ML20125A174
| ML20125A174 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/22/2018 |
| From: | Energy Harbor Nuclear Corp |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20125A185 | List:
|
| References | |
| L-20-137 | |
| Download: ML20125A174 (43) | |
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RTL #A9.621B Beaver Valley Power Station Unit 1/2 1/2-0DC-2.01 ODCM: LIQUID EFFLUENTS Document Owner Manager, Nuclear Environmental and Chemistry Revision Number 17 Level Of Use General Skill Reference Safety Related Procedure Yes Effective Date 01/22/18
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
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1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 2 of 43 TABLE OF CONTENTS 1.0 PURPOSE ......................................................................................................................................... 3 2.0 SCOPE .............................................................................................................................................. 3
3.0 REFERENCES
AND COMMITMENTS ......................................................................................... 3 3.1 References .............................................................................................................................. 3 3.2 Commitments ......................................................................................................................... 6 4.0 RECORDS AND FORMS .......,......................................................................................................... 7 4.1 Records ................................................................................................................................... 7 4.2 Forms ...................................................................................................................................... 7 5.0 PRECAUTIONS AND LIMITATIONS .......................................................................................... 7 5.1 Precautions ............................................................................................................................. 7 5.2 Limitations .............................................................................................................................. 7 6.0 ACCEPTANCE CRITERIA ............................................................................................................. 8 7.0 PREREQUISITES ............................................................................................................................ 8 8.0 PROCEDURE ................................................................................................................................... 8 8.1 Alarm Setpoints ...................................................................................................................... 8 8.1.1 BV-1 Monitor Alarm Setpoint Determination ........................,. .............................. 8 8.1.2 BV-2 Monitor Alarm Setpoint Determination ..................................................... 14 8.2 Compliance With 10 CFR 20 EC Limits (ODCM CONTROL 3.11.1.1) ............................ 20 8.2.1 Batch Releases ..................................................................................................... 20 8.2.2 Continuous Releases ............................................................................................ 22 8.3 Compliance With 10 CFR 50 Dose Limits (ODCM CONTROLS 3.11.1.2 And 3.11.1.3). 22 8.3.1 Cumulation OfDoses (ODCM CONTROL 3.ll.1.2) ......................................... 22 8.3.2 Projection Of Doses (ODCM CONTROL 3.11.1.3) ........................................... 24 8.4 Liquid Radwaste System ...................................................................................................... 25 8.4.1 BV-1 Liquid Radwaste System Components ...................................................... 25 8.4.2 BV-1 Laundry and Contaminated Shower Drain System Components .............. 29 8.4.3 BV-2 Liquid Radwaste System Components ...................................................... 29 8.4.4 BV-1/2 Miscellaneous Sumps ............................................................................. 31 ATTACHMENT A LIQUID SOURCE TERMS .............................................................................. 33 ATTACHMENT B INGESTION DOSE COMMITMENT FACTORS ........................................ 37 ATTACHMENT C LIQUID RADWASTE SYSTEM ..................................................................... 40 ATTACHMENT D SITE BOUNDARY FOR LIQUID EFFLUENTS ........................................... 43
Procedure Number:
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Title:
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ODCM: LIQUID EFFLUENTS 17 3 of 43 1.0 PURPOSE 1.1 This procedure provides the methodology to calculate dose, release concentrations, and alarm setpoints from liquid effluents in accordance with the requirements of in Beaver Valley Technical Specifications [TS] 5.5.2. <3.2.1) 1.1.1 Liquid effluent monitor alarm setpoints [TS] 5.5.2.a 1.1.2 Liquid effluent release concentration calculations [TS] 5.5.2.b) 1.1.3 Liquid effluent dose projection and cumulative dose calculations [TS] 5.5.2.d and
[TS] 5.5.2.e 1.1.4 Liquid Radwaste Treatment System [TS] 5.5.2.f 1.1.5 Site Boundary used for liquid effluents 2.0 SCOPE 2.1 This procedure is applicable to liquid effluents at Beaver Valley Power Station.
3.0 REFERENCES
AND COMMITMENTS 3.1 References 3.1.1 References For BV-1 Liquid Effluent Monitor Setpoints 3.1.1.1 Beaver Valley Power Station, Appendix I Analysis - Docket No. 50-334 and 50-412; Table 2.1-3 3.1.1.2 Beaver Valley Power Station, Appendix I Analysis - Docket No. 50-334 and 50-412; Table 2.1-2 3.1.1.3 Calculation Package No. ERS-SFL-92-039, Isotopic Efficiencies For Unit 1 Liquid Process Monitors 3.1.2 References for BV-2 Liquid Effluent Monitor Setpoints 3.1.2.1 Calculation Package No. ERS-SFL-86-026, Unit 2 DRMS Isotopic Efficiencies 3.1.2.2 Stone and Webster Computer Code LIQlBB; Nopnal Liquid Releases From A Pressurized Water Reactor" 3.1.2.3 Calculation Package No. ERS-JWW-87-015, Isotopic Efficiencies For 2SGC-RQ100 3.1.2.3.1 The Isotopic Efficiencies for 2SGC-RQ100 are superseded by the values presented in Calculation Package No. ERS-SFL-86-026.
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ODCM: LIQUID EFFLUENTS 17 4 of 43 3.1.2.4 Calculation Package No. ERS-WFW-87-021, Conversion Factor for 2SGC-RQ100 3.1.2.4.1 The Monitor Conversion Factor (CF 11) for 2SGC-RQ 100 is superseded by the value presented in Calculation Package No. ERS-ATL-93-021.
3.1.3 References used in other sections of this procedure 3.1.3.1 NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants 3.1.3.2 NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors (Generic Letter 89-01, Supplement No. 1) 3.1.3.3 NUREG-0017; Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from PWRs, Revision 0 3 .1.3 .4 Regulatory Guide 1.113; Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I, April 1977 3.1.3.5 Regulatory Guide 1.109; Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance to 10 CFR Part 50, Appendix I 3.1.3.6 Calculation Package No. ERS-ATL-83-027; Liquid Waste Dose Factor Calculation for Y2-ENV-05.06 3.1.3.7 Calculation Package No. ERS-ATL-93-021, Process Alarm Setpoints For Liquid Effluent Monitors 3.1.3.8 Calculation Package No. ERS-LMR-15-006, Additional Effluent Dose Factors 3.1.3.9 Stone and Webster Calculation Package No. UR(B)-160, BVPS Liquid Radwaste Releases and Concentrations - Expect and Design Cases (per Unit and Site)
"3.1.3.10 10 CFR.20, Appendix B, (20.1001-20.2402) Table 2, Column 2 EC's 3.1.3.11 NUREG-0172; Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake 3.1.3.12 UCRL-50564; Concentration Factors of Chemical Elements in Edible Aquatic Organisms, Revision 1, 1972 3.1.3.13 1/2-ADM-1640, Control of the Offsite Dose Calculation Manual 3.1.3.14 1/2-ADM-0100, Procedure Writers Guide 3.1.3.15 NOP-SS-3001, Procedure Review and Approval
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 5 of 41 3.1.3.16 l/2-0DC-3.03, ODCM: Controls for RETS and REMP Programs 3.1.3.17 CR 02-06174, Tracking of Activities for Unit 1 RCS Zinc Addition Implementation. CA-014, Revise ODCM Procedure l/2-0DC-2.01 (Tables 1.1-la and 1b) to include the addition of Zn-65 to the ODCM liquid source term.
3.1.3.18 CR 03-02466, RFA-Radiation Protection Effluent Control Provide Recommendation on Processing when Performing Weekly Sample of
[1LW-TK-7A/7B]. CA-02, Revise ODCM Procedure l/2-0DC-2.01, (Attachment D) to show the liquid waste flow path cross-connect between Unit 1 and Unit 2.
3.1.3.19 CR 05-03306, Incorporated Improved Technical Specifications (ITS).
3.1.3.20 CR 05-03854, ODCM Figure for Liquid Effluent Release Points Need Updated.
CA-01, revise ODCM procedure l/2-0DC-2.01 (ODCM: Liquid Effluents)
Attachment D, Figure 1.4-3 to incorporate a modified version of Plant Drawing No. 8700-RM-27F.
3.1.3.21 Unit 1 Technical Specification Amendment No. 275 (LAR lA-302) to License No. DPR-66. This amendment to the Unit 1 license was approved by the NRC on July 19, 2006.
3.1.3.22 Vendor Calculation Package No. 8700-UR(B)-223, Impact of Atmospheric Containment Conversion, Power Uprate, and Alternative Source Terms on the Alarm Setpoints for the Radiation Monitors at Unit 1.
3.1.3.23 Engineering Change Package No. ECP-04-0440, Extended Power Uprate.
3.1.3.24 CR 06-04908, Radiation Monitor Alarm Setpoint Discrepancies. CA-03; revise ODCMprocedure l/2-0DC-2.01 to update the alarm setpoints of[RM-lRW-100]
and [RM-lDA-100] for incorporation of the Extended Power Uprate per Unit 1 TS Amendment No. 275.
3.1.3.25 CR 06-6476, Procedure l/2-0DC-2.01 needs revised for Plant Uprate. CA-01; revise ODCM procedure l/2-0DC-2.01 to update the alarm setpoints of
[2SWS-RQ101] for incorporation of the Extended Power Update at Unit 2 (ECP-04-0441) per Unit 2 TS Amendment No. 156.
3.1.3.26 CR 05-00004-15, CR05-00004-17 and SAP Order 200197646-0010 to revise l/2-0DC-2.01. Add the Coolant Recovery Tanks [1BR-TK-4A/4B] as Liquid Waste Tanks to Section 8.4 description and Attachment D Figures 1.4-1 and 1.4-2.
Add a default 2-tank volume recirculation time of 45.7 hrs for the Coolant Recovery Tanks [1BR-TK-4A/4B] to Attachment B Table 1.2-la. Add the Cesium Removal Ion Exchangers [lBR-1-lA/lB and 2BRS-IOE21A/21B] to Section 8.4 description and Attachment B Figures 1.4-1 and 1.4-2. Revise the recirculation times in Attachment B Table 1.2-1 a and 1.2-1 b to indicate the times for nominal tank volume and maximum tank volume.
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 6 of 43 3.1.3.27 SAP Order 200197646-0660. Revise l/2-0DC-2.01 Attachment D Figure 1.4-3 to remove STP Outfalls 113 and 203 due to retirement of the Sewage Treatment Plants and to remove Ul Steam Generator Blowdown Filter Backwash Outfall 501. Water is no longer discharged via these outfalls.
3.1.3.28 SAP Order 200197646-0810. Revise l/2-0DC-2.01 to incorporate alarm setpoints for all possible detector combinations for [RM-lDA-100]. Specifically, due to obsolescence of the original Model 843-30 and 843-32 detectors that were previously installed in [RM-lDA-100], the vendor has upgraded them to Model 843-30R and 843-32R detectors, which include upgraded efficiency data as well.
3.1.3.29 CR 10-86844 revises l/2-0DC-2.01 to remove description that batch releases of liquid waste are processed by recirculation through eductors. Deleted Attachment B which referenced minimum liquid waste batch release recirculation times and added description that liquid waste recirculation times to achieve two tank volumes are calculated based upon actual tank volume and pump capacity.
3.1.3.30 ECP 11-0049 and CR 2012-02583 implement changes to the design of the liquid waste system for Phase 2 of Coolant Recovery Project.
3.1.3.31 SAP Notification 600747531, Update l/2-0DC-2.01 for RM-tRW-100.
3.1.3.32 CR-2012-05875,Antimony-126 identified in the liquid waste system.
3.1.3.33 SAP Notification 600765150, Request from Operations to allow discharge of water in high level drains tanks [LW-TK-2A/B] through low level waste tanks
[LW-TK-3A/B] via RM-LW-104.
3.1.3.34 CA 2012-15547-7, To address the extent of condition with potential gaps in Radiological Effluent Program, evaluate the need to place appropriate ODCM controls on various non-radiological tanks and sumps throughout the site.
3.1.3.35 Engineering Change Package (ECP) 12-0478, Liquid Waste Demin Project:
Installation of new Demineralizers in Solid Waste Building.
3.2 Commitments 3.2.1 Beaver Valley Technical Specifications: [TS] 5.5.2, Radioactive Effluent Controls Program.
Procedure Number:
Beaver Valley Power Station 1/2-0DC-2.01
Title:
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1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 7 of 43 4.0 RECORDS AND FORMS 4.1 Records 4.1.1 Any calculation supporting ODCM changes shall be documented, as appropriate, by a retrievable document (e.g.; letter or calculation package) with an appropriate RTL number.
4.2 Forms 4.2.1 None 5.0 PRECAUTIONS AND LIMITATIONS 5.. 1 Precautions 5.1.1 None 5.2 Limitations 5.2.1 In Section 8.1, Alarm Setpoints, of this procedure effluent monitor setpoints for a conservative mix are based on the individual unit's specific parameters, but effluent monitor setpoints for analysis prior to release permit use the total dilution flow available at the site.
5.2.2 BV-1 and BV-2 utilize the concept of a shared liquid radioactive waste system according to NUREG-0133. <3.1J.1) This permits the mixing of liquid radwaste for processing and allocating of dose due to release as defined in Section 8.4, Liquid Radwaste System.
5.2.3 A difference in alarm setpoint terminology for the radiation monitoring systems ofBV-1 and BV-'2 is described as follows:
5.2.3.1 HIGH and HIGH-HIGH terminology are used for BV-1 monitors and ALERT and HIGH terminology is used for BV-2 monitors.
5.2.3.2 BV-1 alarm setpoint units are expressed as counts per minute (cpm) and BV-2 alarm set'points units are expressed as microcurie per milliliter (uCi/mL). The difference is due to BV-2 software which applies a conversion factor to the raw data (cpm) to convert units to uCi/mL. Note that the uCi/mL presentation is technically correct only for the specific isotopic mix used in the determination of the conversion factors. Therefore, BV-2 setpoints determined on analysis prior to release will be correct for properly controlling dose rate, but the indicated uCi/mL value may differ from the actual value.
5.2.4 This procedure also contains information that was previously contained in Section 5 of the previous BV-1 and 2 Offsite Dose Calculation Manual.
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 8 of 43 5.2.4.1 In regard to this, the site boundary for liquid effluents was included in this procedure.
5.2.4.2 The Site Boundary _for Liquid Effluents is shown in ATTACHMENT D Figure 5-1.
6.0 ACCEPTANCE CRITERIA 6.1 Changes to this procedure shall contain sufficient justification that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR 50, and not adversely impact the accuracy or reliability of effluent dose or alarm setpoint calculation. C3 -13 *2) 6.1.1 Changes to this procedure shall be prepared in accordance with 1/2-ADM-0100, PROCEDURE WRITER'S GUIDE c3.u.1 4) and 1/2-ADM-1640, CONTROL OF THE OFFSITE DOSE CALCULATION MANUAL. c3.u.1 3) 6.1.2 Changes to this procedure shall be reviewed and approved in accordance with NOP-SS-3001 PROCEDURE REVIEW AND APPROVAL c3.u. 15) and 1/2-ADM-1640. ' C3.u. 13 )
7.0 PREREQUISITES 7.1 None 8.0 PROCEDURE 8.1 Alarm Setpoints 8.1.1 BV-1 Monitor Alarm Setpoint Determination This procedure determines the monitor HIGH-HIGH Alarm Setpoint (HHSP) to provide indication if the concentration of radionuclides in the liquid effluent released from the site to unrestricted areas exceeds 10 times the Effluent Concentrations (ECs) specified in 10 CFR 20, Appendix B (20.1001-20.2402), Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases or exceeds a concentration of 2E-4 uCi/mL for dissolved or entrained noble gases. c3.u.s)
The methodology described in Section 8.1.1.2 is an alternative method to be used to determine the [RM-ILW-104], LIQUID WASTE EFFLUENT RADIATION MONITOR or
[RM-lLW-116], LIQUID WASTE CONTAMINATED DRAINS RADIATION MONITOR monitor HHSP. The methodology in Section 8.1.1.2 may be used for any batch release and shall be used when the respective total gamma activity concentration of the liquid effluent prior to dilution exceeds 3.14E-3 uCi/mL and 7.33E-3 uCi/mL. This concentration is equivalent to the respective HHSPs derived in Section 8.1.1.1 and allows for respective tritium concentrations up to 4.26E+O uCi/mL and 9.94E+O uCi/mL. (3.u.s)
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 9 of43 8.1.1.1 BV-1 Setpoint Determination Based On A Conservative Mix The Alarm Setpoints shall be set at the values listed in the following table:
BV-1 LIQUID MONITOR SETPOINTS cpm Above Background Monitor CR HHSP HSP Liquid Waste Effluent Monitor RM-lLW-104 4.15E+5 s 4.15E+5 s 2.9IE+5 (953-36E)
Laundry And Contaminated RM-lLW-116 9.69E+5 s 9.69E+5 s 6.78E+5 Shower Drains Monitor (953-36E)
Component Cooling/ RM-lRW-100 2.57E+4 ::; l .90E+4 <3) ::; l.33E+4 <3)
Recirculation Spray Hx River (843-30)
Water Monitor Component Cooling Hx River RM-lRW-101 9.03E+3 s 9.03E+3 s 6.32E+3 I,
Water Monitor (843-30)
Aux Feed Pump Bay Drain RM-lDA-100 l.22E+4 (I) ::; l.20E+4 (I) :<;:; 8.43E+3 0)
I Monitor (843~30 I 843-32)
I l.05E+4 <2) :-;:; l.05E+4 <2) :<;:; 7.33E+3 <2)
I I
RM-lDA-100 l.22E+4 (I) :<;:; l.20E+4 (I) :<;:; 8.43E+3 (I)
(843-30R I 843-32R) l.22E+4 <2) :-;:; l.22E+4 <2) :-;:; 8.52E+3 <2>
(I) Use these values for a monitor with an analog drawer/meter face. These values are from Calculation No.
8700-UR(B)-223, and are justified for use in Attachment 6 of Calculation Package ERS-ATL-93-021. (3.1. 3 .8) <3* 13 *22)
<2l Use these values when the monitor is upgraded to a digital drawer/meter face. These values are justified for use in Attachment 6 of Calculation Package ERS-ATL-93-021 (3.1.3. 8)
<3) Calculation Package ERS-ATL-93-021 Revision 4, for detector efficiency at elevated river temperatures. <3
- 13 *8>
The setpoint bases for all monitors can be found in Calculation Package ERS-ATL-93-021 and/or Calculation No. 8700-UR(B)-223. (3.u.22) The setpoints for RM-lLW-104 and RM-lLW-116 are based on the following conditions:
- Sourc~ terms given in ATTACHMENT A Table 1.1-la. These source terms (without Zn-65) have been generated from the GALE Computer Code, as described in NUREG-0017. <3 .1.3.3) The inputs to GALE are given in l/2-0DC-3.01 Appendix B. The Zn-65 source term was generated via Calculation Package No.
ERS-ATL-93-021. <3.u.s, 3.u.i 7)
- Dilution water flow rate of22,800 gpm = (15,000 gpm BV-1 + 7,800 gpm BV-2) .
- Discharge flow rate prior to dilution of 3 5 gpm for the Liquid Waste Effluent Monitor (RM-ILW-104).
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 10 of 43
- Discharg~ flow rate prior to dilution of 15 gpm for the Laundry and Contaminated Shower Drains Monitor [RM-lLW-116].
The above setpoints for [RM-1LW-104] and [RM-1 L W-116] can be varied based on actual operating conditions resulting in changes in the discharge and dilution flow rates as follows:
542F HHSP=-- [1.1(1)-1]
f where:
HHSP = Monitor HIGH-HIGH Alarm Setpoint above background (ncpm).
542 = Most restrictive proportionality constant based on nominal flow conditions:
542 = 3.53E+5 ncpm x 35 gpm-,- 22,800 gpm [RM-lLW-104]
542 = 8.24E+5 ncpm x 15 gpm -:-22,800 gpm [RM-lLW-116]
F = Dilution water flow rate (gpm), BV-1 plus BV-2 Cooling Tower Blowdown Rate (not including release through the Emergency Outfall Structure).
f = Discharge flow rate prior to dilution (gpm).
8.1.1.1.1 BV-1 Mix Radionuclides The "mix" (radionuclides and composition) of the liquid effluent was determined as follows:
- The liquid source terms that are representative of the "mix" of the liquid effluent were determined. Liquid source terms are the radioactivity levels of the radionuclides in the effluent from ATTACMENT A Table 1.1-la.
- The fraction of the total radioactivity in the liquid effluent comprised by radionuclide "i" (Si) for each individual radionuclide in the liquid effluent was determined as follows:
[1.1(1)-2]
where:
Ai= Annual release of radionuclide "i" (Ci/yr) in the liquid effluent from ATTACHMENT A Table 1.1-la.
Procedure Number:
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17 11 of 41 8.1.1.1.2 BV-1 Maximum Acceptable Concentration (All Radionuclides)
The maximum acceptable total radioactivity concentration (uCi/mL) of all radionuclides in the liquid effluent prior to dilution (Ct) was determined by:
Ct=....:.F_ _ [1.1(1)-3]
f~ _fu_
i OECi where:
F = Dilution water flow rate (gpm), BV-1 plus BV-2 Cooling Tower Blowdown Rate (not including release through the Emergency Outfall Structure).
= 22,800 gpm = (15,000 gpm BV-1 + 7,800 gpm BV-2) f = Maximum acceptable discharge flow rate prior to dilution (gpm).
= 35 gpm for Liquid Waste Effluent Monitor [RM-lLW-104].
= 15 gpm for Laundry and Contaminated Shower Drains Monitor [RM-lLW-116].
OECi = The ODCM liquid effluent concentration limit for radionuclide "i" (uCi/mL) from ATTACHMENT A Table 1.1-la. The OEC is set at 10 times the 10 CFR 20, Appendix B (20.1001-20.2402) Table 2, Col. 2 EC values.
I Si = The fraction of total radioactivity attributed to radionuclide "i", from Equation [1.1(1)-2].
8.1.1.1.3 BV-1 Maximum Acceptable Concentration (Individual Radionuclide)
The maxjmum acceptable radioactivity concentration (uCi/mL) of radionuclide "i" in the liquid effluent prior to dilution (Ci) was determined by:
[1.1 (1 )-4]
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 12 of 43 8.1.1.1.4 BV-1 Monitor Count Rate The calculated monitor count rate (ncpm) above background attributed to the radionuclides; (CR) was determined by:
[1.1(1)-5]
where:
Ei = Detection efficiency of the monitor for radionuclide "i" (cpm/uCi/mL) from ATTACHMENT A Table 1.1-la. If not listed in Attachment A, then obtained from Calculation Package ERS-SFL-92-039. (3.1.1.4) 8.1.1.1.5 BV-1 Monitor HHSP The monitor HHSP above background (ncpm) should be set at the CR value.
Since only one tank can be released at a time, adjustment of this value is not necessary to compensate for release from more than one source.
8.1.1.2 BV-1 Setpoint Determination Based On Analysis Prior To Release The following method applies to liquid releases when determining the setpoint for the maximum acceptable discharge flow rate prior to dilution and the associated HHSP Alarm Setpoint based on this flow rate for the [RM-lLW-104], LIQUID WASTE EFFLUENT MONITOR and the [RM-lLW-116], LAUNDRY AND CONTAMINATED SHOWER DRAINS MONITOR during all operational conditions.
The monitor alarm setpoint is set slightly above (a factor of 1.25) the count rate that results from the concentration of gamma emitting radionuclides in order to avoid spurious alarms. To compensate for this increase in the monitor alarm setpoint, the allowable discharge flow rate is reduced by the same factor.
When the discharge flow rate is limited by the radwaste discharge pump rate capacity or by administrative selection rather than the allowable flow rate determined from activity concentration, the alarm setpoint will be proportionally adjusted based upon the excess dilution factor provided.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 13 of 43 8.1.1.2.1 BV-1 Maximum Acceptable Discharge Flow Rate The maximum acceptable discharge flow rate (f) prior to dilution (gpm) is determined by: '
f= F [1.1(1)-6]
1.25 ~ Ci i OECi where:
F = Dilution water flow rate, BV-1 plus BV-2 Cooling Tower Blowdown (gpm).
The dilution water flow rate may include the combined cooling tower blowdown flow from both units exiting the discharge structure (but excluding emergency outfall structure flow) when simultaneous liquid discharges are administratively prohibited.
Ci = Radioactivity concentration ofradionuclide "i" in the liquid effluent prior to dilution (uCi/mL) from analysis of the liquid effluent to be released.
1.25 = A factor to prevent spurious alarms caused by deviations in the mixture of radionuclides which affect the monitor response.
OECi = The ODCM liquid effluent concentration limit for radionuclide "i" (uCi/mL) from ATTACHMENT A Table 1.1-la. The OEC is set at 10 times the 10 CFR 20, Appendix B (20.l 001-20.2402) Table 2, Col.
2 EC values.
8.1.1.2.2 BV-1 Monitor Count Rate The calculated monitor count rate (ncpm) above background attributed to the radionuclides, (CR) is determined by:
CR = 1.25 ~ Ci Ei [1.1(1)-7]
I where:
Ei = The detection efficiency of the monitor for radionuclide "i!'
(cpm/uCi/mL) from ATTACHMENT A Table 1.1-la. If not listed in Attachment A, then obtained from Calculation Package ERS-SFL-92-039. (3.1.1. 4) 1.25 = A factor to prevent spurious alarms caused by deviations in the mixture of radionuclides which affect the monitor response.
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 14 of 43 8.1.1.2.3 BV-1 Monitor HHSP The liquid effluent monitor HHSP above background (ncpm) should be set at the CR value adjusted by any excess dilution factor provided as defined in the following equation:
HHSP=CR f [1.1(1)-8]
f where:
HHSP = Monitor HHSP above background.
CR = Calculated monitor count rate (ncpm) from equation [1.1(1)-7].
f = Maximum acceptable discharge flow rate prior to dilution determined by equation [1.1 (1 )-6].
f = Actual maximum discharge flow rate to be maintained for the discharge. The reduced value off may be due to pump limitations or administrative selection. '
8.1.2 BV-2 Monitor Alarm Setpoint Determination This procedure determines the monitor HIGH Alarm Setpoint (HSP) that indicates if the concentration of radionuclides in the liquid effluent released from the site to unrestricted areas exceeds 10 times the ECs specified in 10 CPR 20, Appendix B (20.1001-20.2402),
Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases or exceeds a concentration of 2E-4 uCi/mL for dissolved or entrained noble gases. <3 .1.3. 8)
The methodology described in Section 8.1.2.2 is an alternative method to be used to determine the [2SGC-RQ100], LIQUID WASTE EFFLUENT RADIATION MONITOR HSP. The methodology in Section 8.1.2.2 may be used for any batch release and shall be used when the total gamma radioactivity concentration of the liquid effluent prior to dilution exceeds l.14E-3 uCi/mL. This concentration is equivalent to a monitor response and HSP derived in Section 8.1.2.1 and allows for a tritium concentration ofup to 2.16E+O uCi/mL. The setpoint was obtained by use of a conversion factor of 5.61E-9 uCi/mL/cpm determined for the nuclide mix. <3
- 13 *8>
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
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1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 15 of 43 8.1.2.1 BV-2 Setpoint Determination Based On A Conservative Mix The Alarm Setpoints shall be set at the values listed in the following Table:
BV-2 LIQUID MONITOR SETPOINTS
µCi/mL Above Background Monitor DV HSP ASP Liquid Waste Effluent 2SGC-RQ100 l.14E-3 ::; 1.14E-3 ::; 7.99E-4 Monitor Service Water Monitor 2SWS-RQ101 4.30E-5 ::; 4.30E-5 ::; 3.0lE-5 Service Water Monitor 2SWS-RQ102 4.30E-5 ::; 4.30E-5 ::; 3.0lE-5 The setpoint for [2SGC-RQ 100] is based on the following conditions, however, the setpoint bases for [2SWS-RQ101] and [2SWS-RQ102] can be found in Calculation Package ERS-ATL-93-021. <3.1.3. 8)
- Source terms given in ATTACHMENT A Table 1.1-1 b. These source terms (without Zn-65) have been generated by using models and input similar to NUREG-0017. The inputs are given in l/2-0DC-3.01. The Zn-65 source term was generated via Calculation Package No. ERS-ATL-93-021. (3.n8, 3.ni 7)
- Dilution water flow rate of22,800 gpm = (15,000 gpm BV-1 + 7,800 gpm BV-2).
- Discharge flow rate prior to dilution of 80 gpm for the Liquid Waste Effluent Monitor [2SGC-RQ100].
- A software conversion factor of 5.61E-9 uCi/mL/cpm associated with Liquid Waste Effluent Monitor [2SGC-RQ100]. (3.1.3. 8)
The above setpoint for [2SGC-RQ 100] can be varied based on actual operating conditions resulting in the discharge and dilution flow rates as follows:
HSP = 4.00E-6 F [1.1(2)-1]
f where:
HSP = HSP (uCi/mL) above background.
4.00E-6 = Proportionality constant based on nominal flow conditions:
4.00E-6 = 1.14E-3 net uCi/mL x 80 gpm + 22,800 gpm F = Dilution water flow rate, BV-1 plus BV-2 Cooling Tower Blowdown Rate (gpm).
f = Discharge flow rate prior to dilution (gpm).
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ODCM: LIQUID EFFLUENTS 17 16 of 43 8.1.2.1.1 BV-2 Mix Radionuclides The "mix" (radionuclides and composition) of the liquid effluent was determined as follows: ,
- The liquid source terms that are representative of the "mix" of the liquid effluent were determined. Liquid source terms are the radioactivity levels of the radionuclides in the effluent from ATTACHMENT A Table 1.1-lb.
- The fraction of the total radioactivity in the liquid effluent comprised by radionuclide "i" (Si) for each individual radionuclide in the liquid effluent was determined as follows:
Si= Ai [1.1(2)-2]
LAi i
where:
Ai =Annual release ofradionuclide "i" (Ci/yr) in the liquid effluent from ATTACHMENT A Table 1.1-lb.
8.1.2.1.2 BV-2 Maximum Acceptable Concentration (All Radionuclides)
The maximum acceptable total radioactivity concentration (uCi/mL) of all radionuclides in the liquid effluent prior to dilution (Ct) was determined by:
Ct=_F_ [1.1 (2)-3]
fLi~
OECi where:
F = Dilution water flow rate (gpm), BV-1 plus BV-2 Cooling Tower Blowdown Rate (not including release out through the Emergency Outfall Structure).
= 22,800 gpm = (15,000 gpm BV-1 + 7,800 gpm BV-2).
f = Maximum acceptable discharge flow rate prior to dilution (gpm).
= 80 gpm for Liquid Waste Process Effluent Monitor [2SGC-RQ100].
OECi = The ODCM liquid effluent concentration limit for radionuclide "i" (uCi/mL) from ATTACHMENT A Table 1.1-lb. The OEC is set at 10 times the 10 CPR 20, Appendix B (20.1001-20.2402) Table 2, Col. 2 EC values.
Si = The fraction of total radioactivity attributed to radionuclide "i", from Equation [ 1.1 (2)-2].
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17 17 of 43 8.1.2.1.3 BV-2 Maximum Acceptable Concentration (Individual Radionuclide)
The maximum acceptable radioactivity concentration (uCi/mL) ofradionuclide "i" in the liquid effluent prior to dilution (Ci) was determined by:
[1.1(2)-4]
8.1.2.1.4 BV-2 Monitor Display Value The calculated monitor Display Value (uCi/mL) above background attributed to the radionuclides; (DV), was determined by:
DV = 5.61E-9 Li Ci Ei [l.1(2)-5]
where:
5.61E-9 = Conversion factor (uCi/mL/cpm), an average determined for the source term mix.
Ei = Detection efficiency of the monitor for radionuclide "i" (cpm/uCi/mL) from ATTACHMENT A Table 1.1-lb. If not listed there, from Calculation Package ERS-SFL-86-026. <3 .1.2 -2) 8.1.2.1.5 BV-2 Monitor HSP The monitor HIGH Alarm Setpoint above background (uCi/mL) should be set at the DV value.
8.1.2.2 BV-2 Setpoint Determination Based On Analysis Prior To Release The following method applies to liquid releases when determining the setpoint for the maximum acceptable discharge flow rate prior to dilution and the associated HIGH Alarm Setpoint based on this flow rate for the Liquid Waste Effluent Monitor (2SGC-RQ100) during all operational conditions.
The monitor alarm setpoint is set slightly above (a factor of 1.25) the concentration reading that results from the concentration of gamma emitting radionuclides in order to avoid spurious alarms. To compensate for this increase in the monitor alarm setpoint, the allowable discharge flow rate is reduced by the same factor.
When the discharge flow rate is limited by the radwaste discharge pump rate capacity or by administrative selection rather than the allowable flow rate determined from activity concentration, the alarm setpoint will be proportionally adjusted based upon the excess dilution factor provided.
Procedure Number:
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ODCM: LIQUID EFFLUENTS 17 18 of 43 8.1.2.2.1 BV-2 Maximum Acceptable Discharge Flow Rate The maximum acceptable discharge flow rate (f) prior to dilution (gpm) is determined by:
f= F [1.1(2)-6]
1.25 Li Ci OECi where:
F = Dilution water flow rate, BV-1 plus BV-2 Cooling Tower Blowdown (gpm).
The dilution water flow rate may include the combined cooling tower blowdown flow from both units exiting the discharge structure (but excluding emergency outfall structure flow) when simultaneous liquid discharges from both plants are administratively prohibited.
Ci = Radioactivity concentration ofradionuclide "i" in the liquid effluent prior to dilution (uCi/mL) from analysis of the liquid effluent to be released.
1.25 = A factor to prevent spurious alarms caused by deviations in the mixture of radionuclides which affect the monitor response.
OECi = The ODCM liquid effluent concentration limit for radionuclide "i" (uCi/mL) from Table 1.1-lb. The OEC is set at 10 times the 10 CFR 20, Appendix B (20.1001-20.2402) ATTACHMENT A Table 2, Col. 2 EC values.
Procedure Number:
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17 ' 19 of 43 8.1.2.2.2 BV-2 Monitor Display Value The calculated monitor Display Value (uCi/mL) above background attributed to the radionuclides; (DV) is determined by:
DV = (1.25) (5.61E-9) I:i Ci Ei [1.1(2)-7]
where:
Ei = The detection efficiency of the monitor for radionuclide "i" (cpm/uCi/mL) from ATTACHMENT A Table 1.1-lb. If not listed there, from Calculation Package ERS-SFL-86-026. c3 .1.2 -2l 1.25 = A factor to prevent spurious alarms caused by deviations in the mixture of radionuclides which affect the monitor response.
5.61E-9 = Conversion factor (uCi/mL/cpm), an average determined for the source term mix.
8.1.2.2.3 BV-2 Monitor HSP The liquid effluent monitor HSP above background (uCi/mL) should be set at the DV value adjusted by any excess dilution factor provided as defined in the following equation:
HSP=DVf [1.1(2)-8]
f where:
HSP = HSP above background.
DV = Calculated monitor concentration reading (uCi/mL) from equation
[1.1 (2)-7].
F = Maximum acceptable discharge flow rate prior to dilution determined by equation [l.1(2)-6].
f = Actual maximum discharge flow rate to be maintained for the discharge. The reduced value off may be due to pump limitations or administrative selection.
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,ODCM: LIQUID EFFLUENTS 17 20 of 43 8.2 Compliance With 10 CFR 20 EC Limits (ODCM CONTROL 3.11.1.1}
8.2.l Batch Releases 8.2.1.l Pre-Release The radioactivity content of each batch release will be determined prior to release in accordance with l/2-0DC-3.03, Table 4.11-1. In order to assure representative samples, at least two (2) tank volumes of entrained fluid from each tank to be discharged shall be recirculated. To meet this requirement tank recirculation time is calculated using actual tank volumes and recirculation pump capacity. BV-1 and BV-2 will show compliance with ODCM Control 3.11.1.1 in the following manner:
The activity of the various radionuclides in the batch release, determined in accordance with l/2-0DC-3.03, Table 4.11-1, is divided by the minimum dilution flow to obtain the concentration at the unrestricted area. This calculation is shown in the following equation:
Conci = Ci R [1.2-1]
MDF where:
Conci = Concentration ofradionuclide "i" at the unrestricted area (uCi/mL).
Ci = Concentration ofradionuclide "i" in the potential batch release (uCi/mL).
R = Release rate of the batch (gpm).
MDF = Minimum dilution flow (gpm). (May be combined BV-1/BV-2 flow when simultaneous liquid discharges are administratively prohibited).
The projected concentrations in the unrestricted area are compared to the OECs.
Before a release is authorized, Equation [ 1.2-2] must be satisfied.
~i (Conci/OECi) < 1 [1.2-2]
where:
OECi = The ODCM effluent concentration limit of radionuclide "i" (uCi/mL) from ATTACHMENT A Table 1.1-la and 1.1-lb. The OEC is set at 10 times the 10 CFR 20, Appendix B, (20.1001-20.2402) Table 2, Col. 2 EC values. (3. u.1o)
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17 21 of 43 8.2.1.2 Post-Release Following release from the batch tank, the Post Dose Correction Factor will be calculated in the following manner:
PDCF = {VA1)/(DFA) [1.2-3]
(VI1)/(DFI) where:
PCDF = Post Dose Correction Factor.
VAt = Actual Volume of tank released (gal).
DF A = Actual Dilution Flow during release (gpm).
Vlt = Initial Volume authorized for release (gal).
DFI = Initial Dilution Flow authorized for release (gpm).
The concentration of each radionuclide following release from the batch tank will be calculated in the unrestricted area in the following manner when the Post Dose Correction Factor shown in equation [1.2-3] is> 1: '
The average activity of radionuclide "i" during the time period of release is divided by the actual dilution flow during the period of release to obtain the concentration in the unrestricted area. This calculation is shown in the following equation:
ConCik = Cik V tk [1.2-4]
ADFk where:
Concik = The concentration of radionuclide "i" (uCi/mL) at the unrestricted area, during the release period of time k.
NOTE: Since discharge is from an isolated well-mixed tank at essentially a uniform rate, the difference between average and peak concentration within any discharge period is minimal.
Cik = Concentration of radionuclide "i" (uCi/mL) in batch release during time period k.
Vtk = Volume of Tank released during time period k (gal).
ADFk = Actual volume of Dilution Flow during the time period of release k (gal).
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1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 22 of 43 To show compliance with ODCM CONTROL 3 .11.1.1, the following relationship must be satisfied:
[1.2-5]
8.2.2 Continuous Releases Continuous releases ofliquid effluents do not normally occur at BV-1 or BV-2. When they do occur, the concentration of various radionuclides in the unrestricted area would be calculated using Equation [ 1.2-1] with Cik, the concentration of isotope i in the continuous release. To show compliance with ODCM CONTROL 3.11.1.1, Equation [1.2-5] must again be satisfied.
8.3 Compliance With 10 CFR 50 Dose Limits (ODCM CONTROLS 3.11.1.2 And 3.11.1.3)
BV-1 and 2 utilize the concept of a shared liquid radioactive waste system according to NUREG-0133. (3.1.3.l) This permits mixing of the liquid radwaste for processing. Since the resulting effluent release cannot accurately be attributed to a specific reactor unit, the treated effluent releases are allocated as defined below.
8.3.1 Cumulation Of Doses (ODCM CONTROL 3.11.1.2)
The dose contribution from the release of liquid effluents will be calculated monthly for each batch release during the month and a cumulative summation of the total body and organ doses will be maintained for each calendar month, current calendar quarter, and the calendar year to date. The dose contribution will be calculated using the following equation:
D, = UAF L Ah Lm~tk Cik Fk [1.3-1]
i k=l where:
D. = The cumulative dose commitment to the total body or any organ, t, from the liquid effluents for the total time period m
L ~tk (mrem) k=l
~tk = The length of the kth release over which Cik and Fk are averaged for all liquid releases (hours).
Cik = The average concentration of radionuclide, "i" (uCi/mL), in undiluted liquid effluent during time period ~tk from any liquid release.
Ai, = The site related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma and beta emitter (mrem-mL per hr-uCi) from ATTACHMENT B Table 1.3-1.
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17 23 of 41 m = Number of releases contributing to the cumulative dose, Dt.
UAF = Unit allocation factor. Provides apportionment of dose between BV-1 and BV-2. Normally set at 0.5 for each unit. (Must total to~ 1.0).
Fk = The near field average dilution factor for Cik during any liquid effluent release.
Defined as the ratio of the average undiluted liquid waste flow to the product of the average flow from the site discharge structure during the report period to unrestricted receiving waters, times 3 (3 is the site specific applicable factor for the mixing effect of the BV-1 and BV-2 discharge structure). Since all liquid release pathways lead to the river, it can be assumed that all liquid waste flow mixes with flow from the site discharge structure and can include the factor of 3 in the near field average dilution factor.
= Waste Flow (3) (Dilution Water Flow)
The site specific applicable factor of 3 results in a conservative estimate of the near field dilution factor based upon Regulatory Guide 1.113<3-l.3.4) methodology and is a factor of 10 below the limit specified in NUREG-0133, Section 4.3. <3.1.3.I)
The dose factor Ai, was calculated for an adult for each isotope using the following equation from NUREG-0133 <3.u.1)_
Ah= l.14E5 (730/Dw + 21BFi)DFi, [1.3-2]
where:
l.l 4E 5 = [ 1E6 pCi] x [ 1E3 ml] x [ lyr ]
uCi liter 8760 hr 730 = Adult water consumption rate (liters/yr).
Dw = Far field dilution factor from the near field area within 1/4 mile of the release point to the potable water intake for adult water consumption.
21 = Adult fish consumption (kg/yr).
BFi = Bioaccumulation factor for radionuclide "i" in fish from Table A-1 of Regulatory Guide 1.109<3.u.5) (pCi/kg per pCi/1). However, if data was not available from that reference, it was obtained from Table 6 of UCRL-50564.<3 *13 *12>
The bioaccumulation factor for niobium (300 pCi/kg per pCi/1) was not obtained from either of the above references noted. It was obtained from IAEA Safety Series No. 57. Justification for use of this value is documented in Appendix A to Calculation Package No. ERS-ATL-83-027. <3.u.6)
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ODCM: LIQUID EFFLUENTS 17 24 of 43 DFi, = Dose conversion factor for radionuclide "i" for adults for a particular organ t (mrem/pCi) from Table E-11 of Regulatory Guide 1.109, (3.n5) ERS-LMR-15-006, (3 .u.9) orNUREG-0172. (3.1.3.ll)
A table of Ai, values for an adult at BV-1 and BV-2 are presented in ATTACHMENT B Table 1.3-1.
The far field dilution factor (Dw) for BV-1 and BV-2 is 200. This value is based on a total dilution factor of 600 applicable to the Midland water intake located 1.3 miles downstream and on the opposite bank from BV-1 and BV-2 (i.e., 200 = 600 + 3). The total dilution factor of 600 represents a conservative fully mixed annual average condition. Since the Midland intake is located on the opposite bank and is below the water surface, essentially fully mixed conditions would have to exist for the radioactive effluent to be transported to the intake.
The cumulative doses (from each reactor unit) for a calendar quarter and a calendar year are compared to ODCM CONTROL 3.11.1.2 as follows:
For the calendar quarter, D, < 1.5 mrem total body [1.3-3]
D, < 5 mrem any organ [1.3-4]
For the calendar year, D, < 3 mrem total body [1.3-5]
D, < 10 mrem any organ [l .3-6]
If any of the limits in Equation [1.3-3] through [1.3-6] are exceeded, a Special Report pursuant to ODCM Control 3.11.1.2 of 1/2-0DC-3103 is required. (3 .n 16) 8.3.2 Proiection Of Doses {ODCM CONTROL 3.11.1.3)
Doses due to liquid releases shall be projected at least once per 31 days in accordance with ODCM CONTROL 3 .11.1.3 and this section. The Liquid Radwaste Treatment System shall be used to reduce the radioactive materials in each liquid waste batch prior to its discharge, when the projected doses due to liquid effluent releases from each reactor unit, when averaged over 3 l days would exceed 0.06 mrem to the total body or 0.2 mrem to any organ. Doses used in the projection are obtained according to equation [1.3-1]. The 31-day dose projection shall be performed according to the following equations:
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17 25 of 43 When including pre-relea se data, D 31 A+BJ
= [ -T- 31+C [13-7]
When not including pre-release data,
[1.3-8]
where:
1 D31 = Projected 31 day dose (mrem).
A = Cumulative dose for quarter (mrem).
B = Projected dose from this release (mrem).
T = Current days into quarter.
C = Value which may be used to anticipate plant trends (mrem).
8.4 Liquid Radwaste System The liquid radwaste system has the capability to control, collect, process, store, recycle, and dispose of liquid radioactive waste generated as a result of plant operations, including anticipated operational occurrences. This system also uses some of the components of the steam generator blowdown system for processing.
Simplified flow diagrams of the liquid radwaste systems for BV-1 and BV-2 are provided as ATTACHMENT C Figures 1.4-1 and 1.4-2 respectively. A diagram showing the liquid effluent release points is provided as ATTACHMENT C Figure 1.4-3 .. A diagram of the site boundary for liquid effluents is provided as ATTACHMENT D Figure 5-1.
Since the concept of a shared liquid radwaste system is used, then any liquid waste generated can be stored, processed and discharged from either BV-1 or BV-2.
8.4.1 BV-1 Liquid Radwaste System Components 8.4.1.1 [lBR-I-lA/B], CESIUM REMOVAL ION EXCHANGERS There are two (2) of these ion exchangers, each has a capacity of thirty-five (35) cubic feet. They are located on the east side of the Auxiliary Building (elevation 735'). They receive process fluid (liquid waste) from the reactor coolant system when letdown flow is diverted from the volume control tank.
8.4.1.2 [1BR-TK-4A/B], COOLANT RECOVERY TANKS There are two (2) of these tanks, each tank has a nominal capacity of 195,000 gallons (maximum capacity= 205,578 gallons). They are located in the Solid Waste Building. They receive diverted letdown flow from the volume control tank and various reactor plant non-aerated drains that were processed through the [lBR-I-
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ODCM: LIQUID EFFLUENTS 17 26 of 43 lA/B], CESIUM REMOVAL ION EXCHANGERS from both Unit 1 and Unit 2.
These tanks can also receive unprocessed liquid wastes from either Unit 1 or Unit 2 liquid waste systems. Normally, one (1) tank receives liquid waste while the other tank is placed on recirculation through the demineralizer until the radioactivity concentration is acceptable for discharge. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
8.4.1.3 [1LW-TK-2A/B], HIGH LEVEL WASTE DRAIN TANKS There are two (2) of these tanks, each tank has a nominal capacity of 5,000 gallons (maximum capacity= 4,899 gallons). They are located on the northwest wall of the Auxiliary Building (elevation 735'). They receive liquid wastes from the vent and drain system.
8.4.1.4 [1LW-TK-3A/B], LOW LEVEL WASTE DRAIN TANKS There are two (2) of these tanks, each tank has a nominal capacity of 2,000 gallons (maximum capacity = 1,998 gallons). They are located in the northwest comer of the Auxiliary Building (elevation 735'). They receive liquid wastes from the vent and drain system and can be transferred directly to [1BR-TK-4A/B], COOLANT RECOVERY TANKS. Although not normally used, these tanks can also be utilized to discharge processed or unprocessed liquid wastes. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
8.4.1.5 [lLW-FL-7], LIQUID WASTE PRE-CONDITIONING FILTER A pre-conditioning filter with a fifty (50) is designed to clean liquid waste water of particulate and dissolved radioactive contaminants that is stored in [1LW-TK-2A/B],
HIGH LEVEL WASTE DRAIN TANKS; [1LW-TK-3A/B], LOW LEVEL WASTE DRAIN TANKS; and [1BR-TK-4A/B], COOLANT RECOVERY TANKS. The pre-conditioning filter can be charged with varying grades of activated charcoal (carbon) intended for removal of radionuclides in a colloidal state. The charcoal may consist of course mesh high activated coco carbon, medium mesh high activated coco carbon, fine mesh high activated coco carbon and cobalt selective media. This filter is located in the Solid Waste Building (elevation 735'-6").
8.4.1.6 [lLW-I-3], LIQUID WASTE DEMINERALIZER There are two (2) demineralizer 36" diameter vessels [lLW-I-3-1, lLW-I-3-2] each with a capacity of thirty (30) cubic feet and two (2) demineralizer 24" diameter vessels [lLW-I-3-3, lLW-I-3-4] each with a capacity of fifteen (15) cubic feet. The demineralizers are designed to clean liquid waste water of particulate and dissolved radioactive contaminants and pre-conditioned by [lLW-FL-7]. The primary ion exchange occurs in [lLW-3-3, lLW-I-3-4] exchange vessels. These vessels are located at the end of the process train to maximize the cleaning effect of the media.
[lLW-I-3-1, lLW-I-3-2] "accumulator" vessels are placed at the front of the influent liquid waste demineralizer line. These vessels are used as ion exchange process vessels, however they are normally used for holding partially depleted resin sluiced
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17 27 of 43
\
in from the primary ion exchangers [lLW-1-3-3, lLW-1-3-4]. Each of the demineralizer vessels may be charged with different resins for effective removal of chemical and radioactive contaminants. Resin selection and sequence may be changed dependent upon site liquid waste processing needs. This demineralizer is located in the Solid Waste Building (elevation 735'-6").
8.4.1.7 Liquid Waste Evaporator An evaporator was originally designed to process liquid waste at Unit 1 with a capacity of six (6) gpm. However, this evaporator was retired prior to initial issue of the ODCM, because of concerns for creating a mixed-waste. SINCE the evaporator is no longer in-use, THEN it is not shown on Figure 1.4-1 in ATTACHMENT C.
8.4.1.8 [1LW-TK-7A/B], STEAM GENERATOR DRAIN TANKS There are two (2) of these tanks, each tank has a nominal capacity of 34,500 gallons (maximum capacity= 35,800 gallons). They are located in the Fuel Pool Leakage Monitoring Room (elevation 735'). They normally rec~ive liquid waste that has been processed through the liquid waste demineralizer. These tanks can also receive liquid waste from Unit 2. Upon completion of filling operation, the tank is placed on recirculation through the demineralizer until the radioactivity concentration is acceptable for discharge. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
8.4.1.9 [RM-lLW-104], LIQUID WASTE DISCHARGE RADIATION MONITOR An off-line gamma scintillator radiation monitor continuously analyzes liquid waste as it is being discharged. The normal rate of discharge through this radiation monitor from [1LW-TK-3A/B], LOW LEVEL WASTE TANK DRAIN TANKS and [lLW-TK-7A/B] is less than thirty-five (35) gpm. The normal rate of discharge through this radiation monitor from [1BR-TK-4A/B], COOLANT RECOVERY TANKS is less than 50 gpm. The high alarm on this radiation monitor has a setpoint that would indicate liquid waste discharges that are approaching OEC limits. If a high high alarm is received, liquid waste discharge is automatically terminated by closing the discharge line isolation valve.
8.4.1.10 [lLW-FL-8], RAD WASTE REVERSE OSMOSIS (RWRO) PRE-FILTER A pleated paper mechanical pre-filter upstream of [lLW-RWR0-1], RAD WASTE REVERSE OSMOSIS SKID is designed to prevent fouling of the system by buildup of suspended solids in the reverse osmosis membranes. The shielded pleated-paper filter is sized to filter particulate down to 0.3 microns. The filter is contained within a lockable, shielded cover that allows access for filter replacement. This filter is located in the Solid Waste Building (elevation 735'-6").
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ODCM: LIQUID EFFLUENTS 17 28 of 43 8.4.1.11 [lLW-RWR0-1], RAD WASTE REVERSE OSMOSIS SKID adds a physical membrane barrier to the chemical ion exchange system used in the current liquid waste demineralizer system. This replaceable membrane barrier prevents radwaste particles that were not captured during the ion exchange process. [lLW-RWR0-1] does not capture the particles, but instead produces a reject stream separate from the permeate stream. The [lLW-RWR0-1] accepts the effluent stream of the demineralizer system [lLW-1-3], LIQUID WASTE DEMINERALIZER and is designed to allow boron and silica particles to pass through while rejecting others such as antimony. If antimony is present in the effluent stream it is separated and is sent to [lLW-I-3-5], Antimony Vessel charged with antimony selective resin.
The antimony vessel removes the isotope from the reject stream prior to discharge back to the [1BR-TK-4A/B], COOLANT RECOVERY TANK being processed from. This skid is located in the Solid Waste Building (elevation 735'-6").
8.4.1.12 [lLW-I-3-5], ANTIMONY VESSEL
[lLW-I-3-5], ANTIMONY VESSEL is a two (2) cubic foot ion exchange vessel placed downstream of the [lLW-RWR0-1], RAD WASTE REVERSE OSMOSIS SKID. This vessel removes antimony isotopes from the reject stream of the [lLW-I-3-5] prior to discharge back to the [1BR-TK-4A/B], COOLANT RECOVERY TANK being processed from. This skid is located in the Solid Waste Building (elevation 735'-6").
8.4.1.13 [lLW-I-3-6], POLISHING VESSEL
[lLW-I-3-6], POLISHING VESSEL is a fifteen (15) cubic foot ion exchange vessel placed downstream of the [lLW-RWR0-1], RAD WASTE REVERSE OSMOSIS SKID. This skid is located in the Solid Waste Building (elevation 735'-6").
8.4.1.14 Sample Sinks The sample sinks allow sampling at both influent and effluent vessel streams. In addition, the sinks include gauges that indicate the pressure at each sample point.
One sample sink is located locally (Solid Waste Building - elevation 735'-6") to the system and allows for sampling from eight (8) individual points within the system.
A second sample sink is remotely located and contains four (4) sampling points (Primary Auxiliary Building - elevation 768'-7" along the east wall). A sample sink is also included on the [lLW-RWR0-1], RAD WASTE REVERSE OSMOSIS (RWRO) SKID to provide sample points from within the skid itself.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 29 of 43 8.4.2 BV-1 Laundry and Contaminated Shower Drain System Components 8.4.2.1 [1LW-TK-6A/B], LAUNDRY AND CONTAMINATED SHOWER DRAIN TANKS There are two (2) of these tanks, each has a nominal capacity of 1,200 gallons (maximum capacity= 1,303 gallons). They are located in the northwest comer of the Auxiliary Building (elevation 722'). They receive laundry and contaminated shower drains waste from the Service Building. These tanks can also receive mop water waste and other low-level contaminated waste from Unit 2. The waste in these tanks is not sent to the liquid waste demineralizer for cleanup because this waste may contain organic compounds that will deplete a resin bed. Upon completion of filling operation, the tank must be recirculated a minimum of two (2) tank volumes prior to sampling for discharge permit preparation.
8.4.2.2 [RM-ILW-116], LAUNDRY AND CONTAMINATED SHOWER DRAINS TANK DISCHARGE RADIATION MONITOR An off-line gamma scintillator radiation monitor continuously analyzes laundry and contaminated shower drains waste as it is being discharged. The normal rate of discharge through this radiation monitor from [ILW-TK-6A/B], LAUNDRY AND CONTAMINATED SHOWER DRAIN is less than fifteen (15) gpm. The high alarm on this radiation monitor has a setpoint that would indicate liquid waste discharges are approaching OEC limits. If a high high alarm is received, liquid waste discharge is automatically terminated by closing the discharge line isolation valve.
8.4.3 BV-2 Liquid Radwaste System Components 8.4.3.1 [2BRS-IOE2IA/B], CESIUM REMOVAL ION EXCHANGERS There are two (2) of these ion exchangers, each has a capacity ofthirty-five-(35) cubic feet. They are located on the east side of the Auxiliary Building (elevation 718'). They receive and process liquid wastes from the reactor coolant system during dilution or letdown operations.
8.4.3.2 [2LWS-TK2IA/B], WASTE DRAIN TANKS There are two (2) of these tanks, each tank has a nominal capacity of 10,000 gallons (maximum capacity= 10,184 gallons). They are located in the northeast comer of the Auxiliary Building (elevation 710'). They receive liquid wastes from the vent and drain system. These tanks can also receive liquid wastes from Unit 1. IF further processing is not necessary, then it may be placed on recirculation. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level OfUse:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 30 of 43 8.4.3.3 [2SGC-I0E21A/BSTEAM GENERATORBLOWDOW N CLEANUP ION EXCHANGERS The main purpose of the ion exchangers is to clean liquid waste water of particulate and dissolved radioactive contaminants through an ion exchange process. There is a resin bed, outlets strainer, and cleanup filter associated with each of these ion exchangers. They are located in the Waste Handling Building (elevation 722').
8.4.3.4 Liquid Waste Evaporator Two (2) evaporators were originally designed to process liquid waste at Unit 2 with a capacity of twenty (20) gpm each. However, these evaporators were retired prior to initial issue of the ODCM, because of concerns for creating a mixed-waste.
Since the evaporators are no longer in-use then they are not shown on Figure 1.4-2 in ATTACHMENT C.
8.4.3.5 [2SGC-TK23A/B], STEAM GENERATOR BLOWDOWN TEST TANKS There are two (2) of these tanks, each has a nominal capacity of 18,000 gallons (maximum capacity= 17,955 gallons). They are located in the Auxiliary Building (elevation 755'). They receive liquid waste that has been processed through the cleanup ion exchangers. Upon completion of filling operation, the tank is placed on recirculation through the demineralizer until the radioactivity concentration is acceptable for discharge. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
8.4.3.6 [2SGC-TK21A/B], STEAM GENERATOR BLOWDOWN HOLD TANKS There are two (2) of these tanks, each has a nominal capacity of 50,000 gallons (maximum capacity= 51,460 gallons). They are located in the Waste Handling Building (elevation 722'). These tanks are used to store liquid waste when the radioactive concentration of the steam generator blowdown test tank is not acceptable for discharge. These tanks can also receive liquid wastes from Unit 1.
The contents of this tank may be drained or processed through the Unit 1 or Unit 2 Liquid Radwaste Treatment System until the radioactivity concentration is acceptable for discharge. A minimum of two (2) tank volumes must be recirculated prior to sampling for discharge permit preparation.
8.4.3.7 [2SGC-RQ100], LIQUID WASTE EFFLUENT MONITOR A off-line gamma scintillator radiation monitor continuously analyzes liquid waste as it is being discharged. The normal rate of discharge through this radiation monitor is less than eighty-five (85) gpm. The alert alarm on this radiation monitor has a setpoint that would indicate liquid waste discharges are approaching OEC limits. If a high alarm is received, liquid waste discharge is automatically terminated by closing the discharge line isolation valve.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 31 of 43 8.4.4 BV-1/2 Miscellaneous Sumps 8.4.4.1 Unit 1 Chemical Waste Sump This sump has an approximate capacity of 16,755 gallons and the associated trenches add an additional volume of approximately 5,140 gallons. The waste in this sump is not processed through any liquid waste demineralizer for cleanup.
This sump cannot be completely isolated and does not have any radiation monitors. The sump discharges to Unit 1 Liquid Waste Line that discharges to the Unit 1 Cooling Tower Blowdown Line.
The Chemical Waste Sump typically does not receive radiological liquid waste.
When primary to secondary leakage is greater than 0.1 gpm (142 gpd) releases of radioactive material from this sump are considered to be continuous liquid effluent discharge as specified in ODCM Control 3.11.1.1. When the Turbine Building Sump concentration exceeds 1 OEC, the Chemical Waste Sump accepts flow from the Turbine Building Sump(s).
Prior to discharge the Chemical Waste Sump is sampled. Individual liquid effluent batch discharge permits are not required if no Licensed Radioactive Material (LRM) is present and the discharge of tritium is accounted for with a monthly secondary diffusion permit. If LRM is detected in the Chemical Waste Sump, then a special permit is required. Upon completion of filling operation, the sump must be recirculated a minimum of two (2) volumes prior to sampling for batch discharge permit preparation.
During normal plant operation, this sump may be used for the treatment and disposal of radiological materials for various projects at the discretion of site management. When Licensed Radioactive Material from a source other than a Turbine Building drain of secondary water is added to the sump, the sump may be discharged using batch release methods if concentration and dose limits are maintained.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 32 of 43 8.4.4.2 Unit 2 Cable Vault Sump This sump has a capacity of 2,424 gallons, however the high level alarm will actuate at 1,500 gallons. This sump cannot be completely isolated and does not have any radiation monitors. The waste in this sump is not normally processed through a liquid waste demineralizer for cleanup because the sump may contain contaminants that would deplete the resin bed. The sump normally receives non-radiological liquid waste. It is used for the collection and release of non-contaminated water. The open pit design provides the potential for radiological materials to enter the sump due to its location in the Radiological Controlled Area thus requiring Offsite Dose Calculation Manual (ODCM) controls prior to discharge.
Prior to discharge the Cable Vault Sump is sampled. Licensed Radioactive Material entering the sump may be discharged using batch release methods if concentration and dose limits are maintained. Typically, tritium is accounted for in monthly pre-release batch liquid effluent discharge permit. This sump discharges into the Unit 2 Catch Basin System at 2CB-4. Upon completion of filling operation, the sump must be recirculated a minimum of two (2) volumes prior to sampling for discharge permit preparation.
-END-
Procedure Number:
Beaver Valley Power Station 1/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 33 of 41 ATTACHMENT A Page 1 of 4 LIQUID SOURCE TERMS TABLE I.I-la BV-1 LIQUID SOURCE TERM Ei <4)
Ai <2) DETECTION ANNUAL RELEASE OEC <3l EFFICIENCY NUCLIDE (Ci) (uCi/mL) (cpm/uCi/mL)
Cr-51 l.3E-3 5E-3 l.35E+7 Mn-54 3.IE-4 3E-4 l.03E+8 Fe-55 l.6E-3 IE-3 6.36E-1 Fe-59 8.3E-4 IE-4 9.98E+7 Co-58 l.4E-2 2E-4 1.40E+8 Co-60 2.0E-3 3E-5 l.84E+8 Zn-65 <3.u.1 7) 2.69E-2 5E-5 5.I0E+7 Np-239 1.4E-4 2E-4 8.01E+7 Br-83 2.5E-5 9E-3 l.65E+6 Br-84 2.7E-6 4E-3 l.01E+8 Br-85 2.SE-8 --- 6.60E+6 Rb-86 7.5E-5 7E-5. 8.30E+6 Sr-89 2.9E-4 SE-5 9.64E+3 Sr-90 1.IE-5 5E-6 (5)
Y-90 9.4E-6 7E-5 1.10E+2 Y-9Im 8.7E-6 2E-2 1.10E+8 Y-91 5.7E-5 SE-5 2.42E+5 I
Y-93 7.4E-7 2E-4 1.55E+7 Zr-95 5.IE-5 2E-4 l.06E+8 Nb-95 5.2E-5 3E-4 1.06E+8 Sr-91 l.3E-5 2E-4 8.32E+7 Mo-99 l.lE-2 2E-4 2.91E+7 Tc-99m l.IE-2 lE-2 1.18E+8 Ru-103 3.4E-5 3E-4 1.17E+8 Ru-106 l.OE-5 3E-5 2.74E+7 Rh-103m 3.4E-5 6E-2 (5)
Rh-106 l.OE-5 --- 3.87E+7 Te-125m 2.5E-5 2E-4 l.73E+5 Te-127m 2.6E-4 9E-5 1.69E+4 Te-127 2.7E-4 lE-3 1.58E+6 Te-129m l.IE-3 7E-5 4.41E+6 Te-129 6.7E-4 4E-3 l.43E+7 1-130 1.2E-4 2E-4 3.68E+8 Te-131m l.6E-4 SE-5 1.80E+8 Te-131 3.0E-5 SE-4 1.43E+8 1-131 1.6E-1 lE-5 l.27E+8
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 34 of 41 ATTACHMENT A Page 2 of 4 LIQUID SOURCE TERMS TABLE 1.1-1 a (continued)
BV-1 LIQUID SOURCE TERM Ei <4)
Ai <2) DETECTION ANNUAL RELEASE OECi <3) EFFICIENCY NUCLIDE (Ci) (uCi/mL) (cpm/uCi/mL)
Te-132 4.3E-3 9E-5 l.34E+8 I-132 4.9E-3 lE-3 3.16E+8 I-133 4.0E-2 7E-5 l.21E+8 I-134 8.0E-5 4E-3 2.99E+8 Cs-134 4.6E-2 9E-6 2.44E+8 I-135 4.3E-3 3E-4 l.29E+8 Cs-136 8.9E-3 6E-5 3.05E+8 Cs-137 3.3E-2 lE-5 9.40E+7 Ba-13711J 3.lE-2 -- 9.93E+7 Ba-140 1.IE-4 8E-5 5.05E+7 La-140 l.lE-4 9E-5 l.91E+8 Ce-141 5.IE-5 3E-4 6.72E+7 Ce-143 2.8E-6 2E-4 8.26E+7 Ce-144 3.2E-5 3E-5 l.36E+7 Pr-143 2.7E-5 2E-4 l.28E+O Pr-144 3.2E-5 6E-3 2.27E+6 H-3 5.50E+2 lE-2 (5)
TOTAL (I} 4.05E-1
(!)
Excluding Tritium and Entrained Noble Gases 6
(2) Source Term for (RM-lLW-104 and RM-lLW-116) from Stone and Webster Calculation Package UR(B)-160 (3.1.1. )
3 (3) ODCM Effluent Concentration Limit = 10 times the EC values of 10 CFR 20 < . u. io) 4 (4)
Detection Efficiency for (RM-lLW-104 and RM-lLW-116) from Calculation Package ERS-SFL-92-039 <3.1.1. )
(5)
Insignificant f'.
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 35 of 43 ATTACHMENT A Page 3 of 4 LIQUID SOURCE TERMS TABLE 1.1-lb BV-2 LIQUID SOURCE TERM Ei <4>
Ai <2> DETECTION ANNUAL RELEASE OECi <3> EFFICIENCY NUCLIDE (Ci) (uCi/mL) (cpm/uCi/mL)
Cr-51 l.OOE-4 SE-3 2.01E+7 Mn-54 2.SOE-5 3E-4 1.27E+8 Fe-55 1.30E-4 lE-3 (5)
Fe-59 6.50E-5 lE-4 l.26E+8 Co-58 1.IOE-3 2E-4 l.82E+8 Co-60 l.60E-4 3E-5 2.38E+8 Zn-65 (3.ni 7) 5.IOE-2 5E-5 6.50E+7 Np-239 3.20E-5 2E-4 l.65E+8 Br-83 2.90E-5 9E-3 2.42E+6 Br-84 5.90E-9 4E-3 l.38E+8 Rb-86 3.70E-5 7E-5 l.04E+7 Sr-89 2.20E-5 8E-5 l.83E+4 Sr-90 8.50E-7 5E-6 (5)
Sr-91 5.30E-6 2E-4 1.04E+8 Mo-99 2.30E-3 2E-4 4.47E+7 Tc-99m 2.IOE-3 lE-2 1.40E+8 Te-125m l.90E-6 2E-4 3.94E+5 Te-127m 2.IOE-5 9E-5 1.26E+5 Te-127 2.50E-5 lE-3 2.43E+6 Te-129m 8.20E-5 7E-5 6.53E+6 Te-129 5.30E-5 4E-3 l.96E+7 1-130 2.30E-4 2E-4 5.18E+8 Te-131m 5.20E-5 8E-5 2.85E+8 Te-131 9.40E-6 8E-4 1.88E+8 1-131 l.OOE-1 lE-5 l.96E+8 Te-132 7.80E-4 9E-5 1.76E+8 1-132 2.30E-3 lE-3 4.22E+8 1-133 6.50E-2 7E-5 1.73E+8 1-134 4.60E-6 4E-3 4.06E+8 Cs-134 3.00E-2 9E-6 3.25E+8 1-135 9.20E-3 3E-4 1.71E+8 Cs-136 3.90E-3 6E-5 4.28E+8 Cs-137 2.20E-2 lE-5 l.28E+8 Ba-137m 2.lOE-2 --- l.33E+8 Ba-140 9.30E-6 8E-5 7.50E+7 La-140 8.40E-6 9E-5 3.08E+8
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 36 of 41 ATTACHMENT A Page 4 of 4 LIQUID SOURCE TERMS TABLE 1.1-1 b (continued)
BV-2 LIQUID SOURCE TERM Ei <4)
Ai <2) DETECTION ANNUAL RELEASE OEC <3) EFFICIENCY NUCLIDE (Ci) \ (uCi/mL) (cpm/uCi/mL)
Y-91m 3.60E-6 2E-2 l.59E+8 Y-91 4.40E-6 8E-5 3.55E+5 Y-93 3.00E-7 2E-4 2.03E+7 Zr-95 4.00E-6 2E-4 l.35E+8 Nb-95 4.00E-6 3E-4 l.33E+8 Ru-103 2.70E-6 3E-4 l.71E+8 Ru-106 8.20E-7 3E-5 (5)
Rh-103m 2.70E-6 6E-2 (5)
Rh-106 8.20E-7 --- 5.65E+7 Ce-141 4.00E-6 3E-4 7.75E+7 Ce-143 8.60E-7 2E-4 1.20E+8 Ce-144 2.60E-6 3E-5 l.87E+7 Pr-143 2.30E-6 2E-4 1.63E+O Pr-144 2.60E-6 6E-3 3.40E+6 H-3 5.50E+2 lE-2 (5)*
TOTAL(!) 2.40E-1 (I) Excluding Tritium and Entrained Noble Gases (2) Source Term for (2SGC-RQ100) from Computer Code LIQlBB <3.1. 2.3)
(3) ODCM Effluent Concentration Limit= 10 times the EC values of 10 CPR 20 (3.u.io)
(4) Detection Efficiency for (2SGC-RQ100) from Calculation Package ERS-SFL-86-026 (3 .1. 2 -2) -
(5)
Insignificant i
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 37 of 43 ATTACHMENT B Page I of 3 INGESTION DOSE COMMITMENT FACTORS TABLE 1.3-1 Ait VALVES FOR THE ADULT FOR THE BEAVER VALLEY SITE (mrem/hr per uCi/mL)
NUCLIDE BONE LIVER T-BODY THYROID KIDNEY LUNG GI-ILi H-3 O.OOEOO 2.70E-0I 2.70E-0I 2.70E-01 2.70E-0I 2.70E-0I 2.70E-0I Be-7 l.43E-02 314E-02 l.59E-02 O.OOEOO 3.42E-02 O.OOEOO 5.62EOO C-14 3.13E04 616£03 616£03 616£03 616£03 616£03 616£03 Na-24 4.08E02 4.08E02 4.08E02 4.08E02 4.08E02 4.08E02 4.08E02 P-32 4.62E07 2.87E06 l.79E06 O.OOEOO O.OOEOO O.OOEOO 5.19E06 Cr-51 O.OOEOO O.OOEOO I17EOO 7.62E-OI 2.81E-0I l.69EOO 311E02 J
Mn-54 O.OOEOO 4.38E03 8.35E02 O.OOEOO I.30E03 O.OOEOO l.34E04 Mn-56 O.OOEOO l.10E02 l.95E01 O.OOEOO I.40E02 O.OOEOO 3.52E03 Fe-55 6.59E02 4.56E02 I.06E02 O.OOEOO O.OOEOO 2.54E02 2.61E02 Fe-59 I.04E03 2.45E03 9.38E02 O.OOEOO O.OOEOO 6.83E02 8.15E03 ili-57 O.OOEOO 2.IOEOl 3.50E0I O.OOEOO O.OOEOO O.OOEOO 5.33E02 ili-58 O.OOEOO 8.95E01 2.01E02 O.OOEOO O.OOEOO O.OOEOO l.81E03 Co-60 O.OOEOO 2.57E02 5.67E02 O.OOEOO O.OOEOO O.OOEOO 4.83E03 Ni-63 3.12E04 2.16E03 l.05E03 O.OOEOO O.OOEOO O.OOEOO 4.51E02 Ni-65 I17E02 l.65E0I 7.51EOO O.OOEOO O.OOEOO O.OOEOO 4.17E02 Cu-64 O.OOEOO l.OOEOl 4.70EOO O.OOEOO 2.52EOI O.OOEOO 8.53E02 Zn-65 2.32E04 7.37E04 3.33E04 O.OOEOO 4.93E04 O.OOEOO 4.64E04 Zn-69 4.93E0I 9.43E0I 6.56EOO O.OOEOO 6.13EOI O.OOEOO l.42E01 Se-75 O.OOEOO l.86E03 712E02 O.OOEOO 2.70E03 O.OOEOO 2.69E02 Br-83 O.OOEOO O.OOEOO 4.04EOI O.OOEOO O.OOEOO O.OOEOO 5.82E01 Br-84 O.OOEOO O.OOEOO 514EOI O.OOEOO 0.00EOO O.OOEOO 4.IIE-04 Br-85 O.OOEOO O.OOEOO 2.15EOO O.OOEOO O.OOEOO O.OOEOO O.OOEOO Rb-86 O.OOEOO l.01E05 4.71E04 O.OOEOO O.OOEOO O.OOEOO l.99E04 Rb O.OOEOO 2.90E02 1.54E02 O.OOEOO O.OOEOO O.OOEOO 4.00E-09 Rb-89 O.OOEOO I.92E02 l.35E02 O.OOEOO O.OOEOO O.OOEOO l.12E-11 Sr-89 212E04 O.OOEOO 6.39E02 O.OOEOO O.OOEOO O.OOEOO 3.57E03 Sr-90 5.48E05 O.OOEOO 1.34E05 O.OOEOO O.OOEOO O.OOEOO l.58E04
~
Sr-91 4.10£02 O.OOEOO l.65EOI O.OOEOO 0.00EOO O.OOEOO l.95E03 Sr-92 l.55E02 O.OOEOO 6.72EOO 0.00EOO O.OOEOO O.OOEOO 3.08E03 Y-90 5.80E-01 O.OOEOO l.55E-02 O.OOEOO O.OOEOO O.OOEOO 6.15E03 Y-91m 5.48E-03 O.OOEOO 2.12E-04 O.OOEOO O.OOEOO O.OOEOO l.61E-02 Y-91 8.50EOO O.OOEOO 227E-01 O.OOEOO O.OOEOO O.OOEOO 4.68E03 Y-92 5.09E-02 O.OOEOO l.49E-03 O.OOEOO O.OOEOO O.OOEOO 8.92E02
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 38 of 43
,ATTACHMENT B Page 2 of 3 INGESTION DOSE COMMITMENT FACTORS Ait VALUES FOR THE ADULT FOR THE BEAVER VALLEY SITE (mrem/hr per uCi/mL)
NUCT.IDE. BONE LIVER T-BODY 1HYR0ID KIDNEY LUNG GI-ILi Y-93 1.62&01 O.OOEOO 4.46E-03 O.OOEOO O.OOEOO O.OOEOO 5.12E03 Zr-95 2.53E-01 8.llE-02 5.49E-02 O.OOEOO l27E-Ol O.OOEOO 257E02 Zr-97 l.40E-02 2.82E-03 l29E-03 O.OOEOO 426E-03 O.OOEOO 8.73E02 Nb-95 4.47EOO 2.49EOO l.34EOO O.OOEOO 2.46EOO O.OOEOO l.5IE04 Nb-97 3.75E-02 9.49E-03 3.46E-03 O.OOEOO 1.llE-02 O.OOEOO 3.50E01 Mo-99 O.OOEOO l.05E02 2.00EOl O.OOEOO 238E02 O.OOEOO 2.43E02 Tc-99m 8.97E-03 2.54E-02 323E-Ol O.OOEOO 3.85E-01 l24E-02 l.50E01 Tc-101 923E-03 133E-02 lJOE-01 O.OOEOO 239E-01 6.79E-03 4.00E-14 Ru-103 4.5IEOO O.OOEOO l.94EOO O.OOEOO l.72E01 O.OOEOO 526E02 Ru-105 3.75E-01 O.OOEOO l.48E-01 O.OOEOO 4.85EOO O.OOEOO 229E02 Ru-106 6.70E01 0.00EOO 8.48EOO O.OOEOO l29E02 O.OOEOO 4.34E03 Rh-103m 4.5IEOO 0.00EOO 1.94EOO O.OOEOO l.72E01 0.00EOO 526E02 Rh-105 2.95EOO 2.16EOO 1.42EOO O.OOEOO 9.16EOO O.OOEOO 3.43E02 Rh-106 6.70E01 O.OOEOO 8.48EOO O.OOEOO l29E02 O.OOEOO 434E03 Ag-llOm 9.48E-01 8.77E-Ol 521E-01 O.OOEOO l.72EOO O.OOEOO 3.58E02 Sn-113 5.60E04 l.57E03 322E03 927E02 O.OOEOO O.OOEOO l.57E05 Sn-117m 3.84E03 3.73E02 9.48E02 2.74E02 O.OOEOO O.OOEOO l.62E05 Sb-122 6.55E-Ol l28E-02 l.90E-01 9.08E-03 O.OOEOO 3.40E-01 l.86E02 Sb-124 7.87EOO l.49E-01 3.12EOO 1.9IE-02 O.OOEOO 6.13EOO 223E02 Sb-125 5.03EOO 5.62E-02 120EOO 5.llE-03 O.OOEOO 3.88EOO 5.54E01 Sb-126 323EOO 6.58E-02 l.17EOO l.98E-02 O.OOEOO l.98EOO 2.64E02 Sb-127 725E-Ol 1.59E-02 278E-01 8.71E-03 O.OOEOO 4.30E-01 l.66E02 Te-125m 2.57E03 930E02 3.44E02 7.72E02 l.04E04 O.OOEOO l.03E04 Te-127m . 6.49E03 2.32E03 7.90E02 l.66E03 2.63E04 O.OOEOO 2.17E04 Te-127 l.05E02 3.78E01 228E01 7.8IE01 429E02 O.OOEOO 832E03 Te-129m l.10E04 4.1IE03 1.74E03 3.78E03 4.60E04 O.OOEOO 5.55E04 Te-129 3.0lEOl l.13E01 733EOO 2.3IE01 l26E02 O.OOEOO 227E01 Te-131m l.66E03 8.10E02 6.75E02 128E03 82IE03 O.OOEOO 8.05E04 Te-131 l.89E01 7.88EOO 5.96EOO l.55E01 827EOI O.OOEOO 2.67EOO Te-132 2.4IE03 l.56E03 l.47E03 l.72E03 1.50E04 O.OOEOO 739E04 Te-134 3.lOEOl 2.03E01 l25EOI 2.7IE01 l.96E02 O.OOEOO 3.44E-02 1-129 l.19E02 l.02E02 3.35E02 2.63E05 2.19E02 O.OOEOO l.6IE01 1-130 2.75E01 8.lOEOl 320E01 6.87E03 l26E02 O.OOEOO 6.97E01
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 39 of 43 ATTACHMENT B Page 3 of 3 INGESTION DOSE COMMITMENT FACTORS Ait VALUES FOR THE ADULT FOR THE BEAVER VALLEY SITE (mrem/hr per uCi/mL)
NUCLIDE BONE LIVER T-BODY 1HYROID KIDNEY LUNG GI-LU I-131 L51E02 2.16E02 124E02 7.08E04 3.71E02 O.OOEOO 5.70E01 1-132 7.37EOO l.97E01 6.90EOO 6.90E02 3.14E01 O.OOEOO 3.71EOO 1-133 5.16E01 8.97E01 2.74E01 1.32E04 l.57E02 O.OOEOO 8.06E01 1-134 3.85EOO l.05E01 3.74EOO l.81E02 l.66E01 O.OOEOO 9.12E-03 I-135 l.61E01 421E01 l.55E01 2.78E03 6.76E01 O.OOEOO 4.76E01 Cs-134 2.98E05 7.09E05 5.79E05 O.OOEOO 229E05 7.61E04 124E04 Cs-136 3.12E04 123E05 8.86E04 O.OOEOO 6.85E04 9.39E03 l.40E04 Cs-137 3.82E05 5.22E05 3.42E05 O.OOEOO l.77E05 5.89E04 l.01E04 Cs-138 2.64E02 5.22E02 2.59E02 O.OOEOO 3.84E02 3.79E01 223E-03 Ba-139 9.69E-01 6.90E-04 2.84E--02 O.OOEOO 6.45E-04 3.92E-04 l.72EOO Ba-140 2.03E02 2.55E-Ol l.33E01 O.OOEOO 8.66E--02 l.46E-01 4.18E02 Ba-141 4.71E-01 3.56E-04 l.59E-02 O.OOEOO 3.31E-04 2.02E-04 222E-10 Ba-142 2.13E-Ol 2.19E-04 l.34E-02 O.OOEOO l.85E-04 124E-04 3.00E-19 1.a-140 l.51E-01 7.59E-02 2.0lE-02 O.OOEOO O.OOEOO O.OOEOO 5.57E03 1.a-142 7.71E-03 3.51E-03 8.74E--04 O.OOEOO O.OOEOO O.OOEOO 2.56E01 Ce-141 2.63E-02 l.78E-02 2.02E-03 O.OOEOO 826E-03 o:OOEOO 6.80E01 Ce-143 4.64E-03 3.43EOO 3.79E-04 O.OOEOO l.51E-03 O.OOEOO 128E02 Ce-144 137EOO 5.73E-01 7.36E-02 O.OOEOO 3.40E-01 O.OOEOO 4.64E02 Pr-143 5.54E-01 222E-01 2.75E-02 O.OOEOO 128E-01 O.OOEOO 2.43E03 Pr-144 l.81E-03 7.53E-04 922E-05 O.OOEOO 425E-04 O.OOEOO 2.61E-10 Nd-147 3.79E-Ol 4.38E-01 2.62E-02 O.OOEOO 2.56E-Ol O.OOEOO 2.10E03 Pm-147 4.54EOO 427E-Ol 1.73E-01 O.OOEOO 8.0SE-01 O.OOEOO 5.38E02 Pm-149 9.16E-02 lJOE-02 529E-03 O.OOEOO 2.45E--02 O.OOEOO 2.43E03 Sm-151 4.16EOO 7.17E-Ol l.72E-Ol O.OOEOO 8.02E-01 O.OOEOO 3.17E02 Sm-153 5.16E-02 4.31E-02 3.15E-03 O.OOEOO 1.39E-02 O.OOEOO l.54E03 W-187 2.96E02 2.47E02 8.65E01 O.OOEOO O.OOEOO O.OOEOO 8.10E04 Np-239 2.90E-02 2.85E-03 l.57E-03 O.OOEOO 8.89E-03 O.OOEOO 5.85E02
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 40 of 43 ATTACHMENTC Page 1 of 3 LIQUID RADWASTE SYSTEM FIGURE 1.4-1 BV-1 LIQUID RADWASTE SYSTEM STREAM UNIT 1: COLLECTION, TREATMENT, & DISCHARGE Reactor Coolant Cesium Coolant Recovery Tanks From Unit2 Letdown or Removal Ion 2BRS-IOE21A/21B:
Exchangers 1BR-TK-4A/4B Reactor Plant See Attachment C lBR-1-lA/lB 195,000 gal/tank non-aerated page 2 of3 drains
.---------,i Liquid Waste Ion Waste Drains Tanks Exchanger Pre-filter Liquid Waste BVPS-2 Containment Sump Resin Vessels, RWRO Effluent Radiation Cooling 1LW-TK-2A/2B ILW-FL-7, ILW-I-3-1, Aux Bldg Sumps Monitor Tower
'5,000 gal/tank H~ ILW-I-3-2, lLW-I-3-3, ILW-I-3-4, ILW-FL-8, - Blowdown ILW-RWR0-1, ILW-I-3-5, -i~ RM-lLW-104 7,800 gpm Laboratory Drains 1LW-TK-3A/3B ILW-I-3-6 2,000 gal/tank Reactor Plant Samples' *
- BVPS-1 Cooling Miscellaneous Sources Steam Generator ,, 1, Tower Blowdown Tanks ',__ Blowdown 15,000 gpm 1LW-TK-7A/7B 34,500 gal/tank .--J Steam Gen. Pre-filter 1----+
Backwash Unit 1 Chemical Waste From Unit2 To Unit2 Sump 2LWS-TK21A/21B or GOTO 16,000 gal 2SGC-TK23A/23B:
See Attachment C page 2 of3
- Attachment C page 2 of3 _
Contaminated Laundry and Laundry BVPS 1&2 Drains Eflluent Contaminated Drains Tanks Combined 1--------.i~ Radiation Monitor Shower Drains Cooling Tower 1LW-TK-6A/6B Discharge Point 1200 gal/tank RM-lLW-116 22,800 gpm
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level OfUsc:
1/2 General Skill Reference ODCM: LIQUID EFFLUENTS Revision: Page Number:
17 41 of 43 ATTACHMENT C Page 2 of 3 LIQUID RADWASTE SYSTEM FIGURE 1.4-2 BV-2 LIQUID RADWASTE SYSTEM STREAM UNIT 2: COLLECTION, TREATMENT, & DISCHARGE Reactor Coolant Cesium Removal Ion Exchangers Letdown 2BRS-IOE21A/21B i
To Unit 1 GOTO Attachment C Page 1 of3 i Liquid Waste Ion Steam Generator n Containment Sump Liquid Waste Exchanger Blowdown BVPS-1 Drains Tanks Resin Vessels, Test Tanks Cooling Aux Bldg Sumps f---+ Clean-up Filters f-+
Tower 2LWS-TK21A/21B -+ and Strainers 2SGC-TK23A/23B Blowdown Miscellaneous Sources
-+ 10,000 gal/tank ~ 18,000 gal/tank 15,000 gpm 2SGC-IOE21A/21B Reactor Plant Samples i BVPS-2 Cooling Steam Generator Laboratory Drains - Tower Condensate Condensate Hold Tanks
~
.. Blowdown 7,800 gpm Liquid Waste Demineralizer Rinse Water 2SGC-TK2 l A/2 IB 50,000 gal/tank -
~
Effiuent Radiation Monitor - ..-J From Unit 1 i 2SGC-RQIOO 1LW-'f.K-7A/7B:
See Attachment C - 1 Sump
,r page 1 of3 BVPS1&2 Combined Cooling Tower Discharge Point Cable Vault Sump Unit 2 Catch Basin System 22,800 gpm 2,424 gal
Procedure Number:
Beaver Valley Power Station l/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 42 of4 ATTACHMENT C Page 3 of 3 LIQUID RADW ASTE SYSTEM FIGURE 1.4-3 BV-1 AND 2 LIQUID EFFLUENT RELEASE POINTS ERFBldg HVAC Discharge
&Yard Yard Storm BV-20il Storm Runoff& Water Runoff&
Fire Separator No Fire Protection 22&BV-2 Protection BV-1 Primary & Diesel Secondary Heat Generator &
Exchangers & Chillers Turbine Bldg Drains Main Intake Yard Storm Runoff &
Debris Fire Protection, Removal and BV-1 Pump House Roof Yard Storm Backwash Runoff & Fire BV-2 Service Water Protection Strainer Backwash BV-1 Emergency Cooling Tower BV-2 Oil BV-20il BV-1&2 Overflow, and Water Water Condensate BV-1 Circulating Water System Separator No Separator Slowdown& Drain During outage 23 &BV-2 No21 Uncontaminated Turbine Bldg
- River Water Floor&
Equipment BV-2 Primary & Drains Secondary Heat BV-1 Settling Basin, Exchangers &
BV-1 Clarifier, Chillers BV-20il BV-1 Filters W~r Separator BV-1 Oil Separator, No24&
BV-1 Main Plant Floor Drains BV-2Yard Storm Runoff&
- Fire Protection BV-2 Liquid Radioactive Waste Effluent BV-1 Misc Waste Sumps, BV-2 Control Bldg & Cable Tunnel Transformer Slag Pits,*
BV-1 Chemical BV-1 Liqufd BV-2 Service Water Valve Pits, BV-2.
Waste Sump and Radioactive Condensate Polish &. Clamshell Safeguard Cod Lab Sink Waste Effluent . Valve Pit, Demineralized Water Sump*s Drains
Procedure Number:
Beaver Valley Power Station 1/2-0DC-2.01
Title:
Unit: Level Of Use:
1/2 General Skill Reference Revision: Page Number:
ODCM: LIQUID EFFLUENTS 17 43 of 43 ATTACHMENT D Page 1 of 1 SITE BOUNDARY FOR LIQUID EFFLUENTS FIGURE 5-1 SITE BOUNDARY FOR LIQUID EFFLUENTS
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