6.00E-12 Sr-90 1.00E-11 2.00E 3.00E-12 Cs-134 1.00E-11 8.40E l.81E-13 Cs-137 1.00E-11 9.92E 2.lSE-13 Mn-54 1.00E-11 5.65E 1.91E-13 Fe-59 1.00E-11 1.0SE 2.68E-13 Co-58 1.00E-11 3.93E l.54E-13 Co-60 1.00E-11 6.41E 2.32E-13 Zn-65 1.00E-11 1.83E 4.31E-13 Mo-99 1.00E-11 5.39E 1.llE-12 Ce-141 1.00E-11 6.36E 1.33E-13 Ce-144 1.00E-11 2.28E 5.62E-13 Alpha 1.00E-11 1.llE 2.41E-12 Tritium 1.00E-06 5.30E 5.59E-08 LIQUID Sr-89 5.00E-08 2.00E 5.00E-08 Sr-90 5.00E-08 4.00E 1.00E-08 Cs-134 5.00E-07 1.40E 2.45E-08 Cs-137 5.00E-07 1.97E 3.45E-08 1-131 1.00E-06 1.22E 2.30E-08 Co-58 5.00E-07 7 .28E 2.52E-08 Co-60 5.00E-07 1.12E 3.44B-08 Fe-59 5.00E-07 2.30E 4.49E-08 Zn-65 5.00E-07 3.20E 6.61E-08 Mn-54 5.00E-07 1.07E 3.12E-08 Mo-99 5.00E-07 1.0SE . 1.87E-07 Ce-141 5.00E-07 1.SSE 2.83E-08 Ce-144 5.00E-07 5.43E 1.22E-07 Fe-55 1.00E-06 4.00E 9.OOE-07 Alpha 1.00E-07 6.30E 1.36E-08 Tritium 1.00E-05 1.46E l.54E-06 Xe-133 1.00E-05 2.66E 6. 78E-08 Xe-135 1.00E-05 6.83E l.94E-08 Xe-133m 1.00E-05 6.33E 1.56E-07 Xe-135m 1.00E-05 . 3.66E 7.43E-08 Xe-138 1.00E-05 7.53E 1.93E-07 Kr-87 1.00E-05 1.SOE 3.93E-08 Kr-88 1.00E-05 3.lOE 5.54E-08

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