W3P86-1963, Semiannual Radioactive Effluent Release Rept, Jan-June 1986

Semiannual Radioactive Effluent Release Rept, Jan-June 1986

ML20209E971
Person / Time
Site:	Waterford
Issue date:	06/30/1986
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
W3P86-1963, NUDOCS 8609110405
Download: ML20209E971 (36)

Text

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D Semiannual Radioactive Effluent Release Report January 1,1986 - Jtine 30, 1986 Waterford 3 SES Louisiana Power and Light W3500041I 0609 hq, f g-oseg2. oso6ao

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LOUISI POWER ANA

& LlGHT / 317NEW BARONNESTREET

• P. O. BOX 60340 ORLEANS, LOUISIANA 70160 *

(504)595-3100 NuSNNysS$

August 27, 1986 W3P86-1963 A4.05 QA Mr. Robert D. Martin 3 gg{g]3 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 hp-2[%

Arlington, TX 76011

Subject:

Waterford 3 SES Docket No. 50-382 License No. NPF-38 Semiannual Radioactive Effluent Release Report

Dear Mr. Martin:

l Enclosed is the subject report on effluent releases which covers the period of January 1 to June 30, 1986. This report is submitted per Section 6.9.1.8 in the Waterford 3 Technical Specifications (NUREG-ll17) of l

Appendix A to Facility Operating License No. NPF-38 and pursuant to 10 CFR 50.36a(a)(2).

Very truly yours,

_ i .

\

K.W. Cook kg.g*,-

Nuclear Support & Licensing Manager KWC GEW:ssf )

l Enclosure cc (w/ enclosure): NRC, Director, Office of 16E NRC, Document Control Desk, Washington, D.C.

cc (w/o enclosure): G.W. Knighton, NRC-NRR J.ll. Wilson, NRC-NRR NRC Residtnt Inspectors Office B.W. Churchill W.M. Stevennon

\

wAO aam sount opponrumiry emptovena p

l TABLE OF CONTENTS 1.0 SCOPE 2.0 SUPPLEMENTAL INFORMATION 2.1 Regulatory Limits 2.2 Maximum Permissible Concentrations 2.3 Average Energy 2.4 Measurements and Approximations of Total Radioactivity 2.5 Batch Releases 2.6 Unplanned Abnormal Releases l

3.0 GASEOUS EFFLUENTS 4.0 LIQUID EFFLUENTS l

5.0 SOLID WASTES 6.0 METEOROLOGICAL DATA 7.0 ASSESSMENT OF DOSES

8.0 RELATED INFORMATION 8.1 Changes to the Process Control Program j 8.2 Changes to the Offsite Dose Calculation Manual

! 8.3 Unavailability of REMP Milk Sampling I

j 8.4 Report of Technical Specification Required Instrument Inoperability l 8.5 Violations of Radioactive Gaseous Waste Sampling and Analysis Program, Table 4.11-2 9.0 TABLES i

10.0 ATTACHMENTS .

i

W35000 '+H 1

b .

O 1.0 SCOPE This Semiannual Radioactive Release Report is submitted as required by Louisiana Power and Light's Waterford 3 Technical Specification 6.9.1.8.

It covers the period from January 1, 1986 through June 30, 1986.

Included in this report is a summary of the quantities of radioactive liquid and gaseous effluents and solid wastes released from the plant during the reporting period. The summary of meteorological data and results from the assessment of radiation doses due to the release of liquid and gaseous radioactive effluents will be included in the Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1, 1987.

Information in this report is presented in the format outlined in Appendix B of Regulatory Guide 1.21.

Other required information in this report includes: (1) explanation of why certain instrumentation was not restored to operable status within the time specified in the ACTION Statement, as per Waterford 3 SES Tech-nical Specification 3/4.3.3.11; (2) a report of deviation from Technical Specification Table 4.11-2; and (3) a report of unplanned abnormal releases. .

2.0 SUPPLEMENTAL INFORMATION 2.1 Regulatory Limits Specified as follows are the technical specification limits applicable j to the release of radioactive material in liquid and gaseous effluents.

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2.1.1 Fission and Activation Gases (Noble Gases)

The dose rate due to radioactive noble gases released in gaseous effluents from the site to areas at and beyond the site boundary snall be limited to less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.

The air dose due to noble gases released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

2.1.2 lodines; Particulates, Half Lives > 8 Days; and Tritium The dose rate due to Iodine-131 and 133, tritium, and all radionuclides in particulate form with half lives greater than eight (8) days, released in gaseous effluents from the site to areas at and beyond the site boundary, shall be

, limited to less than or equal to 1500 mrem /yr to any organ.

t

. The dose to a member of the public from Iodine 131 and 133,

! tritium, and all radionuclides in particulate form with half lives greater than eight (8) days in gaseous effluents l

! released to areas at and beyond the site boundary shall be limited to the following:

{

I

! a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, W350004H 3 i

b. During any calendar year: Less than or equal to 15 mrem to any organ.

2.1.3 Liquid Effluents The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in 10CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-4 pCi/ml total activity.

The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.

2.1.4 Uranium Fuel Cycle Sources The dose or dose commitment to any member of the public due to releases of radioactivity and radiation from uranium fuct cycle sources shall be limited to Icss than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.

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I

. 1 2.2 Maximum Permissible Concentrations 2.2.1 Fission and Activation Gases; Iodines; and Particulates, Half Lives > 8 Days For gaseous effluents, maximum permissible concentrations are not directly used in release rate calculations since the applicable limits are expressed in terms of dose rate at the site boundary.

2.2.2 Liquid Effluents The maximum permissible concentration (MPC) values specified in 10CFR20, Appendix B, Table II, Column 2 are used as the permissible concentrations of liquid radioactive effluents at the unreitricted area boundary. A value of 2.0E-4 pC1/mi is used as the MPC for dissolved and entrained noble gases in liquid effluents.

2.3 Averase Energy This is not applicable to Waterford 3 SES's radiological effluent technical specifications.

2.4 Measurements and Approximations of Total Radioactivity The quantification of radioactivity in liquid and gaseous effluents was accomplished by performing the sampling and radiological analysis of ef fluents in accordance with the requirements of Tables 4.11-1 and 4.11-2 of the amended Waterford 3 SES Plant Technical Specifications (see Attachments 1 and 2).

W350004H 5

i' 2.4.1 Fission and Activation Gases (Noble Gases)

For continuous releases, a gas grab sample was analyzed monthly for noble gases. Each week a Gas Ratio (GR) was calculated according to the following equation:

GR = Average Weekly Noble Gas Monitor Reading Monitor Reading During Noble Gas Sampling The monthly sample analysis and weekly Gas Ratio were then used to determine noble gases discharged continuously for the previous week. For gas decay tank and containment purge batch releases, a gas grab sample was analyzed prior to release to determine noble gas concentrations in the batch.

In all cases the measured total radioactivity in gaseous effluents was determined from measured concentrations of each radionuclide present and the total volume discharged.

2.4.2 Iodines and Particulates Iodines and particulates discharged were sampled using a continuous sampler. Each week the charcoal cartridge and particulate filter were analyzed for gamma emitters using gamma spectroscopy. The determined radionuclide concentrations and effluent volume discharged were used to calculate the previous week's activity released.

Composites of the particulate samples were analyzed quarterly for Sr-89 and Sr-90 by a contract laboratory (Teledyne Isotopes). Particulate gross alpha was measured weekly using gas flow proportional counting techniques. The determined

! activities were used to estimate effluent concentrations in

subsequent releases until the next scheduled analysis was

! pe r fo rmed.

i W350004H 6

~- --- -

T

I Grab tritium samples for continuous and batch releases were analyzed monthly. The determined concentrations were used to estimate tritium activity in subsequent releases until the next scheduled analysis was performed.

2.4.3 Liquid Effluents For continuous releases, samples were collected weekly and analyzed using gamma spectroscopy. The measured concentra-tions were used to determine radionuclide concentrations in the previous week's releases. For batch releases, gamma analysis was performed on the sample prior to release.

For both continuous and batch releases, Sr-89, Sr-90, and Fe-55 composite samples were analyzed quarterly by a contract laboratory (Teledyne Isotopes). Tritium and gross alpha composite samples were analyzed monthly using liquid scintil-lation and gas flow proportional techniques respectively.

For radionuclides analyzed in composite samples, the measured concentrations of the previous composite were used in deter-mining concentrations in liquid effluents.

The measured total radioactivity in liquid effluent releases was determined from the measured concentrations of each radio-nuclide present and the total volume of the effluent discharged.

2.5 Batch Releases The summarization of information for gaseous and liquid batch releases is included in Table 1.

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r 2.6 Unplanned Abnormal Releases 2.6.1 Unplanned Abnormal Release February 12 & 13, 1986 Description of Event On February 12 and 13,1986, an unplanned release was made via the plant stack; however, the Technical Specification 3.11.2.1 instantaneous dose rate limit of 500 mrem /yr was not exceeded. Based upon the plant stack monitor reading and grab sample results, the maximum instantaneous dose rate was determined to be 148 mrem per year or 30% of the Tech Spec limit.

Cause of Event A major modification had been underway on the Gas Analyzer System (Waste Gas Holdup System Hydrogen Monitor and Oxygen Monitors) since 1985 (see section 8.4.1 of this report). As of Wednesday, February 12, there were two sample points ready for testing, the Volume Control Tank (VCT) and Gas Surge Tank (GST). Maintenance personnel lined up the system to sequen-tially sample the two points and, shortly thereafter, left it to run overnight. The Plant Stack Radiation Monitor alarmed at 1425, however it was not until approximately 1715 that the release was traced to the analyzer and isolated. The release was believed to have been caused by a value lineup problem which directed the analyzer discharge to the Plant Stack (via the Vent Gas Collection Header) instead of the Gas Surge Header.

Maintenance personnel returned on February 13 to troubleshoot the system and were only able to identify one problem. After eliminating this problem, the analyzer was turned on. Upon reaching the GST sample point, a pressure gauge suddenly W350004H 8

9 indicated in excess of 60 psig. Maintenance personnel quickly secured the analyzer, however some radioactive gas was released to the sample room which resulted in a Plant Stack Radiation Monitor alarm for approximately 10 minutes.

Detailed work instructions were generated to systematically test the system. It was discovered that the GST sample line was crossed with that of Gas Decay Tank (GDT) C. Thus, when the analyzer switched to sample the GST, the pressure of GDT C (104 psig) ruptured the sample pump diaphram, thereby releasing radioactive gas to the sample room where it was discharged to the plant stack by the room ventilation.

Actions Taken to Prevent Recurrence The root cause of the release was the crossing of the sample lines. This error had existed since the original construction of the system. Since this switch was subsequently corrected, the specific problem will not recur.

Ilowever, future testing of the Gas Analyzer System will be performed under detailed test procedures and close Health Physics surveillance of work activities.

Radiological Consequences of the Release The releases occurred on February 12 and 13, 1986. While the first release did reach 30% of the Tech Spec limit, the cumulative gamma and beta doses were trivial. Details are as follows:

1) Tech Spec 3.11.2.1 - Maximum instantaneous release rate j 2/12/86 148 mrem /yr. 30% of Tech Spec limit 2/13/86 11 mrem /yr. 2% of Tech Spec limit W35000411 9

ii) Tech Spec 3.11.2.2 - Quarterly Air Doses 2/12/86 Gamma 4.5E-4 mrad 0.009% of quarterly limit Beta 6.6E-4 mrad 0.007% of quarterly limit 2/13/86 Gamma 4.5E-6 mrad 9E-5% of quarterly limit Beta 6.6E-6 mrad 7E-5% of quarterly limit NOTE: Air doses for the first quarter of 1986 were as follows:

Gamma 0.496 mrad 9.9% of quarterly limit Beta 1.29 mrad 12.9% of quarterly limit These dose rates and doses have been calculated in accordance with the methodology and parameters in the Offsite Dose Calculation Manual. As noted above, the contribution of the two inadvertent releases is negligible with respect to quarterly doses. Instantaneous releases rates were well under Tech Spec limits.

2.6.2 Unplanned Abnormal Release April 5 & 10, 1986 Description of Event On April 5, 1986 at 1648, an unplanned release was made via the plant stack that exceeded Technical Specification 3.11.2.1 instantaneous dose rate limit of 500 mrem /yr to the total body. Based on the plant stack monitor reading and grab sample results, the maximum instantaneous dose rate was determined to be 657 mrem per year. Release duration was determined to be less than 60 minutes. On April 10, 1986, the plant stack radiation monitor again indicated higher than W35000411 10

F normal levels. Based on the plant stack monitor reading and grab sample results, the maximum instantaneous dose rate was determined to be 149 mrem per year. Release duration was determined to be less than 20 minutes.

Cause of Event The normal vent pathway for the VCT is to the Gas Decay Tanks via the Gas Surge Header and Waste Gas Compressors. Station Modification (SM) 461, Revision 3, worked during the March Outage, added piping and a valve to allow venting the Boric Acid Concentrator to the Vent Gas Collection Header (VGCH).

This provided a second vent path in addition to the existing vent to the Gas Surge Header. There are valves on each of these pathways to isolate the discharge; BM-252A on the VGCH and GWM-121A on the Gas Surge Header side. After the April 10 release, a system walkdown revealed that both of these valves were open.

Procedures governing station modifications did not require changes to plant systems to be red lined on the control room drawings until the station modification paperwork had been closed out. Since all work under SM461 had not been completed the control room drawings were not updated, and Operations personnel were not aware of the additional vent. Since both GWM-121A and BM-252A valves were open, this allowed any gas vented to the Gas Surge Header to inadvertently be discharged out the plant stack via the VGCH.

ACTIONS TAKEN TO PREVENT RECURRENCE As soon as it was evident that the instantaneous dose rate limit may have been exceeded, venting of the VCT was secured.

W350004H 11

T O

k After investigation of the incident revealed the release flow path, valves GWM-121A and 121B were danger tagged closed.

The applicable operating procedure was revised to close these valves.

Procedures governing station modifications have been changed so that control room drawings must be red lined prior to the r Shift Supervisor's release of the clearance on the affected

, system.

RADIOLOGICAL CONSEQUENCES OF THE RELEASE Although the instantaneous gaseous limit was exceeded on April 5, due to the short release duration the weekly cumula-tive noble gas beta.and gamma doses for April 4-11, 1986 were not significantly larger than the subsequent week. Details' are as follows:

i) Tech Spec 3.11.2.1 - Maximum instantaneous release rate 4/5/86 657 mrem /yr 130% of Tech Spec limit 4/10/86 149 mrem /yr . 30% of Tech Spec limit i

1 ii) Tech Spec 3.11.2,2 - Quarterly Air Doses 4/5/86 Gamma 3.13C-3 brad 0.06% of Tech Spec limit

( Beta 3.26E-3 mrad 0.03% of Tech Spec limit 4/10/86 Gamma 2.12E-3 mrad 0.04% of Tech Spec limit Beta 2.22E-3 mrad 0.02% of Tech Spec limit i

W350004H 12

r NOTE: Air doses for the second quarter of 1986 were as follows:

Gamma 0.0819 mrad 1.6% of quarterly limit Beta 0.189 mrad 1.9% of quarterly limit These dose rates and doses have been calculated in accordance with the methodology and parameters in the Offsite Dose Calculation Manual. As noted above, the contribution of the two inadvertent releases is negligible with respect to quarterly doses. Instantaneous release rates were well under Tech Spec limits.

3.0 GASEOUS EFFLUENTS The quantities of radioactive material released in gaseous effluents are summarized in Tables 1A, 1B, and IC. Note that there are no elevated releases at Waterford 3 SES.

4.0 LIQUID EFFLUENTS The quantities of radioactive material released in liquid effluents are summarized in Tables 2A and 2B.

5.0 SOLID WASTES The summary of radioactive solid wastes shipped offsite for disposal is listed Table 3.

6.0 METEOROLOGICAL DATA The summary of the hourly meteorological data for this reporting period will be included in the Semiannual Effluent Release Report to be submitted within 60 days after January 1, 1987.

(

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F 7.0 ASSESSMENT OF DOSES The summary of doses from gaseous and liquid effluents for this reporting period will be included in the Semiannual Effluent Release Report to be submitted within 60 days after January 1, 1987.

8.0 RELATED INFORMATION 8.1 Changes to the Process Control Program There were no changes to the Process Control Program for the period this report covers.

8.2 Changes to the Offsite Dose Calculation Manual There were no changes to the Offsite Dose Calculation Manual for the period this report covers.

8.3 Unavailability of REMP Milk Samples Due to the unavailability of three milk sampling locations within five kilometers of the plant, Broad Leaf sampling is performed in accordance with Technical Specification Table 3.12-1. Milk is collected, when available, from the control location and three identi-fied sampling locations as indicated in ODCM, Table 2 and Table 3.

8.4 Report of Technical Specification Required Instrument Inoperability Technical Specification, Limiting Condition for Operation (LCO),

3.3.3.11 requires the reporting in the Semiannual Effluent Release Report of why designated inoperable instrumentation was not restored to operability within the time specified in the ACTION Statement.

W350004H 14

7

8.4.1 Monitor

Waste Gas Holdup System Hydrogen Monitor and Oxygen Monitors Period of Inoperability: 3/21/85 - 6/30/86 (At end of reporting period monitors were still inoperable)

Time Required by Technical Specifications to Restore Operability: 14 days Cause of Inoperability:

Due to design problems excess amounts of moisture were allowed to leak into both the Beckman 0 and Delphi H and 0 2 2 2 analyzer systems.

Reason Inoperability Not Restored Within Allotted Time:

Extensive hours were spent attempting to restore these analyzers to operable status. Several analyzer cells were replaced, the so: Toid and regulator were repaired, and the sample pump was both repaired and replaced. After these efforts failed to return the monitor to service, a station modification was initiated to replace the analyzer cells with less moisture sensitive models and to completely redesign the sample line condensate drain system.

This modification entailed work in several areas of the plant and on four different systems. All existing piping and electronics associated with the Waste Gas Holdup Systen Hydrogen and Oxygen Monitors was essentially scrapped and redesigned.

W350004H 15

r Work included re-routing all sample lines in the Laundry Room, modifying the existing drain header, and fabrication of a new drain header to tie into the Vent Gas Collection Header.

Re-routing of the Gas Surge Header Sample line and fabrication of its drain was performed in Safeguards Room B. On Gas Decay Tank A a new separator and drain line on the Waste Gas Collec-tion Discharge Header was added to the second low point.

Actual work on the Gas Analyzer Panel consisted of (1) adding 12 new solenoid valves in the sample inlets; (2) modifying the panel to accommodate the new exo-sensor units; (3) installing a new pump and its associated tubing; and (4) wiring of all new and relocated components.

While testing the system, the Gas Decay Tank "C" sample line was found to be crossed with the Gas Surge Tank sample line.

Due to greater pressure in the Gas Decay Tank than that of the Gas Surge Tank the sampling pump diaphragm was blown.

This event also identified other problems with the system which required correction.

Design changes have been made to uncross the lines and install a pressure switch to prevent any subsequent over-pressurization. The pump was replaced and new wiring was installed. Additional moisture traps, pressure regulators, valves and pressure indicators have been installed. A detailed test procedure is currently being prepared. Pending the successful completion of the test, the monitors should be operable by the end of 1986.

8.4.2 Monitor

Radiation Monitor PRM-IRE-0647 (Liquid Waste Management System)

Period of Inoperability: 6/2/86 - 6/16/86 W350004H 16

l o

I Time Required by Technical Specifications to Restore Operability: 14 days Cause of Inoperability:

This monitor was declared out-of-service when its reading did not agree with grab sample results. A defective check source drive mechanism caused the check source to remain in the

" check" position. After repair of the check source drive mechanism, the monitor pre-amp board was found to be defec-tive. A new pre-amp board was installed, however, during power-up of the monitor a short circuit was detected in the new pre-amp board. This short circuit caused the amplifier and high voltage to short-out. Repairs had to be made and a functional test was then initiated. At that time it was discovered that due to a modification of the annunciator panel in the control room, a procedure change was necessary prior to completion of the functional test. This monitor was finally declared operable 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> and 11 minutes beyond the 14 day period.

Reason Inoperability Not Restored Within Allotted Time:

Due to extensive troubleshooting, component malfunction and a procedure change to complete the functional test, the monitor could not be restored to service within 14 days.

8.5 Violations of Radioactive Gaseous Waste Sampling and Analysis Program1 Table 4.11-2 8.5.1 Gaseous Release Type: Plant Stack Sampling Frequency: Continuous Minimum Analysis Frequency: Quarterly Composite, Particulate Sample W350004H 17

Type of Activity Analysis: Sr-89, Sr-90 Description of Violation:

Quarterly composites of the air particulate filters are analyzed for Sr-89, Sr-90 by a contract laboratory (Teledyne Isotopes). The 4th Quarter 1985 Plant Stack composite was lost in shipment and therefore could not be analyzed as required by Technical Specification Table 4.11-2.

Upon notification that the sample had been lost, a particulate sample was obtained from the Plant Stack Wide Range Gas Accident Monitor for the time period covered by the lost samples. An analysis for Sr-89, Sr-90 indicated no detectable activity present. Air particulate filters from both plant stack monitors are now being collected. One set of air partic-ulate filters are sent to the contract laboratory while the other set is stored onsite in case of loss during shipment.

9.0 TABLES 1 Batch Release Summary 1A Semiannual Summation of all Releases by Quarter - All Airborne Effluents IB Semiannual Airborne Continuous Elevated and Ground Level Releases IC Semiannual Airborne Batch Elevated and Ground Level Releases 2A Semiannual Summation of All Releases by Quarter - All Liquid Effluents 2B Semiannual Liquid Continuous and Batch Releases 3 Solid Waste Shipped Offsite for Disposal i

10.0 ATTACHMENTS

1. Technical Specification Table 4.11-1

2. Technical Specification Table 4.11-2 W350004H 18

g 4 TABLE 1A (1 of 1)

. REPORT CATEGm Y  : SEMTAlet#L SLM1AT!m 7 ALL PILEASES SY QUMTER TYFE OF ACT1VITY  : ALL AIRBE.'E EFFLUENTS FIPORTING PERIOD  : M TER 8 1 AND GUARTER 4 2

UNIT : QUARTER 1 :QUATCER 2 :EST. TOTAL:

: HOURS :HCURS :EFIG % :

TYPE OF EFFLUENT  :  : 1-2160 :2161-4344 :  :

A. FISSION AND ACTIVATION PRONCTS

1. TOTAL RELEASE :CGIES  : 3.52E 03 : 5.37E 02 : 1.5CE 01:

2. AWRAGE RELEASE RATE FCR PERIOD :UCI/SEC : 2.2 % 02 : 3.44E 01 :

3. FERCENT CF AFPLICABLE LIMIT  :  %  : N/A  : N/A  :

3. Pf.DI0IODDE; UOTAL !0CIPE-131  : CURIES  : 3.33E-04 : 2.?!E-03 : 1.5CE 01:

2. AVERAGE RELEASE RATE FOR PERIOD :UCI/SEC : 2.45E-05 : 1.86E-04 :

3. PERCENT OF APP ICABLE LIMIT  :  %  : N/A  : N/A  :

C. FARTICULATES

1. FARTICULATES(HALF-LPES)8 DAYS) :CWIES  : 0.00E-01 : 0.00E-01 : 1.50E 01:

2. AVERAGE RELEASE RATE FCR PERIOD :UCI/SEC : 0.00E-01 : 0.00E-01 :

3. PERCENT OF AFPLICABLE LIMIT  :  %  : N/A  : N/A  :

4. GROSS ALPHA RADI0 ACTIVITY  : CURIES  : 2.32E-06 : 2.11E-06 :

D. TRITIUM

1. TOTAL RELEASE  : CURIES  : 2.21E 03 : 4.73E 03 : 1.50E 01:

2. AVERAGE RELEASE RATE FOR PERIOD :UCI/SEC : 1.41E 02 : 3.02E 02 :

3. PERCENT & APPLICABLE LIMIT  : I  : N/A  : N/A  :

19

p-TABLE IB (1 of 1)

- REPORT CATEGORY  : SEMTAldCAL A!RBORIE C@ TIM 0JS ELEVATED AND GR0lf0

LE'EL PELEASES. TOTALS FOR EACH NUCLIE RELEASED.

TYPE (F ACTIVITY : FISSI@ GASES. ICDIES. AND FARTICtLATES REPORTING PERIOD : QlARTER # 1 AND QUARTER I 2

ELEVATED RELEASES : GROLAD RELEASES :

LNIT  : QUARTER 1 :Cl#TER 2 :L'JiTER 1 :QUARTERU

: HOWS  : HOURS  : HOURS  : HOURS  :

MILICE  :  : 1-2160 :2161-4344 : 1-2160 :2161-4344 :

F:SSICN CASES G-SSM  : CWIES : 0.00E-01 : 0.0CE-01 : 1.11E 01 : 1.92E 00 :

tR-SS .  : CURIES : 0.00E-01 : 0.00E-01 : 1.51E 00 : 1.S% 00 :

XE-133M  : CURIES : 0.0CE-01 : 0.00E-01 : 1.77E 00 : 0.00E-01 :

XE-133  : CLRIES : 0.00E-01 : 0.0CE-01 : 3.21E 03 : 4.24E 02 :

XE-135  : CURIES : 0.00E-01 : 0.0CE-01 : 1.43E 02 : 2.7SE 01 :

TOTAL FCR PERIOD : CURIES : 0.0CE-01 : 0.0CE-01 : 3.37E 03 : 4.'R 02 :

ICDIES I-131  : CURIES : 0.00E-01 : 0.00E-01 : 3.S3E-04 : 2.91E-03 :

1-133  : CLRIES : 0.0CE-01 : 0.00E-01 : 4.81E-06 : 1.43E-04 :

TOTAL FOR PERICD : CLRIES : 0.00E-01 : 0.00E-01 : 3.SSE-04 : 3.0E-03 :

PARTICtLATES H-3  : CLRIES : 0.00E-01 : 0.00E-01 : 2.21E 03 : 4.73E 03 :

TC-99M  : CLRIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 4.52E-07 :

G ALPHA ,

CURIES : 0.00E-01 : 0.00E-01 : 2.32E-06 : 2.11E-06 :

TOTAL FOR PERIOD : CLRIES : 0.00E-01 : 0.00E-01 : 2.21E 03 : 4.73E 03 :

20

, _ _ ,n t ..

TABLE 1C (1 of 1)

REP WT CATEG M Y t SEMIAMtJAL AIRB0BE BATCH ELEVATED AND GROLND

LEVEL RELEASES. TOTALS F M EACH NUCLIDE RELEASED.

TYPE & ACTIVITY  : FISSION GASES. ICDIES. AND FARTICULATES REPORTING PERIOD  : QUARTER 8 1 AND QUARTER 8 2

ELEVATED RELEASES : GROUND RELEASES :

UNIT :0UARTER 1 :0UARTER 2 : QUARTER 1 : QUARTER 2 :

:HWRS  : HOURS  : HOURS  : HOURS  :

?!JCL!DE  :  : 1-2160 :2161-4344 : 1-2160 :2161-4344 :

FISS!CN CASES KR-85M  : CLRIES : 0.00E-01 : 0.00E-01 : 3.29E-02 : 3.13E-02 :

KR-38  : CURIES : 0.0CE-01 : 0.0CE-01 : 3.89E-02 : 2.90E-02 :

XE-131M  : CURIES : 0.0CE-01 : 0.00E-01 : 1.77E 00 : 6.53E-01 :

XE-132M  : CURIES : 0.00E-01 : 0.0CE-01 : 1.36E 00 : 6.31E-01 :

JE-133  : CURIES : 0.00E-01 : 0.00E-01 : 1.53E 02 : 7.96E 01 :

IE-129.  : CUR!ES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 6.01E-03 :

IE-135  : CTIES : 0.00E-01 : 0.00E-01 : 3.70E-01 : 4.46E-01 :

AR-l' - - " *'re ^ ^^c ^' a ^^c-01 : 1.4?E-01 : 1.65E-02 :

TOTAL FOR PERIOD  : C3!ES : 0.00E-01 : 0.00E-01 : 1.57E 02 : 8.14E 01 :

103IES NONE PARTICULATES H-3  : CURIES : 0.00E-01 : 0.00E-01 : 1.35E 00 : 1.29E CC :

e 21

.g TABLE 2A (L of 1)

REFORT CATEGORY  : SEMIAPNJAL SLMIAT!CN OF ALL RELEASES SY MTER TYFE OF ACTIVITY  : ALL LIQUID EFFLI NTS REPmTING FERIOD  : QUARTER 4 1 AND GUARTER 4 2

UNIT :00ARTER 1 :00ARTER 2 :EST. TOTAL:

:HWRS :HOLSS :EFSCR % :

TYFE CF EFFLLENT  :  : 1-2160 :2161-4344 :  :

A. FISSICN AhD ACTIVATION PRODUCTS

1. TCTAL RE:. EASE (h0T INCLL" DING  :  :  :  :  :

TRITIt?'. GASES. ALFHA)  : CURIES  : 9.81E-01 : 4.*R -01 : 1.50E 01:

2. A'.EFACE DILUTED CCNCENTRATION  :  :  :  :

DURING PER!CD :lC!/ML  : 2.06E-08 : 4.33E-09 :

3. FERCENT CF APPLICABLE LIMIT .:  %  : N/A  : N/A  :

B. TRITILM

1. TCTAL RELEASE  : CURIES  : 7.63E 01 : 1.10E 02 : 1.*4 01:

2. A'.UAGE DILUTED COCENTMIION  :  : .  :  :

'tRING FERICD :UCI/"L  : 1.60E-06 : 1.50E-06 :

3. PERCENT OF APPLICABLE LIMIT  :  %  : N/A  : N/A  :

C. DISSOLVED AND ENTRA!ED GASES

1. TOTAL RELEASE :CGIES  : 3.32E 00 : 7.67E 00 : 1.50E 01:

2. AVERAGE DILUTED CONCENTRATION  :  :  :  :

CLRING FEIOD :UCI/ft.  : 1.75E-07 : 1.07E-07 : .

3. FERCENT CF APPLICABLE LIMIT  :  %  : N/A  : N/A  :

D. GROSS ALPHA RADI0 ACTIVITY

1. TOTAL RELEASE :Cm!ES  : 3.70E-05 : 3.97E-06 : 1.50E 01:

E. WASTE V 1 RELEASED (PRE-DILUTION) :0AL  : 7.36E 05 : 1.07E 06 : 1.50E 01:

F. VOL! N OF DILUTION WATER USED :0AL  : 1.26E 10 : 1.90E 10 : 1.50E 01:

22

- 1

-~ ~.' _

TABLE 2B (1 of 2)

PEP m T CATEGORY  : EMIAMCAL LIQUID CONTIWOUS AW BATCH RELEASES

TOTALS FOR EACH WCLIDE RELEASED.

TYPE CF ACTIVITY  : ALL RADI!huCLIIS REP % TING PERIOD  : QUARTER S 1 AND QUARTER 8 2

CCNTINUQUS RELEASES : BATCH RELEASES :

UNIT :0UARTER 1 : QUARTER 2 :00ARTER 1 :QUNiiER 2 :

: HOWS :HWRS :H0GS lH0$S  :

NUCLIDE  :  : 1-2160 :2161-4344 : 1-2160 :2161-4344 :

ALL NUCLIDE3 H-3  : CWIES : 0.00E-01 : 0.00E-01 : 7.62 01 : 1.10E 02 :

NA-24  : CURIES : 0.00E-01 : 0.00E-01 : 1.32E-03 : 5.29E-03 :

CR-51  : CWIES : 0.00E-01 : 0.00E-01 : 6.50E-03 : 1.64E-02 :

MN-54  : CWIES : 0.00E-01 : 0.00E41 : 1.41E-03 : 3.87E-03 :

FE-55  : CmIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 2.21E-03 :

FE-59  : CURIES : 0.00E-01 : 0.00E-01 : 2.99E-03 : 1.41E-02 :

CD-53  : CmIES : 0.00E-01 : 0.00E-01 : 4.43E-01 : 2.34E-01 :

C0-60

CWIES : 0.00E-01 : 0.00E-01 : 7.33E-03 : 1.0GE-02 :

IN-65  : CGIES : 0.00E-01 : 0.00E-01 : 8.49E-06 : 8.03E-05 :

RB-83  : CWIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 5.38E-03 :

SR-91  : CWIES : 0.00E-01 : 0.00E-01 : 3.78E-04 : 0.00E-01 :

SR-92  : CWIES : 0.00E-01 : 0.00E-01 : 1.18E-04 : 2.99E-04 :

ZR-95  : CWIES : 0.00E-01 : 0.00E-01 : 4.12E-04 : 3.05E-03 :

NB-95  : CURIES : 0.00E-01 : 0.00E-01 : 1.27E-03 : 5.00E-03 :

H0-99  : CURIES : 0.00E-01 : 0.00E-01 : 6.55E-03 : 1.02E-04 :

TC-99M  : CURIES : 0.00E-01 : 0.00E-01 : 9.56E-03 : 4.74E-04 :

RU-103  : CURIES : 0.00E-01 : 0.00E-01 : 2.89E-04 : 4.48E-04 :

AG-110M  : CWIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 8.51E-04 :

I-131  : CURIES : 0.00E-01 : 0.00E-01 : 2.54E-01 : 7.52E-02 :

I-132  : CWIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 2.30E-04 :

!-133  : C m !ES : 0.00E-01 : 0.00E-01 : 1.SSE-01 : 2.65E-02 :

1-135 *

C3!ES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 5.SSE-03 :

CS-134  : CRIES : 0.00E-01 : 0.00E-01 : 3.93E-04 : 6.29E-03 : -

CS-1" CTIES : 0.00E-01 : 0.00E-01 : 6.2SE-04 : 6.90E-04 :

CS-137  : CURIES : 0.00E-01 : 0.00E-01 : 1.14E-03 : 2.23E-02 :

PA-140  : CURIES : 0.00E-01 : 0.00E-01 : 4.06E-04 5.96E-05 :

LA-140  : CWIES : 0.00E-01 : 0.00E-01 : 4.68E-03 : 1.00E-03 :

CE-141  : CWIES : 0.00E-01 : 0.00E-01 : 1.56E4 5 : 1.18E-05 :

23

p-

~

TABLE 2B (2 of 2)

REP W T CATE00RY t SEMIAW4JAL LIQUID CCNTI110US GiD BATCH RELEASES

TOTALS FOR EACH NUCLICE FILEASED.

TYPE OF ACTIVITY  : ALL RADIOMJCLIES REPORTING FERICD  : QUARTER 8 1 AND M TER I 2

CONT!NUOUS RELEASES : BATCH RELEASES :

UNIT  : QUARTER 1:MTER5:M TER 1 :0LG TER 2 :

:H0tRS  : HOURS :HIRS :h0.RS  :

HUCLIDE  :  : 1-2160 :2161-4344 : 1-2160 :2161-4344 :

ALL NUCLIDES CONTIMJED CE-144  : CURIES : 0.00E-01 : 0.00E-01 : 1.06E-03 : 7.81E-04 :

W-187  : CTIES : 0.00E-01 : 0.00E-01 : 6.99E-04 : 2.69E-03 :

KR-SS1  : CLEIES : 0.00E-01 : 0.00E-01 : 2.12E-04 : 1.95E-03 :

KR-SS  : CLRIES : 0.00E-01 : 0.0M-01 : 0.00E-01 : 9.62E-04 :

XE-131M  : CURIES : 0.0CE-01 : 0.00E-01 : 8.46E-02 : 1.05E-01 :

IE-1331  : CURIES : 0.0tE-01 : 0.00E-01 : 5.90E-02 : 5.53E-02 :

XE-133 '

CURIES : 0.00E-01 : 0.00E-01 : 8.16E 00 : 7.38E 00 :

XE-135M  : CURIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 6.3?E-02 :

XE-135  : CURIES : 0.00E-01 : 0.00E-01 : 1.6CE-02 : 6.72E-02 :

G ALPHA  : CURIES : 0.0CE-01 : 0.00E-01 : 3.90E-05 : 3.97E-06 :

C0-57  : CURIES : 0.00E-01 : 0.00E-01 : 2.90E-04 : 2.94E-04 :

53-124  : CLRIES : 0.00E-01 : 0.00E-01 : 4.57E-02 : 8.97E-03 :

SN-113  : CURIES : 0.00E-01 : 0.00E-01 : 8.02E-04 : 1.12E-03 :

NS-97  : CURIES : 0.00E-01 : 0.00E-01 : 0.00E-01 : 2.92E-04 :

SB-122  : CURIES : 0.00E-01 : 0.00E-01 : 2.39E-03 : 1.89E-04 :

TOTAL F M FERIOD  : CURIES : 0.00E-01 : 0.00E-01 : 8.56E 01 : 1.18E 02 :

e 24

• • • REELA.ATCptY SJtDE 1.21 REPORT ++e SCLIS WASTE SNIPPED OFFSITE FOR DIM

.' ** DURINS PERICO FROM 1 186 70 63086 ee e

TABLE 3 (1 of 1)

I.

WASTE TYPE CUBIC METERS CURIES  % ERROR (CI)

E3-9 CM N 10.3 1.8E-03 +/- 25%

SR-D-MA DRMIN 4.8 1.!E-01 +/- 25%

SR-S-CM DEMIN 20.6 3.8E+00 +/- 05%

SR-D-NA DEMIN 15.5 1.6E-01 +/- 25%

SR-D-NA RCS PUR 5.2 * . !E-01 +/- 25%

• • EST! MATES OF MAJOR NUCLIDES BY WASTE TYPE **

WASTE TYPE NUCLIDE ABUNDANCE CURIES EB-S-CM BOTTOMS Co-58 59.% 1.1E-03 Co-60 2.% 3.5E-05 Nt-59 0.% 1.5E-10 N1-63 0. % 0.1E-06 H-3 0.X 4.BE-06 C-14 0.% 0.1E-07 Sr-90 0.% 3.5E-06 Tc-99 0.% 1.7E-06 I-129 0.% 2.9E-06 Cs-137 6.% 1.0E-04 Co-144 12.% 2. E-04 Pu-241 0. 7. 4.3E-05 Cm-24 0.% 1.7E-07 SR-D-NA DEMIN Fe-55 25.% 0. E -02 Co-58 17.% 0.!E-02 Co-eo 17.% 0.:E-02 N1-63 19.% 2. 6E -02 H-3 0. 7. 0.5E- 04

• C-14 0.7 8.1E-06 Tc-99 0.% 4.7E-06 I-129 0.% 7.0E-06 Cs-137 5.% 7.:E-00 Pu-241 0.7 0.:E-05 .

Cm-24: 0.% 1.5E-07 SR-S-CM DEMIN Fe-55 25.% 9.4E-01 Co-58 17.% 6.6E-01 Co-60 17.% 6.4E-o1 N1-63 19.7. 7.4E-01 H-3 0. 7. 1.0E-a:

C-14 0. 7. 2.4E-v4 Tc -99 0.% 1.!E-94 I-129 0.% 0.0E-94 Cs-107 5.% 0.1E-91 e u -241 0.% 6. 4 E- :'4 Cm-24: 0.% 4.4E-06 SR-D-NA DEMIN Fe-55 25.7. 0.8E-02 Co-58 17.% 0.7E-v Co-60 17.% .6E-02 Nt-63 19.% 3.0E-02 H-3 %. 4.1E-04 C-14 0.% 9.5E-06 7c-99 0.% 5.5E-06 I-129 0.% 0. E-06 Cs-137 5.% 0.4E-03 Pu-241 0.% 0.6E-05 Cm-242 0.% 1.8E-07 SR-D-NA RCS PUR Co-60 "S.% 1.!E-01 N1-59 0.% 1.1E-03 Nt-63 21. % 7.1E-02 H-! 0.% 4.5E-06 C-14 0.% .9E-05 Sr-90 0.% 9.8E-05 Tc-99 0.% 5.5E-07 '

I-129 0.% 0.6E-07 Cs-137 16.% 5.:E-02 Pu-241 0.% 4.!E-05 Cm-242 0.% 5.5E-07 NUMME OF TYPE OF TYPE OF MODE CF CLASS SHIPMENTS SHIPtMNT CONTAINER TRANSPORTATION DESTINATION A 1 LSA STRONG TIGHT TRUCK BARNWELL A 1 LSA USNRC CERIFIED CASK TRUCK BARNWELL A 4 LSA STRONG TIGHT TRUCK RICHLAND 25

• l of 6

. AITACHIGTf 1 TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)a TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/mL)

A. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10 7 Tanks b ,f,g,h,i Emitters c

1. Boric Acid I-131 1x10 8 Condensate P M Dissolved and 1x10 5 Entrained Gases

2. Waste One Batch /M (Gamma Emitters)

Condensate P M d

H-3 1x10 s Each Batch Composite

3. Laundry Gross Alpha 1x10 7 Waste

4. Turbine P Q Sr-89, Sr-90 5x10 s Building d Each Batch Composite Industrial Waste Fe-55 1x10.s Sumps *

5. Dry Cooling -

Tower Sumps

1. 1 and #2*

6. Regenerative Waste

7. Filter Flush

8. Waste ,

I

( 'When release from this source is batch in nature.

t PATERFORD - UNIT 3 3/4 11-2 L

F -

, 2 of 6

. ATIACHMENI 1 r

TABLE 4.11-1 (Continued)

.. LOWER LIMII MINIMJM OF OETECTION LIQUID RELEASE SAMPLINE ANALYSIS TYPE OF ACTIVITY (LLD)a TYPE FREQUENCY FREQUENCY ANALYSIS (pci/mt)

8. Continuous W W Principal Gamma 5x10 7 leases e.f e Grab Sample Emitters

1. Turbine I-131 1x10 6 Ouiiding Industrial h Waste Sugs**.

M M Dissolved and 1x10.s

2. Dry Cooling Grab sample Entrained Gases Tower 1 (Gamma Emitters) 54sup #1**

3. Dry Coolin0 W H H-3 1x10 s d

Tower i Grab Sample composita .

Grost Alpha 1x10 7 l 4. Circulating W Q d Srp, SrM 5x1pa l Wataf Grsh Sample Composite Discharge- Fe-55 1x10 8 Steam Gene-rator Slow-down HX

5. Auxiliary Camponent -

Cooling i Water Pumps l

l t

l l

! **When release from this source is continuous in nature.

WATERFORD - UNIT 3 3/4 11 3 M MNTNO.1

3 of 6 A%y TA81.E 4.!!-l (Continued)

LOWER LIMIT MINIMUN LIQUID RELEASE $ANFLING OF DETECTION ANALYSIS TYPE OF ACTIVITY TTFE FEZq0ERCY (LLD)a FREQUENCY ANALYSIS (uC1/aL)

5. Costinuous W W telaases e,f Continuoud Composited PT18CiPal Emitters

• Cama 5x10'I 6 Steam Generator I-131 1x10

-6 Blevdoes M achirge $1 M M Dissolved and 1x10-5 Crab Sample Entrained Gases (Gamma Emitters)

W N 11 - 3 lx10-5 Continoces Composite Cross Alpha Iz10' W Q 3r-89 Sr-90 5x10

-8 Continuousk composited Fe-55 1x10

-6 UATERFOsp - UNIT 3 3/4 11-3a AMENDMENT NO. 1

,-g - ,y,_, - - - - . . . , , . _ _ _ _ _ _ _ _ , - - - , , . _ ,,s.c_-,,__.y

s 4 of 6

• ATTACHtGTI 1 TABLE 4.11-1 (Continued)

TABLE NOTATION a The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66 s b LLD =

E V 2.22 x 10' Y exp (-Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, s[isthestandarddeviationofthebackgroundcountingrateorofthe cDunting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, l

2.22 x 108 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and at for plant effluents is the elapsed time between the midpoint of sample collection and the time of counting.

Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a_ posteriori (after the fact) limit for a particular measurement.

b A batch release is the discharge of liquid wastes of a di ; rete volume.

Prior to sampling for analyses, each batch shall be isola; d, and then thoroughly mixed by a method described in the ODCM to assuie representative sampling.

WATERFORD - UNIT 3 3/4 11-4

f a 5 of 6 I .' ATTACHMENT 1 TABLE 4.11-1 (Continued)

TABLE NOTATIONS C

The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, 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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also t;e analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to

? Specification 6.9.1.8.

d A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

'A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

1. P rior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

9 If the contents of the filter flush tank or the regenerative waste tank contain detectable radioactivity, no discharges from these tanks shall be made

to the UNRESTRICTED AREA and the contents of these tanks shall be directed to the liquid radwaste treatment system.

h Turbine Building Industrial Waste Sump (TBIWS)

The TBIWS shall be required to be sampled and analyzed in accordance with this table if any of the following conditions exist:

(1) Primary to secondary leakage is occurring; or, (2) Activity is present in the secondary system as indicated by either the SGB monitors or secondary sampling and analysis; or, (3) Activity was present in the TBIWS during the previous 4 weeks.

l If none of the above situations exists, then the sampling and analysis of this 1 stream need not be performed. ,

i Sampling and analysis of the dry cooling tower sumps and the auxiliary component cooling water pump discharge will be required only when detectable activity exists in the CCW.

Sampling and analysis of the circulating water discharge-steam generator blowdown heat exchanger discharge (CWD-SGB) will be required only when detectable activity exists in the secondary system.

WATERFORD - UNIT 3 3/4 11-5 l

6 of 6

.' ATTAQ9 TNT 1 Table 4.11-1 (Continued)

TABLE NoTATIoR$

3 Sampling and analysis of the stesa generator blowdown vill be required only when the blowdown is directed to the circulating water system or Eaterford 3 weste pond.

Steam generator blowdova to the Waterford 3 waste pond will be limited to ,

situatises requiring secondary chemia. cry control where the circulating Water system is not available or the secondary chemistry la outside the regstruments for.Cirealattag vetor System discharge. Blowdown to the waste poed will be terminated opeo detection of semple a 'ctivity greater than the uD lisrels of Table 4.11-1 section B. -

ho be repres'entative of the quantities and concentration of radioactive materials in liquid effluents,' samples shall be collected continuously in proportion to the rate of flow of the affluent stream.

I Steam generator blowdown discharge to the waste pond shall be limited to a period of six months with the circulating water system discha'rge' path not available unless. radiation monitoring and automatic isolation i tapabilities am added to the waste pond discharge path.

l 1

W N - UNIT 3 3/4 11-Se AmrmruerxT No.1 O

TABLE 4.11-2 sc RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

$ MINIMUM LOWER LIMIT OF 5 SAMPLING ANALYSIS TYPE OF DETECTION (LLD)*

i GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pci/mL) e P P 5

A. Waste Gas Holdup Each Tank Each Tank Principal Gamma Emitters D 1x10 4 Tanks Grab Sample P P c c B. Containment PURGE Each PURGE Each PURGE Principal Gamma Emitters D 1x10 4 (Plant Stack) Grab Sample M H-3 1x10 8 C

C.1 Plant Stack M 'd'I Principal Noble Gas Gamma b

1x10 4 Grab Sample M Emitters H-3 1x10 6 R

C.2 Fuel Handling M*'d Principal Noble Gas Gamma 1x10 4 Building Ven- Grab Sample M Emitters b U tilation H-3 1x10 6 0 (Normal)

Exhaust D.1 All Release Types Continuousf ,h,j gg I-131 1x10 12

. as listed in Charcoal B., C.1, and Sample I-133 1x10 80 C.2 above D.2 Main Condenser Continuousf ,h,j gg PrincipagParticulateGamma 1x10 81 Evacuation and Particulate Emitters Turbine Gland Sample Sealing System Continuousf ,h,j M Gross Alpha 1x10 88 Composite Particulate Sample Continuous ,h,j f

Q Sr-89, Sr-90 1x10 11 Composite Particulate Continuous ,h,j f

Sample Noble Gas Noble Gases 1x10 8

{*

Monitor Gross Beta or Gamma

' 2 of 3 6

ATTACHMENr 2 l

TABLE 4.11-2 (Continued)

TABLE NOTATION a

The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

D LLD =

E V 2.22 x 10' Y exp (-Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, s h is the standard deviation of the background counting rate or of the ccunting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 10s is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and at for plant effluents is the elapsed time between the midpoint of sample collection and the time of counting.

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

WATERFORD - UNIT 3 3/4 11-11

,s' 3 of 3 o AITACHtENT 2 TABLE 4.11-2 (Continued)

TABLE NOTATIONS b The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, i I-131, I-133, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8.

c Sampling shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following shutdown, startup, i

or a THERMAL POWER change exceeding 15% of RATED THERHAL POWER within a j 1-hour period. Analysis for principle gamma emitters as defined in (b) above shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of sampling.

d Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.

l i ' Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11. 2.1, 3.11. 2. 2, and 3.11. 2. 3.

9 5amples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing, or after removal from sampler.

l Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least l

I 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factnr of 3.

h If no primary to secondary leakage exists, then only the gross beta or gamma noble gases analysis need be performed for the main condenser evacuation and turbine gland sealing system. If a primary to secondary deak exists and the release from the main condenser evacuation and turbine gland sealing system has not been released via the plant stack, then the sampling and analysis must be performed.

Note (c) above is not applicable for the plant stack unless the noble gas monitor shows that effluent activity has increased by a factor of 3.

3 Fuel Handling Building sampling is required whenever irradiated fuel is in the storage pool.

WATERFORD - UNIT 3 3/4 11-12 l- - - - - . .-

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