ML061240158

Offsite Dose Calculation Manual (Odcm), Rev. 9

ML061240158
Person / Time
Site:	Kewaunee
Issue date:	11/29/2005
From:	Nuclear Management Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML061240130	List: ML061240156 ML061240158
References
06-322, FOIA/PA-2010-0209
Download: ML061240158 (187)

Text

KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL (ODCM)

Revision 9 DATE Reviewed By: 4 / 7, , Date:2 2a ' 6$'

Plant Operations Review Committee Approved By: I 2So1 Date: /z/Z/O ReJ~atory Affairs Manager Approved By: <9 Date: ___/_s Ra ation P tion Manager Approved By:, r Date: /003/05" Chemistry anafer

50.59 APPLICABILITY REVIEW (Is the activity exduded from 50.59 review?)

I. Document/Activity number: ODCM revision 9

_ 2. Brief description of proposed activity (what isbeing changed and why):

Implement changes to incorporate new Effluent Conceitration Values of 10 CFR 20, Appendix B, Table 2, Column 2. Change default setpoints for R-16, R-18, R-l9 & R-20 when CW= 0 based on minimum SW flow of 5,000 gpm rather than 10,000 gpm.

_- 3.

Does the proposed activity involve or change any of the folowing documents or processes? Chedc YES or NO for EACH applicability review item. Explain in comments if necessary. [Ref. NMC 50.59 Resource Manual, Section 41 NOTE: If you are unsure if a document or process may be affected, contact the process owner.

Yes No Document or Applicable Contact/Action 3 3 Process Regulation __ __ __ __ __ __

a 0 0 Technical Specifications or Operating License IOCFR50.92 Contact Licensing.

IdentifaNC liettsring cmet.elw rcs b o b0 o Activity/change

~~~license amendmentpreviously or NRC approvedby SER NRC in 1CR09 Identify hne NRC letter incommentsbelow. Process

__ Contact Licensing for assistance.

c O 0i Activity/change covered by an existing approved I OCFR50.59 review, screening, or evaluation.

IOCFR50 LX Appendi B Identify screening or evaluation in comments below.

Process change.

d E0 0 Quality Assurance Program I OOAPDX IOCFRSO.54(a) Contact QA.

e 0 0 Emergency Plan I OCFR50.54(q) Refer to FP-RP02.

f 0 0 Security Plan IOCFRSO.54(p) R efeFP r to y.-SPE-0ll

_8 0O 0 0 IST Plan TRefer IOCFR50.55a(f) Contact IST process owner.

to NAD-01.24.

h 0 ISI Plan 1

_

IOCFR50.55a(g) Contact ISI process owner.

Refer to NADs 01.03, 01.05, and 05.11.

i 0 0 ECCS Acceptance Criteria IOCFR50.46 Contact Licensing.

USAR or any document incorporated by reference - Pross USAR change pet NEP-05.02.

YCheck YES only if change iseditorial (see IOCFR50.71 Contact USAR process owner for assistance.

Attachment A). Cnc__rcs__oaite Commitment - Commitment changes associated Contact Licensing.

k 0 1E with a response to Generic Letters and Bulletins, or IOCFR50 Co Appendix B NAD-05.25.

if described in the USAR require a pre-screening. _____to________5.

Maintenance activity or new/revised maintenance I O 0 procedure - Check YES only if clearly maintenance IOCFR50.65 Evaluate under Maintenance Rule.

and equipment will be restored to its as-designed Refr to NAD-08.20 and NAD08.21.

condition within 90 days (see Attachment C).

New/revised administrative or managerial directive/procedure (e.g., NAD, GNP, Fleet m0 1 Procedure) or a change to any procedtre or other IOCFR50 Appendix B Process procedure/document revision controlled document (e.g., plant drawing) which is clearly editorial/administrative. See Attachments A and B.

4. Conclusion. Check one of the following:

[ All documents'processes listed above are checkedNO. IOCFR50.59 applies to the proposed activity. A 5059 pre-screening shall be performed.

] One or more ofthe docurnents/processes listed above are checked YES, AND controls all aspects of the proposed activity. IOCFR50.59 does NOT apply. Process the change under the applicable program/processfprocedure.

0 One or more of the docunents/processes listed above are checked YES, however, some portion of The proposed activity is not controlled by any of the above processes. IOCFR50.59 applies to that portion. A 50.59 pre-screening shll be performed.

5. Comments:

Implementing the new EC values from 10 CFR 20, Appendix B, Table 2, Column 2 has been previously approved by the NFC as part license TS Amendment No. 186. A 50.59 screening for ODCd rev 9 applies to default setpoint changes with CW - 0

6. Print name followed by signature. Attach completed form to docment/activity/change package.

Prepared by:

(print/sign)

Darryl Holschbach / I Date: 11/10/2005 Reviewed by: o- r jC C' , Date: f /1S/oS (print/sign)

Formn GNP-04.04.O1-1 Rev. F Date: NOV 8 2005 Page 15 of 16 INFORMATION USE

50.59 PRE-SCREENING (is a 50.59 screening required?)

I. Document/Activity number: ODCM revision 9

2. Brief description ofproposed activity (what is being changed and why):

Implement changes to incorporate new Effluent Concertration Values of 10 CFR 20. Appendix B, Table 2, Column 2. Change default selpoints for R-16, R-18, R-19 & R-20 when CW= 0 based on minimum SW flow of 5,000 gem rather than 10,000 gpm.

3. Does the proposed activity involve or change any of the following documents or processes? Explain in Comments if necessary.

Check YES or NO for EACH pre-screening item. [Ref. NMC 50.59 Resource Manual, Section 5.1]

NOTE: If you are unsure if a document or process may be affected, contact the process owner.

NOTE: An asterisk (0) indicates that the document is incorporated by reference in the USAR or is implicitly considered part of the USAR.

NOTE: Check NO if activity/change is considered editorial, administrative, or maintenance as defined in Attachments A, B, and C. Explain in Comments if necessary.

Yes 3 No 3 Document/Process Directive/

Procedure a O 0 Updated Safety Analysis Report (USAR) NEP-05.02 b z* Technical Specifications Bases or Technical Requirements Manual (TRM) NAD-05.14, 0 ~-JNAD-03.25 c O 0

• Commitments made inresponse to NRC Generic Letters and Bulletins, and those described in the USAR NAD-05.25 d O 0 E Environmental Qualification (EQ) Plan NAD-01.08 e E

• Regulatory Guide 1.97 (RG 1.97) Accident Monitoring Instrumentation Plan NAD-05.22 f 0 0

• Fire Plan NAD-01.02 8 E 0

• Appendix R Design Description NAD-01.02 h Q 0

• Fire Protection Program Analysis (FWPA) NAD-01.02 T 0 E

• Offsite Dose Calculation Manual (ODCM) NAD-05.13 _

J El: 0

• Radiological Environmental Monitoring Manual (REMM) NAD-05.13 k E 0

• Station Blackout Design Description El 0

• Control Room Habitability Study m El 0 Plant Drawing Changes/Discrepancies NAD-05.01 n l Calculations/Evaluations/Analyscs/Computer Software -Check YES only if 1) It affects a method of evaluation Various

_ described in the USAR, or 2) It independently (i.e., not part of a modification) affects the licensing or design basis.

o E 0 Permanent Plant Physical Changes - All require a screening. NAD-04.03 El 0 Temporary Plant Physical Changes (TCRs) - Check No only if installed for maintenance AND in effect for less than NAD-04.03

_ E 0 _____

QA USAR.Typing Determinations - Check YES only if reduction in classification, or affects design function as described in NAD-01.01 r 0 Setpoint or Acceptance Criteria - Check YES only if change affects plant monitoring, performance, or operation. Various s Plant Procedures/Revisions - Check YES only if the change directly or indirectly involves operating, controlling or NAD-03.01 configuring an SSC differently than described or credited in USAR._

t El 0 Engineering Specifications - Check YES only if a design function or design requirement may be affected. NAD-05.03 u OperationsNight Orders or Operator WorkArounds - Check YES only if SSCs are operatedor configured differently NAD-12.0S E] __ than described in USAR.

v Temporary plant alterations (e.g., jumpers, scaffolding, shielding, barriers) - Check YES only if installed (or in effect) NAD-08.14, El 0 for maintenance for longer than 90 days at power conditions. GMP-127, FPP-08-09 w E Temporary plant alterations - Check YES only if not associated with maintenance.

x E 0 Corrective/Compensatory Actions - Check YES only if degradedhon-conforming plant condition accepted "as-is" or GNP-1 1.08.03

_ ___ compensatory action taken.

4. Conclusion. Check one of the following:

El All of the documerts or processes listed above are checked NO. A 50.59 screening is NOT required. Process change in accordance with the applicable program/process/procedure.

0 One or more of the documents or processes listed above are chedced YES. A 50.59 screening shall be performed.

5. Comments:

Implementipg the new EC values from 10 CFR 20, Appendix B, Table 2, Column 2 has been previously approved by the NRC as part license TS AmendmentNo.

186. A 50.59 screening for ODCM rev 9 applies to default setpoint changes with CW -0

6. Print name followed by sigpature. Fither th preparer or reviewer shall be scweingqgplifis4. Attach completed form to documentlactivity/change package.

Prepared by: Adz / Ake / >nJAC Date: _ _ __ _ __ _

(print/sign) 9& AdD Reviewed by: I' Date: Asf Form GNP-04.04.01-2 Rev. F Date: NOV 8 2005 Page 16 of 16 INFORMATION USE

10CFR60.59 SCREENING Page 1of5 t-o Document/Activity Nlumher: OffsiteDoseCalculationManualRev.9 SCRN# 05-146 PART I: Describe the Proposed Activity and Search the KNPP USAR (Refer to NMC 50.59 Resource Manual Section 5.3.1) 1.1. Describe the proposed activity, and scope ofthe activity covered by this screening. Appropriate descriptive materials may be attached.

Revision 9 of the ODCM changes the default set points for radiation monitors, R-18, R-19, R-20 & R-16 with no CW flow as given in Table 1.1 to more conservative values based on the lowest possible Service Water dilution flow of 5,000 gpm rather than the previous default of 10,000 gpm. Calculation

1. 10690 rev 0 addendum B. [CAP 25888]

_ I.2. Search the Updated Safety Analysis Report (USAR) including those documents incorporated by reference. Describe relevant function(s), performance requirements, and methods of evaluation of the affected SSCs, and where this information is described in the USAR. In general, any USAR information potentially affected by the activity should be identified (consider both support functions and indirect affects). It is acceptable to attach and highlight applicable portions of the USAR.

E_ USAR 11.0 Waste Disposal System and Radiation Protection System 11.1 Waste Disposal System 11.1.1 Design Basis Control of Releases of Radioactivity to the Environment page 11.1-1 Criterion: The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases the design for radioactivity control must be justified:

In all cases the design for radioactivity control must be justified:

(a) on the basis of 10 CFR 20 requirements, for both normal operations and for any transient situation that might reasonably be anticipated to occur, and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence (GDC 70).

Continued on SCRN # 05-146 Attachment I 1.3. Does the activity involve a chanr to the Technical Specifications?

(Changes to the Technical Specifications require a License Amendment request.)

EJ Yes E No Form GNP-04.04.02-1 Rev. D Date: JUN 23 2005 Page 1 of 20 INFORMATION USE

10CFR50.59 SCREENING Page 2 of 5 SCRN# 05-146 PART II: Determine if the Activity Involves a Desi2n Function (Refer to NMC 50.59 Resource Manual Section 5.3.2)

Compare the proposed activity to the relevant portions of the USAR and answer the following questions:

YES NO QUESTION

1. E] 0 Does the proposed activity involve Safety Analysm or an SSC(s) credited in the Safety Analyses?

2. 0 0 Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?

3. o Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc) or accident?

4 Does the proposed activity involve SSCs whose failure could impact SSC(s) credited in the Safety Analyses?

Does the proposed activity involve USAR-described SSCs or procedural controls that perform functions

5. 0 0 that are required by, or otherwise necessary to comply with, regulations, license conditions, orders, or Technical Specifications?

6. 0 0 Does the activity involve a method of evaluation described in the USAR?

7. 0 0 Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)

8. Elo Does the activity exceed or potentially affect a design basis limit for a fission product barrier (DBLFPB)? If this question is answered YES, this activity requires a I CFR50.59 Evaluation.

If the answer to all of these questions is NO, answer PART III as Not Applicable, and proceed to PART IV. A 10CFR50.59 evaluation is not required.

If any of the above questions are checked YES, identify the specific design function, method of evaluation, or DBLFPB involved:

The radiation monitors, R-16, R-18, R-19 & R 20 are used to control radioactive effluents as required by 10 CFR 20.1302,40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50 and described in Technical Specification 6.16.b.1.

Form GNP-04.04.02-1 Rev. D Date: JUN 23 2005 Page 1 of 20 INFORMATION USE

10CFR60.59 SCREENING Page 3 of 5 SCRN# 05-146 PART III: Determine Whether the Activity Involves Adverse Effects (Refer to NMC 50.59 Resource Manual Section 5.3.3)

If all the questions in Part II were answered NO, then Part III is:

E Not Applicable Answerthe following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a I OCFR50.59 Evaluation is required, except where noted in Question 111.3.

111.1. Changes to the Facility or Procedures YES NO QUESTION

a. El 0 Does the activity adversely affect the design function(s) identified in Part II?

b. El 0 Does the activity introduce an accident of a different type than previously described in the USAR? (see RM Section 6.2.5)

c. 05 0 Does the activity introduce new type of malfunction directly or indirectly affecting an SSC having a design function identified in Part II? (See definition in GNP-04.04.02, Section 3.0)

d. 5 0 Does the activity adversely affect the method of performing or controlling the design function(s) identified in Part 11?

If any answer is YES, a IOCFR50.59 Evaluation is required. For each answer given. describe the basis for the conclusion (attach additional discussion, as necessary):

See SCRN # 05-146 Attachment 2 111.2. Changes to a Method of Evaluation If the activity does not involve a method of evaluation, these questions are:

0 Not Applicable YES NO QUESTION

[a a Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the USAR?

5 a Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the USAR?

If eitheransweris YES, a 10CFRSO.59 Evaluation is required. For each answer aiven, describe the basis forthe conclusion (attach additional discussion, as necessary):

Form GNP-04.04.02-1 Rev. D Date: JUN 23 2005 Page 1 of 20 INFORMATION USE

10CFR50.59 SCREENING Page 4 of 5 SCRN# 05-146 111.3. Tests or Experiments If the activity is not a test or experiment, the questions in 111.3.a and 1II.3.b are:

Z Not Applicable

a. Answer these two questions first:

YES NO QUESTION O El Is the proposed test or experiment bounded by other tests or experiments that are described in the USAR?

C E Are the SSCs affected by the proposed test or experiment isolated from the facility?

If the answerto both questions is NO, continue to 1II.3.b. For each answer given, describe the basis for the conclusion (attach additional discussion, as necessary!:

b. Answer these additional questions only for tests or experiments which do not meet the criteria given above. If the answer to either question in 111.3.a is YES, then these three questions are:

0 Not Applicable YES NO QUESTION _

[ [l Does the activity use or control an SSC in a manner that is outside the reference bounds ofthe design bases as described in the USAR?

0 0 Does the activity use or control an SSC in a manner that is inconsistent with the analyses or descriptions in the USAR?

o E Does the activity place the facility in a condition not previously evaluated or that could affect the capability of an SSC to perform its intended functions?

If any answer in 111.3.b is YES, a IOCFR50.59 Evaluation is required. For each answergLiven, describe the basis for the l conclusion (attach additional discussion, as necessary):

Form GNP-04.04.02-1 Rev. D Date: JUN 23 2005 Page 17 of 20 INFORMATION USE

I 0CFR50.59 SCREENING Page S of 5 SCRN# 05-146 PART IV: Conclusion (Refer to NMC 50.59 Resource Manual Section 5.3.4)

Check all that apply:

I. A 10CFR50.59 Evaluation is a required, OR ED NOT required

2. A change to the USAR and/or any document incorporated by reference is O required (Process change in accordance with applicable plant program/process/procedure.),

OR 0 NOT required Additional comments:

Print name followed by signature. The preparer and reviewer shall be 50.59 screening or evaluation qualified. The completed screening is part of the documentlactivity/change package. Provide a copy of 50.59 screening to the 50.59 Process Owner/Program Coordinator.

Prepared By: PC '-',y,'/ -C _ / r-a Date: t p //ir/'g 5 (print/sign)

Reviewed By: 8i4kc. 6ernsJdofn 1% & ~4 Date: i////S/o5 (print/sign)

r. '*ieJ du: xrj I l /ols ll

• A "ll/5/c0 5-Form GNP-04.04.02-1 Rev. D Date: JUN 23 2005 Page 18 of 20 INFORMATION USE

SCRN # 05-146 Attachment 1 USAR Search Part 1.2 USAR 9.3 Auxiliary Coolant System 9.3.2 System Design and Operation page 9.3-4 9.6 Facility Services 9.6.2 Service Water System Design and Operation page 9.6-3 USAR 11.0 Waste Disposal System and Radiation Protection System 11.1 Waste Disposal System 11.1.1 Design Basis Control of Releases of Radioactivity to the Environment page 11.1-1 Criterion: The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment.

In all cases the design for radioactivity control must be justified:

(a) on the basis of 10 CFR 20 requirements, for both normal operations and for any transient situation that might reasonably be anticipated to occur, and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence (GDC 70).

11.1.2 System Design and Operation pages 11.1-2 & 11.1-4 thru 11.1-5 11.1.4 Minimum Operating Conditions page 11.1-13 All liquid waste releases are continuously monitored for gross activity during discharges to ensure that the activity limits specified in 10 CFR 20 for unrestricted areas are not exceeded. The Off-Site Dose Calculation Manual provides guidance when continuous monitoring is unavailable.

11.2 Radiation Protection 11.2.1 Design Basis Monitoring Radioactive Releases pages 11.1-1 11.2.3 Radiation Monitoring System pages 11.2-7 thru 11.2-10 and 11.2-15 thru 11.2-17 11.2.5 Minimum Operating Conditions page 11.2-27 All liquid waste releases are continuously monitored for gross activity during discharge to ensure that the activity concentration limits are below those specified in 10 CFR 20 for unrestricted areas.

Table 11.2-7 (Radiation Monitor Setpoints "Calculated In Accordance With KNPP Off-site Dose Calculation Manual")

USAR 14.2 Standby Safety Features Analysis 14.2.2 Accidental Release-Recycle of Waste Liquidpages 14.2-5 thru 14.2-7 14.2.4 Steam Generator Tube Rupture pages 14.2-10 thru 14.2-12 TS 6.16.b ODCM Sections 1.2.1 and 1.2.2 ODCM Table 1.1 Calculation # 10690

[CAP 25888]

SCRN # 05-146 Attachment 2 Safety Review Screening PART III Determine Whether the Activity Involves Adverse Effects Answer to questions:

M11.1. Changes to the Facility or Procedures

a. No, the proposed setpoint changes to radiation monitors, R-16, R-18, R-19 & R 20 are not adverse to the level of radioactive effluent control as required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50 and described in Technical Specification 6.16.b.1.

Previous revisions of the ODCM listed calculated default setpoints as given in table 1.1 without CW flow using the default Service Water flow of 10,000 gpm. CAP 25888 identified from operator aid 94-1 and comparison with the normal operating limits of N-SW-02 that SW flow could be as low as 5000 gpm. CAP2588 determined that 5000 gpm rather than 10,000gpm should be used to calculate the default set points for radiation monitors, R-18, R-19, R-20 & R-16 with no CW flow Using the minimum SW flow of 5,000 gpm as input to calculation # 10690, new default setpoints were determined (Calculation # 10690 rev 0 addendum B). The setpoints will become more conservative being reduced by a factor of two (2). Thus they are not adverse to the level of radioactive effluent control.

b. No, the proposed setpoint changes to radiation monitors, R-16, R-18, R-19 & R 20 do not create the possibility of an accident of a different type than any previously evaluated In the USAR.

The radiation monitors are not an accident initiator. They are used for monitoring and controlling effluent releases from the plant and do not affect nuclear safety.

c. No, the proposed setpoint changes to radiation monitors, R-16, R-18, R-19 & R 20 do not Introduce a new type of malfunction directly or indirectly affecting an SSC used to control radioactive effluents.

Changing the radiation monitor setpoints does not change the analyzed design of any plant equipment and cannot create the possibility for a different type of malfunction of equipment than previously evaluated in the USAR.

d. No, the proposed setpoint changes to radiation monitors, R-16, R-18, R-19 & R 20 do not adversely affect the method of performing or controlling radioactive effluents required by 10 CFR 20.1302,40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50.

The methodology to calculate the setpoints has not changed but rather the dilution water input to the calculation. The proposed changes to the setpoints are more conservative. The procedural controls for radioactive liquid discharges remain intact. The bases for protection of the public from radiation per 10 CFR 20; 10 CFR 50, Appendix I; and 40 CFR 190 are maintained. These default setpoint changes in no way reduce the margin of safety as defined in TS 6.16.b. .

C." toUNITED STATES re iNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

• * ***

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 186 TO FACILITY OPERATING LICENSE NO. DPR-43 DOMINION ENERGY KEWAUNEE. INC.

KEWAUNEE POWER STATION

'e DOCKET NO. 50-305

1.0 INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated February 3,2005 (Agencywide Document Access and Management System (ADAMS)

Accession Number ML050450372), Nuclear Management Company, LLC (the former licensee),

proposed an amendment to the Technical Specifications (TSs) for the Kewaunee Nuclear Power Plant (Kewaunee). On July 5, 2005, the NRC issued Amendment No. 185 that reflected the transfer of the license to Dominion Energy Kewaunee, Inc., and the change in the facility name to Kewaunee Power Station.

The proposed changes would revise TS 6.16.b.1, "Radioactive Effluent Controls Program," and TS 6.18, "Off-site Dose Calculation Manual," to be consistent with Title 10 of the Code of Federal Regulations (10 CFR) Part 20, "Standards for Protection Against Radiation," and NUREG-1 431, Revision 3, "Standard Technical Specifications [STS] Westinghouse Plants."

Several of the proposed changes are consistent with TS Task Force (TSTF) Change Travelers TSTF-258A, Revision 4, 'Changes to Section 5.0, Administrative Controls," dated July 31, 2003, and TSTF-308. Revision 1, "Determination of Cumulative and Projected Dose Contnbution in RECP [Radioactive Effluents Control Program)," dated June 13, 2000. While TSTF-308 recommended changes for several TS administrative controls, the licensee proposed only those changes referring to the RECP for inclusion into the Kewaunee TSs.

2.0 REGULATORY EVALUATION

The TSTF process is an industry and NRC-controlled process for proposing and incorporating improvements to the STS. Several of the revisions proposed in this amendment are consistent with TSTF-258A, Revision 4, and TSTF-308, Revision 1. However, since Kewaunee has not adopted the STS, the NRC staff has made a plant-specific evaluation of the licensee's application using TSTF-258A and TSTF-308 as guidance.

The applicable regulatory requirements and guidelines are:

1. NUREG-1 431, Revision 3

2. TSTF-258A, Revision 4

3. TSTF-308, Revision 1

- 4. 10 CFR Part 20

5. 40 CFR Part 190, "Environmental Standards for the Uranium Fuel Cycle"

6. 10 CFR 50, Appendix I, "Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion 'As Low As Is Reasonably Achievable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents"

7. Generic Letter (GL) 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications [RETS] in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program," dated January 31, 1989.

3.0 TECHNICAL EVALUATION

The licensee's proposed changes, and NRC staffs evaluation of the changes, are discussed below.

3.1 TS 6.16.b.1 3.1.1 Proposed TS Changes Revise TS 6.16.b.1 to insert parentheses around the "S" in "MEMBERS OF THE PUBLIC" in the second line of the introductory paragraph and delete "OPERATING' before procedures in the fourth line of the introductory paragraph.

3.7.2 NRC Staff Evaluation Inserting parentheses around the "SW in MEMBERS OF THE PUBLIC makes the wording consistent with the Kewaunee TS title for definition TS 1.0.o.1. NUREG-1431, Section 1.1, "Definitions," does not include a definition of "member(s) of the public." Notwithstanding, NUREG-1431 and Kewaunee TS are written such that the defined terms in TS Section 1 appear identically In capitalized type throughout the TS and Bases as a means of reflecting the use of the term as TS Section 1. The NRC staff finds that this change is administrative with no impact on the safety of the unit and is therefore, acceptable.

The NRC staff finds that the deletion of the word 'operating' in the specification introduction is administrative with no impact on the safety of the unit and is consistent with the wording in NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.2 TS6.16.b.1. B 3.2.1 Proposed TS Change Revise TS 6.16.b.1.B to change "10 CFR Part 20, Appendix B, Table II, Column 2" to "ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402."

3.2.2 NRC Staff Evaluation

• The 1994 revision to 10 CFR Part 20, Appendix B, Table 2, Column 2, incorporated a change in the dose base for the liquid effluent concentration release rate limit from 500 millirem (mrem) per year to 50 mrem per year. Therefore, in order to retain the same dose base upon which the TS 6.1 6.b.1 .b is based (500 mrem per year), the new Appendix B, Table 2, Column 2 values are multiplied by ten. This change is intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements and Is consistent with TSTF-258, Revision 4 and NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.3 TS 6.16.b.1.C 3.3.1 Proposed TS Change Revise TS 6.16.b.1.C to change reference from 10 CFR 20.106 to 10 CFR 20.1302.

3.3.2 NRC Staff Evaluation The 1994 revision of 10 CFR Part 20 replaced 10 CFR 20.106 with 10 CFR 20.1302.

Subpart D, Sections 20.1301 and 20.1302 now state the requirements for dose limits for individual members of the public, and compliance with the dose limits, respectively. This change is consistent with NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.4 TS6.16.b.1.E 3.4.1 Proposed TS Change Revise TS 6.1 6.b.1 .E to separate out the determination of cumulative and projected dose as follows: "Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days."

3.4.2 NRC Staff Evaluation This revision clarifies that determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM is required at least once every 31 days. The existing TS could be interpreted as requiring determining projected dose

contribution of the current calendar quarter and current calendar year every 31 days. The NRC staff finds that this change is administrative in nature and is consistent with TSTF-308, Revision 1, and NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.5 TS6.16.b.1.G 3.5.1 Proposed TS Change Revise TS 6.16.b.1.G to delete current specification and insert:

Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be limited to the following:

1. For noble gases: a dose rate <:500 mrermyr to the total body and a dose rate of c3000 mrem/yr to the skin, and

2. For iodine-1 31, iodine-1 33, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: a dose rate of <1500 mrem/yr to any organ.

3.5.2 NRC Staff Evaluation This revision maintains the same dose rate limits as referenced in the current wording for TS 6.1 6.b.1 .G, while also maintaining consistency with the methods of the ODCM for calculating these dose rates. Recognizing that this limit is an as low as reasonably achievable (ALARA) constraint on the release rate for gaseous effluents and not an annual dose limit, reference to the 10 CFR 20, Appendix B, Table 2, Column I values is not necessary. Change

_ , traveler TSTF-258 states that the change was made to eliminate possible confusion or improper implementation of the revised 10 CFR Part 20 requirements.

The NRC staff finds that the change is administrative in nature and is consistent with TSTF-258 and NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.6 TS 6.16.1.b.J 3.6.1 Proposed TS Change Revise TS 6.16.1 .b.J to insert the phrase", beyond the site boundary," immediately following "MEMBER(S) OF THE PUBLIC".

3.6.2 NRC Staff Evaluation This change is made to provide clarification. The NRC staff finds that the change is administrative in nature and is in accordance with TSTF-258, NUREG-1 431, and 40 CFR Part 190. On this basis, the NRC staff finds the proposed change acceptable.

.5-3.7 Surveillance Provisions 3.7.1 Proposed TS Change Insert the following sentence immediately after TS 6.16.1 .b.J: "The provisions of TS 4.0.b and 4,0.c are applicable to the Radioactive Effluent Controls Program surveillance frequency."

3.7.2 NRC Staff Evaluation This change specifies that the provisions of Kewaunee TS 4.0.b and 4.0.c, which contain provisions for maximum allowable surveillance frequency extension and missed surveillances, are applicable to the RECP surveillances. TS 4.0.b permits a 25 percent extension of the interval specified in the frequency (31 days). Allowing a 25 percent extension in the frequency of performing the monthly cumulative dose and projected dose calculation for the current quarter/year will have no effect on the outcome of the calculations.

As applied to the 31 -day frequency, TS 4.0.b and 4.0.c would allow up to 31 days to complete the surveillance (dose calculation) if it is discovered that the surveillance was not performed within 38 days and 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> (the specified interval plus the 25 percent extension). Allowing 31 days to complete the cumulative dose and projected dose calculation for the current quarter/year is acceptable because it will have no effect on the outcome of the calculations and has no Impact on the risk associated with plant operation. In addition, operating experience has shown that the calculated dose is usually well within limits. Thus, it is considered unlikely that a potential greater time interval between dose calculations will result in inadvertent effluent releases exceeding the specified limits. Additionally, TS 6.16.b.1 requires the RECP, which is contained in the ODCM, to include remedial actions to be taken whenever program limits are exceeded.

The NRC staff finds this change is consistent with the guidance contained in GL 89-01 and NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.8 TS 6.18.b.1.B 3.8.1 Proposed TS Change Revise TS 6.18.b.1.B to change the reference from 10 CFR 20.106 to 10 CFR 20.1302.

3.8.2 NRC Staff Evaluation The 1994 revision of 10 CFR Part 20 replaced 10 CFR 20.106 with 10 CFR 20.1302.

Subpart D,Sectons 20.1301 and 20.1302 now state the requirements for dose limits for Individual members of the public, and compliance with the dose limits, respectively. This change isconsistent with NUREG-1431. On this basis, the NRC staff finds the proposed change acceptable.

3.9 Technical Evaluation Summarv The NRC staff has reviewed the licensee's submittal, and based on our review discussed above, the NRC staff finds the proposed changes to be acceptable.

4.0 VERIFICATIONS AND COMMITMENTS The licensee's application dated February 5, 2005, made the following plant-specific regulatory commitment:

The Kewaunee ODCM and affected plant procedures will be updated during implementation of the approved TS amendment.

The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment can be provided by the licensee's administrative processes, including its commitment management program. The NRC staff has agreed that Nuclear Energy Institute 99-04, Revision 0,

'Guidelines for Managing NRC Commitment Changes," provides reasonable guidance for the control of regulatory commitments made to the NRC staff (see Regulatory Issue Summary 2000-17, "Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff," dated September 21, 2000). Should the licensee choose to incorporate a regulatory commitment into the final safety analysis report or other document with established regulatory controls, the associated regulations would define the appropriate change-control and reporting requirements.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (70 FR 15944). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

7.0 CONCLUSION

The NRC staff has concluded, based on the considerations discussed above, that: (1)there is reasonable assurance that the health and safety of the public will not be endangered by

operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: D. Beaulieu Date: Otcer 4, 2005

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KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL (ODCM)

Revision 9 December 2, 2005 Reviewed By: Tom Webb Date: 11/29/2005 Plant Operations Review Committee Approved By: Thomas Breene Date: 11/28/2005 Regulatory Affairs Manager Approved By: Dan Shannon Date: 11/28/2005 Radiation Protection Manager Approved By: Wally Flint Date: 11/23/2005 Chemistry Manager

Abstract This document has been developed in accordance with the Wisconsin Public Service Corporation (WPSC) commitment made by letter dated August 21, 1984 (from D.C. Hintz to S.A. Varga). It provides the current methodologies and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and gaseous and liquid effluent monitoring alarm/trip setpoints for the Kewaunee Power Station. To develop this document, WPSC contracted the J. Stewart Bland Consultants, Inc. of Maryland; however, rigorous review and final acceptance of this document has been provided by WPSC. Implementation of this' document is the responsibility of the current owner/operator of the Kewaunee Power Station (KPS).

REV. 9 12102/2005

KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction ....... 0-1 Definitions ...... I.... 0-2 1.0 Liquid Effluent .......... 1-1 1.1 Radiation Monitoring Instrumentation and Controls .1-1 1.2 Liquid Effluent Monitor Setpoint Determination .................................. 1-1 1.2.1 Liquid Effluent Monitors (Radwaste, Steam Generator Blowdown and Service Water) .1-2 1.2.2 Conservative Default Values .1-3 1.3 Liquid Effluent Concentration Limits - 10CFR20 . .1-4 1.4 Liquid Effluent Dose Calculations - 10CFR50 . .1-5 1.5 Liquid Effluent Dose Projections .. 1-7 1.6 Onsite Disposal of Low-Level Radioactively Contaminated Waste Streams . 1-8 1.7 Heating Boiler Blowdown Operation with Primary-to-Secondary Leak .. 1-9 Figure 1 Liquid Radioactive Effluent Flow Diagram .1-10 Table 1.1 Parameters for Liquid Alarm Setpoint Determinations .1-11 Table 1.2 Site Related Ingestion Dose Commitment Factors .1-12 Table 1.3 Bioaccumulation Factors (BFi) .1-14 2.0 Gaseous Effluents .2-1 2.1 Radiation Monitoring Instrumentation and Controls . .2-1 2.1.1 Waste Gas Holdup System .2-1 2.1.2 Condenser Evacuation System .2-1 2.1.3 Containment Purge .2-1 2.1.4 Auxiliary Building Vent .2-1 2.1.5 Containment Mini-Purge/Vent System .2-2 2.2 Gaseous Effluent Monitor Setpoint Determination . .2-2 2.2.1 Containment and Auxiliary Building Vent Monitor .2-2 2.2.2 Conservative Default Values .2-3 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10CFR20 .. 2-4 2.3.1 Site Boundary Dose Rate - Noble Gas. 2-4 2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates .2-5 i REV. 9 12/02/2005

KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents (con't) 2.4 Gaseous Effluent Dose Calculations - 10CFR50 . ...............................2-6 2.4.1 Unrestricted Area Dose - Noble Gases ............................................... 2-6 2.4.2 Unrestricted Area Dose - Radioiodine and Particulates .......................... 2-7 2.5 Gaseous Effluent Dose Projection ............................................... 2-9 2.6 Environmental Radiation Protection Standards - 40CFR190 ............................ 2-10 2.7 Incineration of Radioactively Contaminated Oil .............................................. 2-10 2.8 Total Dose ............................................... 2-10 Figure 2 Gaseous Radioactive Effluent Flow Diagram ............................... 2-12 Figure 3 Simplified Heating Boiler Fuel Oil Piping System ....................... 2-13 -

Table 2.1 Dose Factors for Noble Gases ................................................ 2-14 Table 2.2 Parameters for Gaseous Alarm Setpoint Determinations ............... 2-15 Table 2.3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations ............................................... 2-16 Table 2.4 Inhalation Pathway Dose Factors - ADULT ................................. 2-17 Table 2.5 Inhalation Pathway Dose Factors - TEEN ..................................... 2-19 Table 2.6 Inhalation Pathway Dose Factors - CHILD ................................... 2-21 Table 2.7 Inhalation Pathway Dose Factors - INFANT ................................ 2-23 Table 2.8 Vegetation Pathway Dose Factors - ADULT ................................ 2-25 Table 2.9 Vegetation Pathway Dose Factors - TEEN ................................... 2-27 Table 2.10 Vegetation Pathway Dose Factors - CHILD ................................. 2-29 Table 2.11 Grass-Cow-Milk Pathway Dose Factors - ADULT ....................... 2-31 Table 2.12 Grass-Cow-Milk Pathway Dose Factors - TEEN .......................... 2-33 Table 2.13 Grass-Cow-Milk Pathway Dose Factors - CHILD ........................ 2-35 Table 2.14 Grass-Cow-Milk Pathway Dose Factors - INFANT ..................... 2-37 Table 2.15 Ground Plane Pathway Dose Factors ............................................ 2-39 3/4 Radiological Effluent Specifications and Surveillance Requirements ..................... 3-1 3/4.0 Applicability and Surveillance Requirements .................................................. 3-1 3/4.1 Radioactive Liquid Effluent Monitoring Instrumentation .................................. 3-2 3/4.2 Radioactive Gaseous Effluent Monitoring Instrumentation ................................ 3-3 ii REV. 9 12/02/2005

KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents (con't) 3/4.3 Liquid Effluents .............................. 3-4 Concentration. 3-4 Dose .3-5 Liquid Radwaste Treatment System .3-6 3/4.4 Gaseous Effluents .3-7 Dose Rate .3-7 Dose-Noble Gases ................................................. 3-9 Dose-Iodine-131, Iodine-133, Tritium and Radionuclides in Particulate Form 3-10 Gaseous Radwaste Treatment System .3-12 3/4.5 Total Dose .3-14 3/4.6 Reporting Requirements .. 3-16 Table 3.1 Radioactive Liquid Effluent Monitoring Instrumentation. 3-17 Table 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation . 3-18 Table 4.0 Frequency Notation .3-20 Table 4.1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements .3-21 Table 4.2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements .3-22 Table 4.3 Radioactive Liquid Waste Sampling and Analysis Program . 3-23 Table 4.4 Radioactive Gaseous Waste Sampling and Analysis Program . 3-25 Appendices Appendix A Technical Basis for Effective Dose Factors - Liquid Radioactive Effluents ..... A-1 Table A-1 Adult Dose Contributions Fish and Drinking Water Pathways .... A-5 Table A-2 Adult Liver and Total Body Dose Assessment .................. ............... A-6 Appendix B Technical Basis for Effective Dose Factors - Gaseous Radioactive Effluents .. B-1 Table B-1 Effective Dose Factors - Noble Gases ........................................... B-S Appendix C Evaluation of Conservative, Default Effective EC Value for Liquid Effluents . C-I Table C-I Calculation of Effective EC (EC,) ................................................ C4 Appendix D Site Maps ........................................................ D-1 Figure D-1 Gaseous and Liquid Effluent Release Points ................................ D-3 Appendix E Onsite Disposal of Low-Level Radioactively Contaminated Waste Streams .... E-1 iii REV. 9 12/02/2005

KEWAUNEE POWER STATION OFFSITE DOSE CALCULATION MANUAL Introduction The Kewaunee Offsite Dose Calculation Manual (ODCM) describes the methodology and parameters used in:

1) The calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints; and

2) The calculation of radioactive liquid and gaseous concentrations, dose rates and cumulative quarterly and yearly doses.

The methodology stated in this manual is acceptable for use in demonstrating compliance with 10CFR20. 1302, 10CFR50, Appendix 1, and 40CFR190. I More conservative calculational methods and/or conditions (e.g., location and/or exposure pathways) expected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations.

The ODCM will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations. Changes will be made to the ODCM calculational methodologies and parameters as is deemed necessary to assure reasonable conservatism in keeping with the principles of 10CFR50.36a and Appendix I for demonstrating radioactive effluents are ALARA.

Rev. 9 0-1 12/02/2005

Definitions

1. ACTION ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

2. GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting off-gases from the primary coolant system and providing for delay or holdup for the purpose of reducing the total radioactivity released to the environment.

3. INSTRUMENTATION SURVEILLANCE

a. CHANNEL CHECK

b. CHANNEL FUNCTIONAL TEST

c. CHANNEL CALIBRATION

d. SOURCE CHECK As defined in the Technical Specifications.

4. MEMBER(S) OF THE PUBLIC MEMEBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

5. OPERABLE-OPERABILITY As defined in the Technical Specifications.

6. PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other OPERATING condition, in such a manner that replacement air or gas is required to purify the confinement.

Rev. 9 0-2 12/02/2005

7. RADIOLOGICAL ENVIRONMENTAL MONITORING MANUAL (REMM)

The REMM shall contain the current methodology and parameters used in the conduct of the radiological environmental monitoring program.

8. SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

9. UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

10. VENTILATION EXHAUST TREATMENT SYSTEM A VENTIATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature atmospheric cleanup systems (i.e., Auxiliary Building special ventilation, Shield Building ventilation, spent fuel pool ventilation) are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

Rev. 9 0-3 12/02/2005

1.0 Liquid Effluents 1.1 Radiation Monitoring Instrumentation and Controls The liquid effluent monitoring instrumentation and controls installed at Kewaunee for controlling and monitoring normal radioactive material releases in accordance with 10 CFR 50, Appendix A, Criteria 60 and 64, are summarized as follows:

1) Alarm (and Automatic Termination) - R-18 provides this function on the liquid radwaste effluent line, R-19 on the Steam Generator blowdown.

2) Alarm (only) - R-20 and R-16 provide alarm functions for the Service Water discharges.

3) Composite Samples - Samples are collected weekly from the steam generator blowdown and analyzed by gamma spectroscopy. Samples are collected weekly from the Turbine Building Sump and analyzed by gamma spectroscopy. The weekly samples are composited for monthly tritium and gross alpha analyses and for quarterly Sr-89, Sr-90 , and Fe-55 analyses. During periods of identified primary-to-secondary leakage (with the secondary activity > 1.OE-05 pCi/ml), grab samples from the Turbine Building sump are collected daily and analyzed by gamma spectroscopy.

These samples are composited for monthly tritium and gross alpha analyses and for quarterly Sr-89, Sr-90, and Fe-55 analyses.

4) Liquid Tank Controls - All radioactive liquid tanks are located inside the Auxiliary Building and contain the suitable confinement systems and drains to prevent direct, unmonitored release to the environment. A liquid radioactive waste flow diagram with the applicable, associated radiation monitoring instrumentation and controls is presented as Figure 1.

1.2 Liquid Effluent Monitor Setpoint Determination Per the requirements of Technical Specification 6.16.b.1.B and ODCM Specification 3.1, alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of ODCM Specification 3.3.1 are met (i.e., the concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides and 2.OE-04 jiCi/ml for dissolved or entrained noble gases). The following equation' must be satisfied to meet the liquid effluent restrictions:

c< 10XC(F+f) (1.1)

'Adapted from NUREG-0133 to include the application of 10 times the Effluent Concentration (EC) of 10 CFR 20, Appendix B,Table 2, Column 2.

1-1 REV. 9 12/02/2005

where:

lOxC = ten times the effluent concentration limit of 10 CFR 20, Appendix B, Table 2, Column 2, in gCi/ml. For dissolved and entrained noble gases equals 2xI0 4 jxCi/mI.

c = the setpoint, in gCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely proportional to the volumetric flow of the effluent line and proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value which, if exceeded, would result in concentrations exceeding the limits of ODCM Specification 3.3.1.

f = the flow rate at the radiation monitor location in volume per unit time, but in the same units as F, below.

F = the dilution water flow rate as measured prior to the release point, in volume per unit time.

[Note that if no dilution is provided, c < C. Also, note that when (F) is large compared to (f), then (F + f) = F.]

1.2.1 Liquid Effluent Monitors (Radwaste. Steam Generator Blowdown and Service Water)

The setpoints for the liquid effluent monitors at the Kewaunee Power Station are determined by the following equations:

SP< CWx+(gxSEN;) bkg (1.2)

-' xRR M0xEC, where:

SP = alarm setpoint corresponding to the maximum allowable release rate (cpm)

Ci = the concentration of radionuclide "i" in the liquid effluent (gCi), to include gamma emitters only 10 x ECi = ten times the EC value corresponding to radionuclide "i" from 10 CFR 20, Appendix B, Table 2, Column 2 (gCi/ml) 1-2 REV. 9 12/02/2005

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SENi = the sensitivity value to which the monitor is calibrated for radionuclide "i" (cpm per pCi/ml). The default calibration value from Table 1.1 may be used for gamma emitting radionuclides in lieu of nuclide specific values.

CW = the circulating water flow rate (dilution water flow) at the time of release (gal/min)

RR = the liquid effluent release rate (gal/min) bkg = the background of the monitor (cpm)

The radioactivity monitor setpoint equation (1.2) remains valid during outages when the circulating water dilution is at its lowest. Reduction of the waste stream flow (RR) may be necessary during these periods to meet the discharge criteria. At its lowest value, CW will equal RR and equation (1.2) reverts to the following equation:

SP C + bkg (1.3)

Ci X (l0xEC) 1.2.2 Conservative Default Values Non-gamma emitting radionuclides (H-3, Fe-55, Sr-89/90) are not detected by the effluent monitor and, therefore, are not directly included in the above setpoint equation. These non-gamma radionuclides can, however, contribute a sizable fraction of the total EC limit (refer to Appendix C). The method specified below for establishing default setpoints provides conservatism to account for these non-gamma emitters and ensures that the setpoint meets the requirements of ODCM Specification 3.1 including all radionuclides. Refer to Appendix C for further discussion.

Conservative alarm setpoints have been determined through the use of generic, default parameters. Table 1.1 summarizes all current default values in use for Kewaunee. They are based upon the following:

a) substitution of the default effective EC (EC,) value of 1.OE-06 l pCi/ml (refer to Appendix C for justification),

where, EC = Ec (1.4)

X(ECi) 1-3 REV. 9 12/02/2005

b) substitution of the lowest operational circulating water flow, in gal/min; and, c) substitution of the highest effluent release rate, in gal/min, d) substitution of the default monitor sensitivity.

The default setpoint equation is provided below:

SP < ECex1OxSENxCW + bkg (1.5) 13 Liquid Effluent Concentration Limits -10 CFR 20 ODCM Specification 3.3.1 limits the concentration of radioactive material in liquid effluents (after dilution in the Circulating Water System) to less than ten times the concentrations as specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than noble gases. Noble gases are limited to a diluted concentration of 2E-04 gCi/ml. Release rates are controlled and radiation monitor alarm setpoints are established to ensure that these concentration limits are not exceeded. In the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance with the concentration limits of ODCM Specification 3.3.1 may be performed using the following equation:

where:

[(Ci + (1OxECi))x(RR + CW)]S1 (1.6)

Ci = concentration of radionuclide "i" in the undiluted liquid effluent (IiCi/ml) 10 x ECi = ten times the EC value corresponding to radionuclide "i" from 10 CFR 20, Appendix B, Table 2, Column 2 (gCi/m)l

= 2E-04 jiCi/ml for dissolved or entrained noble gases RR = the liquid effluent release rate (gal/min)

CW = the circulating water flow rate (dilution water flow) at the time of the release (gal/min) 1-4 REV. 9 12/02/2005

IA Liquid Effluent Dose Calculation - 10 CFR 50 ODCM Specification 3.3.2 limits the dose or dose commitment to members of the public from radioactive materials in liquid effluents from the Kewaunee Power Station to:

• during any calendar quarter;

< 1.5 mrem to total body

< 5.0 mrem to any organ

• during any calendar year,

< 3.0 mrem to total body

• 10.0 mrem to any organ.

Per Surveillance Requirement 4.3.2, the following calculational methods may be used for determining the dose or dose commitment due to the liquid radioactive effluents from Kewaunee.

Do= l. 6 7 E 2 xVOLx (Cjx A-.) (1.7) cw where:

Do = dose or dose commitment to organ "o", including total body (mrem)

Ajo = site-related ingestion dose commitment factor to the total body or any organ "o" for radionuclide "i" (mrem/hr per gCi/ml) (Table 1.2)

C3 = average concentration of radionuclide "i", in undiluted liquid effluent representative of the volume VOL (pCi/ml)

VOL = volume of liquid effluent released (gal)

CW = average circulating water discharge rate during release period (gal/min) 1.67E-02 = conversion factor (hr/min) 1-5 REV. 9 12/02/2005

The site-related ingestion dose/dose commitment factors (Ai.) are presented in Table 1.2 and have been derived in accordance with guidance of NUREG-0133 by the equation:

Aio = 1.14E + 05[(Uw foDw)+ (UFXBFi)]DFi (1.8) where:

AO = composite dose parameter for the total body or critical organ "o" of an adult for radionuclide "i", for the fish ingestion and water consumption pathways (mrem/hr per pCi/ml) 1.14E+05 = conversion factor (pCi/gCi x ml/kg + hr/yr)

UW = adult water consumption (730 kg/yr)

DW = dilution factor from the near field area within 1 mile of the release point to the nearest potable water intake for the adult water consumption (842, unitless)

UF = adult fish consumption (21 kg/yr)

BF, = bioaccumulation factor for radionuclide "i" in fish from Table 1.3 (pCi/kg per pCi/i)

DF1 = dose conversion factor for nuclide "i" for adults in pre-selected organ "o", from Table E-1 1 of Regulatory Guide 1.109, 1977 and NUREG 0172, 1977 (mrem/pCi)

The radionuclides included in the periodic dose assessment per the requirements of ODCM Specification 3.3.2 and Surveillance Requirement 4.3.2 are those as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per Surveillance Requirement 4.3.1.1, Table 4.3.

Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Table 4.3.

In lieu of the individual radionuclide dose assessment as presented above, the following simplified dose calculational equation may be used for demonstrating compliance with the dose limits of ODCM Specification 3.3.2. (Refer to Appendix A for the derivation and justification for this simplified method.)

2 Adapted from the Kewaunee Final Environmental Statement,Section V.

1-6 REV. 9 12/02/2005

Total Body DIb = 9.67E+03x VOL XC (1.9) ew Maximum Organ

= 1.18E+04xVOL Xc (1.10)

CW where:

Ci = average concentration of radionuclide "i", in undiluted liquid I effluent representative of the volume VOL (gCi/ml)

VOL = volume of liquid effluent released (gal)

CW = average circulating water discharge rate during release period (gal/min)

Dti = conservatively evaluated total body dose (mrem)

Dmax = conservatively evaluated maximum organ dose (mrem) 9.67E+03 = product of the hour-to-minute conversion factor (hr/min) and the I conservative total body dose conversion factor (Cs-134, total body

-- 5.79E+05 mrem/hr per gCi/ml) 1.1 8E+04 = product of the hour-to-minute conversion factor (hr/min) and the I conservative maximum organ dose conversion factor (Cs-134, liver

-- 7.09E+05 mrem/hr per ACi/ml) 1.5 Liquid Effluent Dose Projections ODCM Specification 3.3.3 requires that the liquid radioactive waste processing I system be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed:

0.18 mrem to the total body, or

• 0.62 mrem to any organ.

The applicable liquid waste streams and processing systems are as delineated in Figure 1.

1-7 REV. 9 12/02/2005

Dose projections are made at least once per 31 days by the following equations:

Dbp= Dtb(91 + d) (1.11)

Dfnxp = D(91 +d) (1.12) where:

Dtbp = the total body dose projection for current calendar quarter (mrem)

Dtb = the total body dose to date for current calendar quarter as determined by equation (1.7) or (1.9) (mrem)

Dffmtp = the maximum organ dose projection for current calendar quarter (mrem)

Dma~x = the maximum organ dose to date for current calendar quarter as determined by equation (1.7) or (1.10) (mrem) d = the number of days to date for current calendar quarter 91 = the number of days in a calendar quarter 1.6 Onsite DisDosal of Low-Level Radioactivelv Contaminated Waste Streams During the normal operation of Kewaunee, the potential exists for in-plant process streams, which are not normally radioactive to become contaminated with very low levels of radioactive materials. These waste streams are normally separated from the radioactive streams. However, due mainly to infrequent, minor system leaks, and anticipated operational occurrences, the potential exists for these systems to become slightly contaminated. At Kewaunee, the secondary system demineralizer resins, the service water pretreatment system sludges, the make-up water system resins, and the sewage treatment plant sludges are waste streams that have the potential to become contaminated at very low levels. During the yearly testing of a batch of pre-treatment sludge, it was found that approximately 15,000 cubic feet of sludge had been contaminated with Cs-137 and Co-60.

The potential radiation doses to members of the public from these onsite disposal methods are well below 1 mrem per year. This dose is in keeping with the guidelines of the National Council on Radiation Protection (NCRP) in their Report No. 91, in which the NCRP established a "negligible individual risk level" at a dose rate of 1 mrem per year.

1-8 REV. 9 12/02/2005

It is for these type wastes that the NRC acknowledged in Information Notice No. 83-05 and 88-22 that the levels of radioactive material are so low that control and disposal as a radwaste are not warranted. The potential risks to man are negligible and the disposal costs as a radwaste are unwarranted and costly.

This waste material will be monitored and evaluated prior to disposal to ensure its radioactive material content is negligible. It shall then be disposed of in a normal conventional manner with records being maintained of all materials disposed of using these methods.

Approvals for specific alternate disposal methods are listed in Appendix E. Currently, only service water pretreatment (SWPT) facility lagoon sludge and sewage treatment plant sludge have been approved for disposal by land spreading.

1.7 Heating Boiler Blowdown Ooeration with Primary-to-Secondary Leak During operation with a primary-to-secondary leak, the potential exists for non-radioactive systems to become contaminated. One such system is the heating system. Activity is transferred from the reactor coolant system into the secondary main steam system through the leak and then into the heating system. Heating boiler operation following operation with a primary-to-secondary leak will result in the heating boiler becoming contaminated.

When the heating boiler is operated, it must be periodically blown down to remove impurities, which collect in the system. This blowdown is normally directed to the steam generator blowdown tank but can be diverted to the circulating water discharge. Either way, the blowdown becomes a release path for radioactivity to the environment. The heating boiler blowdown is sampled, using current plant procedures, whenever the primary-to-secondary leakage exceeds 10 gallons per day and the gross gamma activity or tritium activity exceeds 1.OE-05 ZCi/ml. The results of these samples allow for the activity being released to the environment to be quantified. This is similar to the method used for the turbine building sump release path. The radioactive effluent limits of 10 CFR Part 20,40 CFR 190, and Technical Specifications can therefore be maintained.

1-9 REV. 9 12/02/2005

Figure 1 Luid Radioactive Efuent Flow Diagram Ilm (1w wu T I _ . . _ . - . _ .

t._

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ba, L4.....q........................ rnae amuwaeaQ Wlc4c tM F C CL)cMFIGIYLD 1-10 REV. 9 12/02/2005 I lI I I I I I I I I I I

Table 1.1 Parameters for Liquid Alarm Setpoint Determinations Parameter Actual Value Default Value* Units Comments EC, calculated 1.OE06** ACi/nm Calculate for each batch to be released emesrdN/A gindTaken from gamma spectral analysis of measured _liquid effluent a nTaken from 10 CFR 20, Appendix B, lECas determinedN/Ai Table 2, Col. 2 Sensitivity (SEN)

R-18 as determined 1.OE+08 Radwaste effluent R-19 as determined 1.OE+08 cpm per Steam Generator blowdown R-20 as determined 1.OE+08 aCi/ml Service Water - component cooling R-16 as determined 9.8E+07 Service Water - Containment fan cooling CW as determined 2.58E+05 gpm Circulating Water System default -

winter, single CW pump Release Rate (RR)

R-18 as determined 8.0E+01 Determined prior to release; release rate can be adjusted for Technical

. gpm Specification compliance lR-19 as determined 2.01+02 team Generator A and B combined R-20 as determined 5.OE+03 Service Water - component cooling R-16 as determined 1.5E1+03 Service Water - Containment fan cooling Background (bkg)

R-18 as determined 2.0E+03 cpm Nominal values only; actual values may R-19 as determined 8.OE+01 be used in lieu of these reference values R-20 as determined 6.0E1+01 R-16 as determined 8.OE+01 Setpoint* (SP) Default alarm setpoints; more R-18 calculated 3.22E+06 + bkg cpm conservative values may be used as deem R-19 calculated 1.29E+06 + bkg appropriate and desirable for assuring R-20 calculated 5.16E+04 + bkg regulatory compliance and for R-16 calculated 1.69E+05 + bkg maintaining releases ALARA.

etpoint* (SP) with no Circulating Water System flow, CW=0 R-18 calculated 6.25E+04+ bkg For outages with no Circulating Water R-19 calculated 2.50E+04 + bkg System flow (CW=0) and a dilution flow R-20 calculated 1,OOE+03 + bkg cpm as provided by the Service Water system R-16 calculated 3.26E+03 + bkg of 5,000 gpm total.***

Refer to Calculation # C10690 for the default setpoint calculation.

• $ Refer to Appendix C for derivation

• SW flow is based on N-SW-02 Operating Parameters and Service Water Pump Flow Curves.

1-11 REV. 9 12/02/2005

Table 1.2 Site Related Ingestion Dose Commitment Factors (mrem/hr per pCI/mi)

Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI H-3. - 3.30E-1 3.30E-1 3.30E-1 3.30E-1 3.30E-1 3.30E-1 C-14 3.13E+4 6.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 Na-24 4.09E+2 4.09E+2 4.09E+2 4.09E+2 4.09E+2 4.09E+2 4.09E+2 P-32 1.39E+6 8.62E+4 5.36E+4 - - - 1.56E+5 Cr-51 - - 1.28E+0 7.63E-1 2.81 E-1 1.69E+O 3.21 E+2 Mn-54 - 4.38E+3 8.36E+2 - 1.30E+3 - 1.34E+4 Mn-56 - 1.1OE+2 1.96E+1 - 1.40E+2 - 3.52E+3 Fe-55 6.61 E+2 4.57E+2 1.06E+2 - - 2.55E+2 2.62E+2 Fe-59 1.04E+3 2.45E+3 9.40E+2 - - 6.85E+2 8.17E+3 Co-57 - 2.11 E+1 3.51 E+1 - - - 5.36E+2 Co-58 - 8.99E+1 2.02E+2 - - - 1.82E+3 Co-60 - 2.58E+2 5.70E+2 - - - 4.85E+3 Ni-63 3.13E+4 2.17E+3 1.05E+3 - - - 4.52E+2 Ni-65 1.27E+2 1.65E+1 7.52E+O - - 4.18E+2 Cu-64 - 1.01 E+1 4.72E+0 - 2.53E+1 - 8.57E+2 Zn-65 2.32E+4 7.38E+4 3.33E+4 - 4.93E+4 - 4.65E+4 Zn-69 4.93E+1 9.43E+1 6.56E+0 - 6.13E+1 - 1.42E+1 Br-82 - - 2.27E+3 - - - 2.61 E+3 Br-83 - - 4.05E+1 - - - 5.83E+1 Br-84 - - 5.24E+1 - - - 4.12E-4 Br-85 - - 2.15E+O - -

Rb-86 - 1.01 E+5 4.71 E+4 - - - 1.99E+4 Rb-88 - 2.90E+2 1.54E+2 - - - 4.OOE-9 Rb-89 - 1.92E+2 1.35E+2 - - - -

Sr-89 2.24E+4 - 6.44E+2 - - - 3.60E+3 Sr-90 5.52E+5 - 1.35E+5 - - - 1.59E+4 Sr-91 4.13E+2 - 1.67E+1 - - - 1.97E+3 Sr-92 1.57E+2 - 6.77E+0 - - - 3.10E+3 Y-90 5.85E-1 - 1.57E-2 - - - 6.21 E+3 Y-91m 5.53E-3 - 2.14E-4 - - - 1.62E-2 Y-91 8.58E+0 - 2.29E-1 - - - 4.72E+3 Y-92 5.14E-2 - 1.50E-3 - - - 9.OOE+2 Y-93 1.63E-1 - 4.50E-3 - - - 5.17E+3 Zr-95 2.70E-1 8.67E-2 5.87E-2 - 1.36E-1 - 2.75E+2 Zr-97 1.49E-2 3.01 E-3 1.38E-3 - 4.55E-3 - 9.34E+2 Nb-95 4.47E+2 2.49E+2 1.34E+2 - 2.46E+2 - 1.51 E+6 Nb-97 3.75E+0 9.48E-1 3.46E-1 - 1.11 E+O 3.50E+3 Mo-99 - 1.07E+2 2.04E+1 - 2.43E+2 - 2.49E+2 Tc-99m 9.11 E-3 2.58E-2 3.28E-1 - 3.91 E-1 1.26E-2 1.52E+1 Tc-101 9.37E-3 1.35E-2 1.32E-1 - 2.43E-1 6.90E-3 Ru-103 4.61E+O - 1.99E+0 - 1.76E+1 - 5.39E+2 Ru-105 3.84E-1 - 1.52E-1 - 4.96E+0 - 2.35E+2 Ru-106 6.86E+1 - 8.68E+0 - 1.32E+2 - 4.44E+3 Rh-103Mr 1-12 REV. 9 12/02/2005

Table 1.2 Site Related Ingestion Dose Commitment Factors (mrem/hr per ;LCIml)

Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI Rh-106 . - -

Ag-i1Om 1.04E+0 9.62E-1 5.71 E-1 - 1.89E+0 - 3.92E+2 Sb-124 9.48E+O 1.79E-1 3.76E+0 2.30E-2 - 7.38E+O 2.69E+2 Sb-125 6.06E+0 6.77E-2 1.44E+0 6.16E-3 - 4.67E+0 6.67E+1 Te-1 25m 2.57E+3 9.31 E+2 3.44E+2 7.73E+2 1.04E+4 - 1.03E+4 Te-1 27m 6.49E+3 2.32E+3 7.91 E+2 1.66E+3 2.64E+4 - 2.1 8E+4 Te-127 1.05E+2 3.79E+1 2.28E+1 7.81E+1 4.29E+2 - 8.32E+3 Te-129m 1.10E+4 4.11E+3 1.74E+3 3.79E+3 4.60E+4 5.55E+4 Te-129 3.01 E+1 1.13E+1 7.33E+0 2.31 E+1 1.27E+2 - 2.27E+1 Te-131m 1.66E+3 8.11E+2 6.76E+2 1.28E+3 8.22E+3 - 8.05E+4 Te-131 1.89E+1 7.89E+O 5.96E+0 1.55E+1 8.27E+1 - 2.67E+O Te-1 32 2.42E+3 1.56E+3 1.47E+3 1.73E+3 1.50E+4 - 7.39E+4 1-130 2.79E+1 8.23E+1 3.25E+1 6.97E+3 1.28E+2 - 7.08E+1 1-131 1.54E+2 2.20E+2 1.26E+2 7.20E+4 3.76E+2 - 5.79E+1 1-132 7.49E+0 2.0OE+1 7.01E+0 7.01E+2 3.19E+1 - 3.76E+O 1-133 5.24E+1 9.11E+1 2.78E+1 1.34E+4 1.59E+2 - 8.19E+1 1-134 3.91E+0 1.06E+1 3.80E+0 1.84E+2 1.69E+1 - 9.26E-3 1-135 1.63E+1 4.28E+1 1.58E+1 2.82E+3 6.86E+1 - 4.83E+1 Cs-1 34 2.98E+5 7.09E+5 5.79E+5 - 2.29E+5 7.61 E+4 1.24E+4 Cs-136 3.12E+4 1.23E+5 8.86E+4 - 6.85E+4 9.39E+3 1.40E+4 Cs-137 3.82E+5 5.22E+5 3.42E+5 - 1.77E+5 5.89E+4 1.01E+4 Cs-1 38 2.64E+2 5.22E+2 2.59E+2 - 3.84E+2 3.79E+1 2.23E-3 Ba-1 39 1.02E+0 7.30E-4 3.OOE-2 - 6.83E-4 4.14E-4 1.82E+O Ba-140 2.15E+2 2.69E-1 1.41E+1 - 9.16E-2 1.54E-1 4.42E+2 Ba-141 4.98E-1 3.76E-4 1.68E-2 - 3.50E-4 2.13E-4 -

Ba-142 2.25E-1 2.31E-4 1.42E-2 - 1.95E-4 1.31E-4 -

La-140 1.52E-1 7.67E-2 2.03E-2 - - - 5.63E+3 La-142 7.79E-3 3.54E-3 8.82E-4 - - - 2.59E+1 Ce-141 3.17E-2 2.14E-2 2.43E-3 - 9.95E-3 - 8.19E+1 Ce-143 5.58E-3 4.13E+O 4.57E-4 - 1.82E-3 - 1.54E+2 Ce-144 1.65E+0 6.90E-1 8.87E-2 - 4.1OE-1 - 5.58E+2 Pr-143 5.60E-1 2.25E-1 2.77E-2 - 1.30E-1 - 2.45E+3 Pr-144 1.83E-3 7.61E-4 9.31E-5 - 4.29E-4 - -

Nd-147 3.83E-1 4.42E-1 2.65E-2 - 2.59E-1 - 2.12E+3 W-187 2.96E+2 2.47E+2 8.65E+1 - - - 8.10E+4 Np-239 2.97E-2 2.92E-3 1.61 E-3 - 9.1OE-3 - 5.98E+2 1-13 REV. 9 12/02/2005

Table 1.3 Bioaccumulatlon Factors(BFi)

(pCilkg per pC~liter)*

.Element Freshwater Fish H 9.0E-01 C 4.6E+03 Na 1.0E2+02 P 3.0E+03 Cr 2.OE+02 Mn 4.OE+02 Fe 1.OE+02 Co 5.OE+01 Ni 1.OE+02 Cu 5.0E+01 Zn 2.OE+03 Br 4.2E+02 Rb 2.OE+03 Sr 3.OE+01 Y 2.5E+01 Zr 3.3E+OO Nb 3.OE+04 Mo 1.OE+01 Tc 1.5E+01 Ru 1.OE+01 Rh 1.OE+01 Ag 2.3E+OO Sb 1.OE+OO Te 4.OE+02 I 1.5E+01 Cs 2.OE+03 Ba 4.OE+OO La 2.5E+01 Ce 1.OE+OO Pr 2.5E+01 Nd 2.5E+01 W 1.2E+03 Np 1.OE+01

• Values in this Table are taken from Regulatory Guide 1.109 except for phosphorus which is adapted from NUREG/CR-1336 and silver and antimony which are taken from UCRL 50564, Rev. 1, October 1972.

1-14 REV. 9 12/02/2005

9 Gaseous Effluents 2.1 Radiation Monitorin! Instrumentation and Controls The gaseous effluent monitoring instrumentation and controls at Kewaunee for controlling and monitoring normal radioactive material releases in accordance with 10 CFR 50, Appendix A, Criteria 60 and 64, are summarized as follows:

2.1.1 Waste Gas Holdup System The vent header gases are collected by the waste gas holdup system. Gases may be recycled to provide cover gas for the CVCS hold-up tanks or held in the waste gas tanks for decay prior to release. Waste gas decay tanks are batch released after sampling and analysis. The tanks are discharged via the Auxiliary Building vent. R-13 and/or R-14 provide noble gas monitoring and automatic isolation.

2.1.2 Condenser Evacuation System The air ejector discharge is monitored by R-15. Releases from this system are normally via the Auxiliary Building vent and are monitored by R-13 and/or R-14.

2.1.3 Containment Purge Containment purge and ventilation is via the containment stack for the 36-inch RBV system but via the auxiliary building stack for the 2-inch vent and mini-purge blower system. The stack radiation monitoring system consists of:

• a noble gas activity monitor providing alarm and automatic termination of release (R-12 and R-21),

• an iodine sampler, and

• a particulate sampler.

Effluent flow rates are determined empirically as a function of fan operation (fan curves).

Sampler flow rates are determined by flow rate instrumentation.

2.1.4 Auxiliary Building Vent The Auxiliary Building vent receives discharges from the waste gas holdup system, condenser evacuation system, fuel storage area. ventilation, Auxiliary Building radwaste processing area ventilation, 2-inch containment pressure relief purge/vent system, and Auxiliary Building general area. All effluents pass through the R-13 and/or R-14 channels which contain:

• a noble gas monitor,

• an iodine sampler, and

• a particulate sampler.

2-1 REV. 9 12/02/2005

The noble gas monitor provides auto isolation of any waste gas decay tank release and diverts other releases through the special ventilation system. Effluent flow rates are determined by installed flow measurement equipment or as a function of fan operation (fan curves). Sampler flow rates are determined by flow rate instrumentation.

2.1.5 Containment Mini-PurgeNent System Slight pressure buildup in containment is a recurring event resulting from normal operation of the plant. Prior to exceeding 2 psig in containment, this excess pressure is vented off. Air from containment is routed to the Auxiliary Building ventilation system, via the post-LOCA hydrogen recombiner piping and then out through the Auxiliary Building vent stack. The system is also designed to allow a continuous supply of fresh air to be introduced into containment via a mini-blower to purge gases. An alarm of the Auxiliary Building vent stack monitor (R-13 or R-14) or the containment building airborne radioactivity monitors (R-11, R-12) provides automatic isolation.

A gaseous radioactive waste flow diagram with the applicable, associated radiation monitoring instrumentation and controls is presented as Figure 2.

2.2 Gaseous Effluent Monitor Setpoint Determination _

2.2.1 Containment and Auxiliary Building Vent Monitor Per the requirements of ODCM Specification 3.2, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed corresponding dose rate at the site boundary of 500 mrem/year to the total body or 3000 mrem/year to the skin. Based on a grab sample analysis of the applicable release (i.e., grab sample of the Containment vent or Auxiliary Building vent), the radiation monitoring alarm setpoints may be established by the following calculational method:

FRACtb = [4.72E + 02x X/Qx VFx (CxK.)]+ 500 (2.1) _

FRACskin = [4.72E + 02x X/Q x VFx E (Ci x (L + 1.lMi))]J 3000 (2.2) where:

FRACtb = fraction of the allowable release rate for the total body based on the identified radionuclide concentrations and the release flow rate FRACski. = fraction of the allowable release rate for skin based on the identified radionuclide concentrations and the release flow rate xQ = annual average meteorological dispersion for direct exposure to noble gas at the controlling site boundary location (sec/m 3 , from Table 2.3)

VF = ventilation system flow rate for the applicable release point and monitor (ftW/min, from Table 2.3) 2-2 REV. 9 12/02/2005

Ci = concentration of noble gas radionuclide "i" as determined by radioanalysis of grab sample (gCi/cm3 )

K; = total body dose conversion factor for noble gas radionuclide "i" (mrem/yr per gCi/m3, from Table 2. 1)

= beta skin dose conversion factor for noble gas radionuclide "i" (mrem/yr per pCi/m 3, from Table 2.1)

= gamma air dose conversion factor for noble gas radionuclide "i" (mrad/yr per pCi/M3 , from Table 2.1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad) 4.72E+02 = conversion factor (cm3 Ift3 x min/sec) 500 = total body dose rate limit (mremlyr) 3000 = skin dose rate limit (mrem/yr)

Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoint for the Containment and Auxiliary Building vent monitors at Kewaunee may be calculated:

SP = [(CixSENi) FRAC]+bkg (2.3) where:

SP = alarm setpoint corresponding to the maximum allowable release rate (cpm)

SEN 3 = the sensitivity value to which the monitor is calibrated for radionuclide "i" (cpm per pCi/cm 3 ), use the default value from Table 2.2 if radionuclide specific sensitivities are not available bkg = background of the monitor (cpm) 2.2.2 Conservative Default Values A conservative alarm setpoint can be established, in lieu of the individual radionuclide evaluation based on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default values presented in Table 2.2. These values are based upon:

a) substitution of the maximum ventilation flow rate, I 2-3 REV. 9 12/02/2005

b) substitution of a radionuclide distribution 3 comprised of 95% Xe-133, 2%

Xe-135, 1%Xe-133m, 1% Kr-88 and 1% Kr-85; and, c) application of an administrative multiplier of 0.5 to conservatively assure that any simultaneous releases do not exceed the maximum allowable release rate.

For this radionuclide distribution, the alarm setpoint based on the total body dose rate is more restrictive than the corresponding setpoint based on the skin dose rate. The resulting conservative, default setpoints are presented in Table 2.2.

2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 2.3.1 Site Boundary Dose Rate - Noble Gases.

ODCM Specification 3.4.1.a limits the dose rate at the site boundary due to noble gas l releases to < 500 mrem/yr to the total body, and < 3000 mrem/yr to the skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded. In the event any gaseous releases from the station results in the alarm setpoints being exceeded, an evaluation of the unrestricted area dose rate resulting from the release may be performed using the following equations:

Dt =X/QxE K-xQi (2.4) and D,=X/IQxE(L +1.1M-)xQi (2.5) where:

D t = total body dose rate (mrem/yr)

D. = skin dose rate (mrem/yr) 71Q = atmospheric dispersion for direct exposure to noble gas at the controlling site boundary (sec/M3 , from Table 2.3)

Qi = average release rate of radionuclide "i" over the release period under evaluation (gCi/sec)

'Adopted from ANSI N237-1976/ANS-18.1, Source Term Specifications, Table 6.

2-4 REV. 9 12/02/2005

Ki = total body dose conversion factor for noble gas radionuclide "i" (mrem/yr per pCi/M3 , from Table 2.1)

L; = beta skin dose conversion factor for noble gas radionuclide "i" (mrem/yr per gCi/m3 , from Table 2.1)

Mi = gamma air dose conversion factor for noble gas radionuclide "i" (mrad/yr per pCi/M3 , from Table 2.1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)

Actual meteorological conditions concurrent with the release period or the default, annual average dispersion parameters as presented in Table 2.3 may be used for evaluating the gaseous effluent dose rate.

2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates ODCM Specification 3.4.1.b limits the dose rate to < 1500 mrem/yr to any organ for I-131, I-133, tritium and particulates with half-lives greater than 8 days. To demonstrate compliance with this limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period for continuous releases (e.g.,

nominally once per 7 days) and for batch releases on the time period over which any batch release is to occur. The following equation may be used for the dose rate evaluation:

Do = XY Rix i (2.6) where:

DO = average organ dose rate over the sampling time period (mremlyr)

X/Q = atmospheric dispersion to the controlling site boundary for the inhalation pathway (sec/m 3 , from Table 2.3)

R. = dose parameter for radionuclide "i", (mrem/yr per pLCi/i 3 ) for the child inhalation pathway from Table 2.6 Qi = average release rate over the appropriate sampling period and analysis frequency for radionuclide i, 1-131, 1-133, tritium or other radionuclide in particulate form with half-life greater than 8 days (pCilsec)

By substituting 1500 mrem/yr for Do solving forQj, an allowable release rate for I-131 can be determined. Based on the annual average meteorological dispersion (see Table 2.3) and the most limiting potential pathway, age group and organ (inhalation pathway, child thyroid - R1 = 1.62E+07 mrem/yr per gCi/m3) the allowable release rate for 1-131 is 6.43 gCi/sec. An added conservatism factor of 0.25 has been included in this calculation to account for any potential dose contribution from other radioactive particulate material.

2-5 REV. 9 12102/2005

For a 7-day period, which is the nominal sampling and analysis frequency for 1-131, the cumulative allowable release is 3.9 Ci. Therefore, as long as the 1-131 releases in any 7-day period do not exceed 3.9 Ci, no additional analyses are needed to verify compliance with the ODCM Specification 3.4.1.b limits on allowable release rate. I 2.4 Gaseous Effluent Dose Calculations - 10 CFR 50 2.4.1 Unrestricted Area Dose - Noble Gases ODCM Specification 3.4.2 requires a periodic assessment of releases of noble gases to evaluate I compliance with the quarterly dose limits of (< 5 mrad, gamma-air and < 10 mrad, beta-air) and the calendar year limits (< 10 mrad, gamma-air and < 20 mrad, beta-air). The following equations may be used to calculate the gamma-air and beta-air doses:

-d DY = 3.17E - 08x XzQ x E (Mix Qi) (2.7) and D = 3.17E-08xxIQxX(NixQi) (2.8) where:

= air dose due to gamma emissions for noble gas radionuclides (mrad)

= air dose due to beta emissions for noble gas radionuclides (mrad)

XIQ = atmospheric dispersion to the controlling site boundary (sec/m3, from Table 2.3)

Qi = cumulative release of noble gas radionuclide "i" over the period of interest (COi)

= air dose factor due to gamma emissions from noble gas radionuclide "i" (mrad/yr per pCi/m3 from Table 2.1)

N2 = air dose factor due to beta emissions from noble gas radionuclide "i" (mrad/yr per tLCi/m3, Table 2.1) 3.17E-08 = conversion factor (yr/sec) 2-6 REV. 9 12/02/2005

In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculational equation may be used for verifying compliance with the dose limits of ODCM Specification 3.4.2. (Refer to Appendix B for the derivation and justification for this simplified method.)

DI = 3.7Eo-s XX/Q x Meff X Q (2.9) 0.50 and Dp = 3 . 7 E 0 8 xx/QxNfxEQi (2.10) 0.50 where:

3 Muff = 5.3E+02 effective gamma-air dose factor (mradlyr per gCi/m )

Nff = 1.LE+03 effective beta-air dose factor (mrad/yr per pCi/M3) 0.50 = conservatism factor Actual meteorological conditions concurrent with the release period or the default, annual average dispersion parameters as presented in Table 2.3, may be used for the evaluation of the gamma-air and beta-air doses.

2.4.2 Unrestricted Area Dose - Radioiodine and Particulates Per the requirements of ODCM Specification 3.4.3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit (< 7.5 mrem) and calendar year limit (< 15 mrem) to any organ. The following equation may be used to evaluate the maximum organ dose due to releases of 1-131, 1-133, tritium and particulates with half-lives greater than 8 days:

Daop=3.17E-08XWxSFpx (RiXQi) (2.11) where:

Diap = dose or dose commitment for age group "a" to organ "o", including the total body, via pathway "p" from 1-131, 1-133, tritium and radionuclides in particulate form with half-life greater than eight days (mrem)

W = atmospheric dispersion parameter to the controlling location(s) as identified in Table 2.3 X/Q = atmospheric dispersion for inhalation pathway and H1-3 dose contribution via other pathways (sec/M3) 2-7 REV. 9 12/02/2005

D/Q = atmospheric deposition for vegetation, milk and ground plane exposure pathways (1/m2)

Ri = dose factor for radionuclide "i", (rnrem/yr per glCi/m 3 ) or (m2 - mrem/yr per ILCi/sec) from Table 2.4 through 2.15 for each age group "a" and the applicable pathway "p" as identified in Table 2.3. Values for R; were derived in accordance with the methods described in NUREG-0133.

Qj = cumulative release over the period of interest for radionuclide "i" -- I-131 or radioactive material in particulate form with half-life greater than 8 days (Coi).

SFp = seasonal correction factor to account for the fraction of the period that the applicable exposure pathway does exist.

1) For milk and vegetation exposure pathways:

1. of months in the period that grazing occurs total # of months in period

= 0.5 for annual calculations

2) For inhalation and ground plane exposure pathways: = 1.0 In lieu of the individual radionuclide (1-131 and particulates) dose assessment as presented above, the following simplified dose calculational equation may be used for verifying compliance with the dose limits of ODCM Specification 3.4.3.

Dgnx =3.17E-08xWxSFpxR -131XXQi (2.12) where:

D~x = maximum organ dose (mrem)

RI-131 = 1-131 dose parameter for the thyroid for the identified controlling pathway

= 1.05E+12, infant thyroid dose parameter with the grass-cow-milk pathway controlling (M2 - mrem/yr per liCi/sec)

The ground plane exposure and inhalation pathways need not be considered when the above-simplified calculational method is used because of the overall negligible contribution of these pathways to the total thyroid dose. It is recognized that for some particulate radionuclides (e.g.

Co-60 and Cs-137), the ground plane exposure pathway may represent a higher dose contribution than either the vegetation or grass-cow-milk pathway. However, use of the I-131 thyroid dose parameter for all radionuclides will maximize the organ dose calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than 1-131 for the thyroid via the grass-cow-milk pathway.

2-8 REV. 9 12/02/2005

6- The location of exposure pathways and the maximum organ dose calculation may be based on the available pathways in the surrounding environment of Kewaunee as identified by the annual land-use census, see ODCM Specification 3.6.2. Otherwise, the l

_ dose will be evaluated based on the predetermined controlling pathways as identified in Table 2.3.

2.5 Gaseous Effluent Dose Proiection ODCM Specification 3.4.4 requires that the Ventilation Exhaust Treatment System be used to reduce radioactive material levels prior to discharge when projected doses exceed one-half the annual design objective rate in any calendar quarter, i.e., exceeding:

• 0.62 mrad/quarter, gamma air,

• 1.25 mradlquarter, beta air, or

• 0.94 mrem/quarter, maximum organ.

The applicable gaseous release sources and processing systems are as delineated in Figure 2.

Dose projections are performed at least once per 31 days by the following equations:

Dyp = Dyx(91 +d) (2.13)

Dop = DO X (91 + d) (2.14)

Damxp =-:Dnmx x(91 +d) (2.15) 6* where:

Dyp = gamma air dose projection for current calendar quarter (mrad)

DY = gamma air dose to date for current calendar quarter as determined by equation (2.7) or (2.9) (mrad)

Do~p = beta air dose projection for current calendar quarter (mrad)

Da = beta air dose to date for current calendar quarter as determined by equation (2.8) or (2.10) (mrad)

Dmiaxp = maximum organ dose projection for current calendar quarter (mrem)

DXx = maximum organ dose to date for current calendar quarter as determined by equation (2.11) or (2.12) (mrem) d = number of days to date in current calendar quarter 91 = number of days in a calendar quarter 2-9 REV. 9 12/02/2005

2.6 Environmental Radiation Protection Standards 40 CFR 190 For the purpose of implementing ODCM Specification 3.5 on the EPA environmental radiation protection standard and Technical Specification 6.9.b.2 on reporting requirements, dose calculations may be performed using the above equations with the substitution of average or actual meteorological parameters for the period of interest and actual applicable pathways. Any exposure attributable to on-site sources will be evaluated based on the results of the environmental monitoring program (TLD measurements) or by calculational methods. NUREG-0543 describes acceptable methods for demonstrating compliance with 40 CFR Part 190 when radioactive effluents exceed the Appendix I portion of the specifications.

2.7 Incineration of Radioactively Contaminated Oil During plant operation, radioactively contaminated oils are generated from various pieces of equipment operating in the plant. The largest source of contaminated oil is the reactor coolant pump lubricating oil, which is periodically changed for preventive maintenance reasons. 10 CFR Part 20 allows licensees to incinerate radioactively contaminated oils on site provided that the total radioactive effluents from the facility conform to the requirements of 10 CFR Part 50, Appendix L Radioactively contaminated oil, which is designated for incineration, will be collected in containers, which are uniquely serialized such that the contents can be identified and tracked.

Each container will be sampled and analyzed for radioactivity. The isotopic concentrations will be recorded for each container.

The heating boiler will be utilized to incinerate the radioactively contaminated oil collected on site. A gaseous radwaste effluent dose calculation, as prescribed in Section 2.3 of the ODCM, will be performed to insure that the limits established by ODCM Specifications 3.4.1, 3.4.2 and 3.4.3 are not exceeded. Release of the activity is assumed to occur at the time the contaminated oil is transferred into the heating boiler fuel oil storage tank and will be accounted for using established plant procedures. This will be valid for an assumed release from the fuel oil storage tank vent, fill piping, or from the boiler exhaust stack. See Figure 3 for a description of the heating boiler fuel oil system.

2.8 Total Dose The purpose of this section is to describe the method used to calculate the cumulative dose contributions from liquid and gaseous effluents in accordance with KPS Technical Specifications for total dose. This method can also be used to demonstrate compliance with the Environmental Protection Agency (EPA) 40CFR190, "Environmental Standards for the Uranium Fuel Cycle".

Compliance with the KPS Technical Specification dose objectives for the maximum individual demonstrates compliance with the EPA limits to any member-of the public, since the design dose objectives from IOCFR50, Appendix I are much lower than the 40CFR190 dose limits to the general public. With the calculated doses from the releases of radioactive materials in liquid or gaseous effluents exceeding twice the limits outlined in ODCM Specifications 3.3.2, 3.4.2, and 3.4.3, a special analysis shall be performed. The purpose of this analysis is to demonstrate if the total dose to any member of the public (real individual) from all uranium fuel cycle sources (including direct radiation contributions from the reactor unit, from outside storage areas and from all real pathways) is limited to less than or equal to 25 mrem per year to the total body or any organ, except the thyroid, which is limited to 75 mrem per year.

2-10 REV. 9 12/02/2005

If required, the total dose to a member of the public will be calculated for all significant effluent release points for all real pathways including direct radiation. Effluent releases from Point Beach Nuclear Plant must also be considered due to its proximity. Calculations will be based on the equations in Sections 1.4, 2.4.1, and 2.4.2, with the exception that usage factors and other site specific parameters may be modified using more realistic assumptions, where appropriate.

The direct radiation component from the facility can be determined using environmental TLD results. These results will be corrected for natural background and for actual occupancy time of any areas accessible to the general public at the location of maximum direct radiation. It is recognized that by including the results from the environmental TLDs into the sum of total dose component, the direct radiation dose may be overestimated. The TLD measurements may include the exposure from noble gases, ground plane deposition, and shoreline deposition, which have already been included in the summation of the significant dose pathways to the general public. However, this conservative method can be used, if required, as well as any other method for estimating the direct radiation dose from contained radioactive sources within the facility.

The methodology used to incorporate the direct radiation component into total dose estimates will be outlined whenever total doses are reported.

Therefore, the total dose will be determined based on the most realistic site specific data and parameters to assess the real dose to any member of the public.

2-11 REV. 9 12/02/2005

toAtmosphere FIgure 2 Gaseous Radioactive Effluent Flow Diagram

• The shield building ventilation and special ventiation are ESF systems and are not part of the normal ef Iuent processing sy stem. They are included I or completeness only.

• • The containment air sampler (R-11) and radiation monitor (R-12) can also be aligned as needed forsamplhngconlainmentvent.

2-12 REV. 9 12/02/2005 I ' ( I I (I I I I I I I t [ I

I Ir r r I . r r r r Figure 3 Simplified Heating Boiler Fuel Oil Piping System exhoustutack rFOr Iplhntwi Y vent cap with flame arrestor

- 11 unit

/ -A 3' rod &sample unit i,

K Valve Rack F j- II___ L ----- J Heetrng Bonler Fuel Oil Storage Tank heun bofler 30,000 lotllons fuel oan pumps 2-13 REV. 9 12/02/2005

Table 2.1 Dose Factors for Noble Gases Total Body Skin Dose Gamma Air Beta Air Dose Factor Factor Dose Factor Dose Factor K, L; ml N1 Radionuclide (mremlyr (mrem/yr (mrad/yr (mrad/yr

___________ per ILCi/m 3) per ILCi/m 3 ) per gCim3) per gCim 3)

Kr-83m 7.56E-02 ------ 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.O1E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m - 9.15E+01 4.76E+02 1.56E+02 1.l1IE+03 Xe-133m 2.5 1E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05SE+03 Xe-135m 3.12E+03 7.1 lE+02 3.36E+03 7.39E+02 Xe-135 1.8 1E+03 E.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 2-14 REV. 9 12/02/2005

2-15 REV. 9 12/02/2005

Table 2.3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations Atmospheric Dispersion ODCMI X/Q D/Q I-Specification Location Pathway(s) (sec/m 3 ) (1/m;)

3.4.1La site boundary noble gases 3.6E-06 N/A (1300 mi, N) direct exposure ___________

3.4.1.b site boundary inhalation 3.6E-06 N/A (1300 m, N) _________

3.4.2 site boundary gamma-air 3.6E-06 N/A 3..2(1300 m, N) beta-air___________

eii .inhalation, I 3.4.3 residence/dairy vegetation, milk and 5.6E-07 5.6E-09 (1 mile 'W) ground plane ____________

2-16 REV. 9 12/02/2005

Table 2.4 RI Inhalation Pathway Dose Factors - ADULT (mremlyr per IACVm 3)

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 1.26E+3 1.26E+3 1.26E+3 1.26E+3 1.26E+3 1.26E+3 C-14 1.82E+4 3.41 E+3 3.41 E+3 3.41 E+3 3.41 E+3 3.41 E+3 3.41 E+3 Na-24 1.02E+4 1.02E+4 1.02E+4 1.02E+4 1.02E+4 1.02E+4 1.02E+4 P-32 1.32E+6 7.71 E+4 - - 8.64E+4 5.01 E+4 Cr-51 - - 5.95E+1 2.28E+1 1.44E+4 3.32E+3 1.OOE+2 Mn-54 - 3.96E+4 - 9.84E+3 1.40E+6 7.74E+4 6.30E+3 Mn-56 - 1.24E+0 - 1.30E+O 9.44E+3 2.02E+4 1.83E-1 Fe-55 2.46E+4 1.70E+4 - 7.21 E+4 6.03E+3 3.94E+3 Fe-59 1.18E+4 2.78E+4 - 1.02E+6 1.88E+5 1.06E+4 Co-57 - 6.92E+2 - - 3.70E+5 3.14E+4 6.71 E+2 Co-58 - 1.58E+3 - - 9.28E+5 1.06E+5 2.07E+3 Co-60 - 1.15E+4 - 5.97E+6 2.85E+5 1.48E+4 Ni-63 4.32E+5 3.14E+4 - - 1.78E+5 1.34E+4 1.45E+4 Ni-65 1.54E+0 2.10E-1 - - 5.60E+3 1.23E+4 9.12E-2 Cu-64 - 1.46E+0 - 4.62E+0 6.78E+3 4.90E+4 6.15E-1 Zn-65 3.24E+4 1.03E+5 - 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Zn-69 3.38E-2 6.51 E-2 - 4.22E-2 9.20E+2 1.63E+1 4.52E-3 Br-82 - - - - - 1.04E+4 1.35E+4 Br-83 - - - - - 2.32E+2 2.41 E+2 Br-84 - - - - - 1.64E-3 3.13E+2 Br-85 - - - - - - 1.28E+1 Rb-86 - 1.35E+5 - - - 1.66E+4 5.90E+4 Rb-88 - 3.87E+2 - - - 3.34E-9 1.93E+2 Rb-89 - 2.56E+2 - - - - 1.70E+2 Sr-89 3.04E+5 - - - 1.40E+6 3.50E+5 8.72E+3 Sr-90 9.92E+7 - - - 9.60E+6 7.22E+5 6.1OE+6 Sr-91 6.19E+1 - - - 3.65E+4 1.91 E+5 2.50E+O Sr-92 6.74E+0 - - - 1.65E+4 4.30E+4 2.91 E-1 Y-90 2.09E+3 - - - 1.70E+5 5.06E+5 5.61 E+1 Y-91m 2.61 E-1 - - - 1.92E+3 1.33E+O 1.02E-2 Y-91 4.62E+5 - - - 1.70E+6 3.85E+5 1.24E+4 Y-92 1.03E+1 - - - 1.57E+4 7.35E+4 3.02E-1 YV93 9.44E+1 - - - 4.85E+4 4.22E+5 2.61 E+0 Zr-95 1.07E+5 3.44E+4 - 5.42E+4 1.77E+6 1.50E+5 2.33E+4 Zr-97 9.68E+1 1.96E+1 - 2.97E+1 7.87E+4 5.23E+5 9.04E+0 Nb-95 1.41 E+4 7.82E+3 - 7.74E+3 5.05E+5 1.04E+5 4.21 E+3 Nb-97 2.22E-1 5.62E-2 - 6.54E-2 2.40E+3 2.42E+2 2.05E-2 Mo-99 - 1.21EE+2 - 2.91 E+2 9.12E+4 2.48E+5 2.30E+1 Tc-99m 1.03E-3 2.91 E-3 - 4.42E-2 7.64E+2 4.16E+3 3.70E-2 Tc-101 4.18E-5 6.02E-5 - 1.08E-3 3.99E+2 - 5.90E-4 2-17 REV. 9 12/02/2005

Table 2.4 RI Inhalation Pathway Dose Factors - ADULT (mrem/yr per jICVm3 )

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-103 1.53E+3 - - 5.83E+3 5.05E+5 1.10E+5 6.58E+2 Ru-105 7.90E-1 - - 1.02E+0 1.10E+4 4.82E+4 3.i1 E-1 Ru-i06 6.91 E+4 - - 1.34E+5 9.36E+6 9.12E+5 8.72E+3 Rh-103m -

Rh-106 - - - - -

Ag-110m 1.08E+4 1.00E+4 - 1.97E+4 4.63E+6 3.02E+5 5.94E+3 Sb-124 3.12E+4 5.89E+2 7.55E+1 - 2.48E+6 4.06E+5 1.24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+1 - 1.74E+6 1.01 E+5 1.26E+4 Te-125m 3.42E+3 1.58E+3 1.05E+3 1.24E+4 3.14E+5 7.06E+4 4.67E+2 Te-127m 1.26E+4 5.77E+3 3.29E+3 4.58E+4 9.60E+5 1.50E+5 1.57E+3 Te-127 1.40E+0 6.42E-1 1.06E+0 5.10E+0 6.51 E+3 5.74E+4 3.10E-1 Te-129m 9.76E+3 4.67E+3 3.44E+3 3.66E+4 1.16E+6 3.83E+5 1.58E+3 Te-129 4.98E-2 2.39E-2 3.90E-2 1.87E-1 1.94E+3 1.57E+2 1.24E-2 Te-131m 6.99E+1 4.36E+1 5.50E+1 3.09E+2 1.46E+5 5.56E+5 2.90E+1 Te-131 1.11E-2 5.95E-3 9.36E-3 4.37E-2 1.39E+3 1.84E+1 3.59E-3 Te-132 2.60E+2 2.15E+2 1.90E+2 1.46E+3 2.88E+5 5.10E+5 1.62E+2 1-130 4.58E+3 1.34E+4 1.14E+6 2.09E+4 7.69E+3 5.28E+3 1-131 2.52E+4 3.58E+4 1.19E+7 6.13E+4 - 6.28E+3 2.05E+4 1-132 1.16E+3 3.26E+3 1.14E+5 5.18E+3 - 4.06E+2 1.16E+3 1-133 8.64E+3 1.48E+4 2.15E+6 2.58E+4 - 8.88E+3 4.52E+3 1-134 6.44E+2 1.73E+3 2.98E+4 2.75E+3 1.01E+0 6.15E+2 1-135 2.68E+3 6.98E+3 4.48E+5 1.11 E+4 - 5.25E+3 2.57E+3 Cs-134 3.73E+5 8.48E+5 - 2.87E+5 9.76E+4 1.04E+4 7.28E+5 Cs-1 36 3.90E+4 1.46E+5 - 8.56E+4 1.20E+4 1.17E+4 1.10E+5 Cs-1 37 4.78E+5 6.21 E+5 - 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Cs-1 38 3.31 E+2 6.21 E+2 - 4.80E+2 4.86E+1 1.86E-3 3.24E+2 Ba-1 39 9.36E-1 6.66E-4 - 6.22E-4 3.76E+3 8.96E+2 2.74E-2 Ba-1 40 3.90E+4 4.90E+1 - 1.67E+1 1.27E+6 2.18E+5 2.57E+3 Ba-1 41 1.OOE-1 7.53E-5 - 7.00E-5 1.94E+3 1.16E-7 3.36E-3 Ba-142 2.63E-2 2.70E-5 - 2.29E-5 1.19E+3 - 1.66E-3 La-140 3.44E+2 1.74E+2 - - 1.36E+5 4.58E+5 4.58E+1 La-142 6.83E-1 3.10E-1 - - 6.33E+3 2.11E+3 7.72E-2 Ce-141 1.99E+4 1.35E+4 - 6.26E+3 3.62E+5 1.20E+5 1.53E+3 Ca-143 1.86E+2 1.38E+2 - 6.08E+1 7.98E+4 2.26E+5 1.53E+1 Ce-144 3.43E+6 1.43E+6 - 8.48E+5 7.78E+6 8.16E+5 1.84E*5 Pr-1 43 9.36E+3 3.75E+3 - 2.16E+3 2.81E+5 2.00E+5 4.64E+2 Pr-144 3.01E-2 1.25E-2 - 7.05E-3 1.02E+3 2.15E-8 1.53E-3 Nd-147 5.27E+3 6.10E+3 - 3.56E+3 2.21E+5 1.73E+5 3.65E+2 W-187 8.48E+0 7.08E+0 - - 2.90E+4 1.55E+5 2.48E+O Np-239 2.30E+2 2.26E+1 - 7.00E+1 3.76E+4 1.19E+5 1.24E+1 2-18 REV. 9 12/02/2005

Table 2.5 RI Inhalation Pathway Dose Factors - TEEN (mrem/yr per ,uCVm 3 )

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 C-14 2.60E+4 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 Na-24 1.38E+4 1.38E+4 1.38E+4 1.38E+4 1.38E+4 1.38E+4 1.38E+4 P-32 1.89E+6 1.10E+5 - - - 9.28E+4 7.16E+4 Cr-61 - 7.50E+1 3.07E+1 2.10Ed+4 3.OOE+3 1.35E+2 Mn-54 - 5.11 E+4 - 1.27E+4 1.98E+6 6.68E+4 8.40E+3 Mn-56 - 1.70E+O - 1.79E+O 1.52E+4 5.74E+4 2.52E-1 Fe-55 3.34E+4 2.38E+4 - - 1.24E+5 6.39E+3 5.54E+3 Fe-59 1.59E+4 3.70E+4 - - 1.53E+6 1.78E+5 1.43E+4 Co-57 - 6.92E+2 - 5.86E+5 3.14E+4 9.20E+2 Co-58 - 2.07E+3 - 1.34E+6 9.52E+4 2.78E+3 Co-60 - 1.51 E+4 - - 8.72E+6 2.59E+5 1.98E+4 Ni-63 5.80E+5 4.34E+4 - 3.07E+5 1.42E+4 1.98E+4 Ni-65 2.18E+O 2.93E-1 - 9.36E+3 3.67E+4 1.27E-1 Cu-64 - 2.03E+O - 6.41 E+O 1.11 E+4 6.14E+4 8.48E-1 Zn-65 3.86E+4 1.34E+5 - 8.64E+4 1.24E+6 4.66E+4 6.24E+4 Zn-69 4.83E-2 9.20E-2 - 6.02E-2 1.58E+3 2.85E+2 6.46E-3 Br-82 - - - - - 1.82E+4 Br-83 - - - - - 3.44E+2 Br-84 - - - - - - 4.33E+2 Br-85 - - - - - 1.83E+1 Rb-86 - 1.90E+5 - - - 1.77E+4 8.40E+4 Rb-88 - 5.46E+2 - - - 2.92E-5 2.72E+2 Rb-89 - 3.52E+2 - - - 3.38E-7 2.33E+2 Sr-89 4.34E+5 - - - 2.42E+6 3.71 E+5 1.25E+4 Sr-90 1.08E+8 - - - 1.65E+7 7.65E+5 6.68E+6 Sr-91 8.80E+1 - - - 6.07E+4 2.59E+5 3.51 E+O Sr-92 9.52E+O - 2.74E+4 1.19E+5 4.06E-1 Y-90 2.98E+3 - - - 2.93E+5 5.59E+5 8.OOE+1 Y-91 rn 3.70E-1 - - - 3.20E+3 3.02E+1 1.42E-2 Y-91 6.61 E+5 - - - 2.94E+6 4.09E+5 1.77E+4 Y-92 1.47E+1 - - - 2.68E+4 1.65E+5 4.29E-1 Y-93 1.35E+2 - - - 8.32E+4 5.79E+5 3.72E+O Zr-95 1.46E+5 4.58E+4 - 6.74E+4 2.69E+6 1.49E+5 3.15E+4 Zr-97 1.38E+2 2.72E+1 4.12E+1 1.30E+5 6.30E+5 1.26E+1 Nb-95 1.86E+4 1.03E+4 1.OOE+4 7.51 E+5 9.68E+4 5.66E+3 Nb-97 3.14E-1 7.78E-2 - 9.12E-2 3.93E+3 2.17E+3 2.84E-2 Mo-99 - 1.69E+2 - 4.11 E+2 1.54E+5 2.69E+5 3.22E+1 Tc-99m 1.38E-3 3.86E-3 - 5.76E-2 1.15E+3 6.13E+3 4.99E-2 Tc-1 01 5.92E-5 8.40E-5 - 1.52E-3 6.67E+2 8.72E-7 8.24E-4 2-19 REV. 9 12/02/2005

Table 2.5 RI Inhalation Pathway Dose Factors - TEEN (mrem/yr per IICVm 3 )

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-103 2.1 OE+3 - - 7.43E+3 7.83E+5 1.09E+5 8.96E+2 Ru-105 1.12E+0 - - 1.41 E+O 1.82E+4 9.04E+4 4.34E-1 Ru-1 06 9.84E+4 - - 1.90E+5 1.61 E+7 9.60E+5 1.24E+4 Rh-103m - - - - - - -

Rh-106 - - - - - -

Ag-1I Om 1.38E+4 1.31 E+4 - 2.50E+4 6.75E+6 2.73E+5 7.99E+3 Sb-1 24 4.30E+4 7.94E+2 9.76E+1 - 3.85E+6 3.98E+5 1.68E+4 Sb-125 7.38E+4 8.08E+2 7.04E+1 - 2.74E+6 9.92E+4 1.72E+4 Te-125m 4.88E+3 2.24E+3 1.40E+3 - 5.36E+5 7.50E+4 6.67E+2 Te-127m 1.80E+4 8.16E+3 4.38E+3 6.54E+4 1.66E+6 1.59E+5 2.18E+3 Te-127 2.01 E+O 9.12E-1 1.42E+0 7.28E+O 1.12E+4 8.08E+4 4.42E-1 Te-129m 1.39E+4 6.58E+3 4.58E+3 5.19E+4 1.98E+6 4.05E+5 2.25E+3 Te-1 29 7.1 OE-2 3.38E-2 5.18E-2 2.66E-1 3.30E+3 1.62E+3 1.76E-2 Te-131m 9.84E+1 6.01 E+1 7.25E+1 4.39E+2 2.38E+5 6.21 E+5 4.02E+1 Te-131 1.58E-2 8.32E-3 1.24E-2 6.18E-2 2.34E+3 1.51E+1 5.04E-3 Te-1 32 3.60E+2 2.90E+2 2.46E+2 1.95E+3 4.49E+5 4.63E+5 2.19E+2 1-130 6.24E+3 1.79E+4 1.49E+6 2.75E+4 - 9.12E+3 7.17E+3 1-131 3.54E+4 4.91 E+4 1.46E+7 8.40E+4 - 6.49E+3 2.64E+4 1-132 1.59E+3 4.38E+3 1.51 E+5 6.92E+3 - 1.27E+3 1.58E+3 1-133 1.22E+4 2.05E+4 2.92E+6 3.59E+4 - 1.03E+4 6.22E+3 1-134 8.88E+2 2.32E+3 3.95E+4 3.66E+3 - 2.04E+1 8.40E+2 1-135 3.70E+3 9.44E+3 6.21 E+5 1.49E+4 - 6.95E+3 3.49E+3 Cs-1 34 5.02E+5 1.1 3E+6 - 3.75E+5 1.46E+5 9.76E+3 5.49E+5 Cs-136 5.15E+4 1.94E+5 - 1.10E+5 1.78E+4 1.09E+4 1.37E+5 Cs-1 37 6.70E+5 8.48E+5 - 3.04E+5 1.21 E+5 8.48E+3 3.11 E+5 Cs-1 38 4.66E+2 8.56E+2 - 6.62E+2 7.87E+1 2.70E-1 4.46E+2 Ba-1 39 1.34E+0 9.44E-4 - 8.88E-4 6.46E+3 6.45E+3 3.90E-2 Ba-140 5.47E+4 6.70E+1 - 2.28E+1 2.03E+6 2.29E+5 3.52E+3 Ba-141 1.42E-1 1.06E-4 - 9.84E-5 3.29E+3 7.46E-4 4.74E-3 Ba-142 3.70E-2 3.70E-5 - 3.14E-5 1.91 E+3 - 2.27E-3 La-140 4.79E+2 2.36E+2 - - 2.14E+5 4.87E+5 6.26E+1 La-i 42 9.60E-1 4.25E-1 - - 1.02E+4 1.20E+4 1.06E-1 Ce-141 2.84E+4 1.90E+4 - 8.88E+3 6.14E+5 1.26E+5 2.17E+3 Ce-143 2.66E+2 1.94E+2 - 8.64E+1 1.30E+5 2.55E+5 2.16E+1 Ce-144 4.89E+6 2.02E+6 - 1.21 E+6 1.34E+7 8.64E+5 2.62E+5 Pr-i 43 1.34E+4 5.31 E+3 - 3.09E+3 4.83E+5 2.14E+5 6.62E+2 Pr-144 4.30E-2 1.76E-2 - 1.01E-2 1.75E+3 2.35E-4 2.18E-3 Nd-147 7.86E+3 8.56E+3 - 5.02E+3 3.72E+5 1.82E+5 5.13E+2 W-1 87 1.20E+1 9.76E+0 - - 4.74E+4 1.77E+5 3.43E+O Np-239 3.38E+2 3.19E+1 - 1.OOE+2 6.49E+4 1.32E+5 1.77E+1 2-20 REV. 9 12/02/2005

Table 2.6 RI Inhalation Pathway Dose Factors - CHILD (mrenmyr per guCVm 3)

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 1.12E+3 1.12E+3 1.12E+3 1.12E+3- 1.12E+3 1.12E+3 C-14 3.59E+4 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 Na-24 1.61 E+4 1.61 E+4 1.61 E+4 1.61 E+4 1.61 E+4 1.61 E+4 1.61 E+4 P-32 2.60E+6 1.14E+5 - - - 4.22E+4 9.88E+4 Cr-51 - - 8.55E+1 2.43E+1 1.70E+4 1.08E+3 1.54E+2 Mn-54. - 4.29E+4 - 1.OOE+4 1.58E+6 2.29E+4 9.51 E+3 Mn-56 - 1.66E+0 - 1.67E+0 1.31E+4 1.23E+5 3.12E-1 Fe-55 4.74E+4 2.52E+4 - 1.11 E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 - 1.27E+6 7.07E+4 1.67E+4 Co-57 - 9.03E+2 - 5.07E+5 1.32E+4 1.07E+3 Co-58 - 1.77E+3 - - 1.11E+6 3.44E+4 3.16E+3 Co-60 - 1.31 E+4 - - 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21 E+5 4.63E+4 - 2.75E+5 6.33E+3 2.80E+4 Ni-65 2.99E+O 2.96E-1 - - 8.18E+3 8.40E+4 1.64E-1 Cu-64 - 1.99E+O - 6.03E+0 9.58E+3 3.67E+4 1.07E+0 Zn-65 4.26E+4 1.13E+5 - 7.14E+4 9.95E+5 1.63E+4 7.03E+4 Zn-69 6.70E-2 9.66E-2 - 5.85E-2 1.42E+3 1.02E+4 8.92E-3 Br-82 - - - - - 2.09E+4 Br-83 - - - - - - 4.74E+2 Br-84 - - - - - - 5.48E+2 Br-85 - - - - - 2.53E+1 Rb-86 - 1.98E+5 - - - 7.99E+3 1.14E+5 Rb-88 - 5.62E+2 - - - 1.72E+1 3.66E+2 Rb-89 - 3.45E+2 - - 1.89E+0 2.90E+2 Sr-89 5.99E+5 - - - 2.16E+6 1.67E+5 1.72E+4 Sr-90 1.01 E+8 - - - 1.48E+7 3.43E+5 6.44E+6 Sr-91 1.21 E+2 - - - 5.33E+4 1.74E+5 4.59E+0 Sr-92 1.31 E+1 - - - 2.40E+4 2.42E+5 5.25E-1 Y-90 4.11 E+3 - - - 2.62E+5 2.68E+5 1.11 E+2 Y-91 m 5.07E-1 - - - 2.81 E+3 1.72E+3 1.84E-2 Y-91 9.14E+5 - - - 2.63E+6 1.84E+5 2.44E+4 Y-92 2.04E+1 - - - 2.39E+4 2.39E+5 5.81 E-1 Y-93 1.86E+2 - - - 7.44E+4 3.89E+5 5.11 E+O Zr-95 1.90E+5 4.18E+4 - 5.96E+4 2.23E+6 6.11 E+4 3.70E+4 Zr-97 1.88E+2 2.72E+1 - 3.89E+1 1.13E+5 3.51 E+5 1.60E+1 Nb-95 2.35E+4 9.18E+3 - 8.62E+3 6.14E+5 3.70E+4 6.55E+3 Nb-97 4.29E-1 7.70E-2 - 8.55E-2 3.42E+3 2.78E+4 3.60E-2 Mo-99 - 1.72E+2 - 3.92E+2 1.35E+5 1.27E+5 4.26E+1 Tc-99m 1.78E-3 3.48E-3 - 5.07E-2 9.51 E+2 4.81 E+3 5.77E-2 Tc-101 8.10E-5 8.51 E-5 1.45E-3 5.85E+2 1.63E+1 1.08E-3 2-21 REV. 9 12/02/2005

Table 2.6 RI Inhalation Pathway Dose Factors - CHILD (mremlyr per jCvm3 )

Nucilde Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-1 03 2.79E+3 - - 7.03E+3 6.62E+5 *4.48E+4 1.07E+3 Ru-1 05 1.53E+0 - - 1.34E+0 1.59E+4 9.95E+4 5.55E-1 Ru-106 1.36E+5 - - 1.84E+5 1.43E+7 4.29E+5 1.69E+4 Rh-103m - - - - - -

Rh-106 - -

Ag-1i1m . 1.69E+4 1.14E+4 - 2.12E+4 5.48E+6 1.OOE+5 9.14E+3 Sb-i124 5.74E+4 7.40E+2 1.26E+2 - 3.24E+6 1.64E+5 2.OOE-4 Sb-i125 9.84E+4 7.59E+2 9.10E+1 - 2.32E+6 4.03E+4 2.07E+4 Ta-125m 6.73E+3 2.33E+3 1.92E+3 - 4.77E+5 3.38E+4 9.14E+2 Te-127m 2.49E+4 8.55E+3 6.07E+3 6.36E+4 1.48E+6 7.14E+4 3.02E*3 Te-127 2.77E+0 9.51 E-1 1.96E+0 7.07E+0 1.OOE+4 5.62E+4 6.11 E-1 Te-129m 1.92E+4 6.85E+3 6.33E+3 5.03E+4 1.76E+6 1.82E+5 3.04Ei3 To-129 9.77E-2 3.50E-2 7.14E-2 2.57E-1 2.93E+3 2.55E+4 2.38E-2 Te-131m 1.34E+2 5.92E+1 9.77E+1 4.00E+2 2.06E+5 3.08E+5 5.07E+l To-131 2.17E-2 8.44E-3 1.70E-2 5.881E-2 2.05E+3 1.33E+3 6.59E-3 Te-1 32 4.81 E+2 2.72E+2 3.17E+2 1.77E+3 3.77E+5 1.38E+5 2.63E+2 1-130 8.18E+3 1.64E+4 1.85E+6 2.45E+4 - 5.1 1E+3 8.44E+3 1-131 4.81 E+4 4.81 E+4 1.62E+7 7.88E+4 - 2.84E+3 2.73E+4 1-132 2.12E+3 4.07E+3 1.94E+5 6.25E+3 - 3.20E+3 1.88E+3 1-133 1.66E+4 2.03E+4 3.85E+6 3.38E+4 - 5.48E+3 7.70E+3 1-134 1.17E+3 2.16E+3 5.07E+4 3.30E+3 - 9.55E+2 9.95E+2 1-135 4.92E+3 8.73E+3 7.92E+5 1.34E+4 - 4.44E+3 4.14E+3 Cs-1 34 6.51 E+5 1.01 E+6 - 3.30E+5 1.21 E+5 3.85E+3 2.25E+5 Cs-136 6.51E+4 1.71E+5 - 9.55E+4 1.45E+4 4.18E+3 1.16E+5 Cs-137 9.07E+5 8.25E+5 - 2.82E+5 1.04E+5 3.62E+3 1.28E+5 Cs-1 38 6.33E+2 8.40E+2 - 6.22E+2 6.81 E+1 2.70E+2 5.55E+2 Ba-139 1.84E+0 9.84E-4 - 8.62E-4 5.77E+3 5.77E+4 5.37E-2 Ba-140 7.40E+4 6.48E+1 - 2.1 1E+1 1.74E+6 1.02E+5 4.33E+3 Ba-1 41 1.96E-1 1.09E-4 - 9.47E-5 2.92E+3 2.75E+2 6.36E-3 Ba-142 5.OOE-2 3.60E-5 - 2.91 E-5 1.64E+3 2.74E+0 2.79E-3 La-1 40 6.44E+2 2.25E+2 - - 1.83E+5 2.26E+5 7.55E+1 La-142 1.30E+0 4.1 1E-1 - - 8.70E+3 7.59E+4 1.29E-1 Ce-1 41 3.92E+4 1.95E+4 - 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-1 43 3.66E+2 1.99E+2 - 8.36E+1 1.1 5E+5 1.27E+5 2.87E+1 Ce-144 6.77E+6 2.12E+6 - 1.17E+6 1.20E+7 3.89E+5 3.61 E+5 Pr-143 1.85E+4 5.55E+3 - 3.OOE+3 4.33E+5 9.73E+4 9.14E+2 Pr-44 5.96E-2 1.85E-2 - 9.77E-3 1.57E+3 1.97E+2 3.OOE-3 Nd-147 1.08E+4 8.73E+3 - 4.81 E+3 3.28E+5 8.21 E+4 6.81 E+2 W-187 1.63E+1 9.66E+0 - - 4.11 E+4 9.10E+4 4.33E+0 Np-239 4.66E+2 3.34E+1 - 9.73E+1 5.81 E+4 6.40E+4 2.35E+1 2-22 REV. 9 12/02/2005

Table 2.7 RI Inhalation Pathway Dose Factors - INFANT (mremlyr per jCiIm 3 )

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 2.65E+4 5.31 E+3 5.31 E+3 5.31 E+3 5.31 E+3 5.31 E+3 5.31 E+3 Na-24 1.06E+4 1.06E+4 1.06E+4 1.06E+4 1.06E+4 1.06E+4 1.06E+4 P-32 2.03E+6 1.12E+5 - 1.61 E+4 7.74E+4 Cr-51 - - 5.75E+1 1.32E+1 1.28E+4 3.57E+2 8.95E+1 Mn-54 - 2.53E+4 - 4.98E+3 1.OOE+6 7.06E+3 4.98E+3 Mn-56 - 1.54E+O - 1.10E+0 1.25E+4 7.17E+4 2.21E-1 Fe-55 1.97E+4 1.17E+4 - 8.69E+4 1.09E+3 3.33E+3 Fe-59 1.36E+4 2.35E+4 - - 1.02E+6 2.48E+4 9.48E+3 Co-57 - 6.51 E+2 - - 3.79E+5 4.86E+3 6.41 E+2 Co-58 1.22E+3 - 7.77E+5 1.11 E+4 1.82E+3 Co-60 8.02E+3 - - 4.51 E+6 3.19E+4 1.18E+4 Ni-63 3.39E+5 2.04E+4 2.09E+5 2.42E+3 1.16E+4 Ni-65 2.39E+O 2.84E-1 - - 8.12E+3 5.01 E+4 1.23E-1 Cu-64 - 1.88E+O 3.98E+O 9.30E+3 1.50E+4 7.74E-1 Zn-65 1.93E+4 6.26E+4 - 3.25E+4 6.47E+5 5.14E+4 3.11 E+4 Zn-69 5.39E-2 9.67E-2 - 4.02E-2 1.47E+3 1.32E+4 7.18E-3 Br-82 - 32 - - - 1.33E+4 Br-83 3 - - - 2 - 3.81 E+2 Br-84 - - - - - 4.00E+2 Sr-85 1 - - - - 2.013 +1 Rb-86 - 1.90E+5 - - 2 3.04E+3 8.82E+4 Rb-88 - 5.57E+2 - - -E 3.39E+2 2.87E+2 Rb-89 - 3.21 E+2 --- 6.82E+1 2.06E+2 Sr-89 3.98E+5 - - - 2.03E+6 6.40E+4 1.14E+4 Sr-90 4.09E+7 - - - 1.12E+7 1.31E+5 2.59E+6 Sr-91 9.56E+1 - - - 5.26E+4 7.34E+4 3.46E+0 Sr-92 1.05E+1 - - 3 2.38E+4 1.40E+5 3.91 E-1 Y-90 3.29E+3 - - - 2.69E+5 1.04E+5 8.82E+1 Y-91 m 4.07E-1 - - - 2.79E+3 2.35E+3 1.39E-2 Y-91 5.88E+5 --- 2.45E+6 7.03E+4 1.57E+4 Y-92 1.64E+1 - - - 2.45E+4 1.27E+5 4.61 E-1 Y-93 1.50E+2 1 - - 7.64E+4 1.67E+5 4.07E+0 Zr-95 1.15E+5 2.79E+4 - 3.1 1E+4 1.75E+6 2.17E+4 2.03E+4 Zr-97 1.50E+2 2.56E+1 - 2.59E+1 1.10E+5 1.40E+5 1.17E+1 Nb-95 1.57E+4 6.43E+3 - 4.72E+3 4.79E+5 1.27E+4 3.78E+3 Nb-97 3.42E-1 7.29E-2 - 5.70E-2 3.32E+3 2.69E+4 2.63E-2 Mo-99 - 1.65E+2 - 2.65E+2 1.35E+5 4.87E+4 3.23E+1 Tc-99m 1.40E-3 2.88E-3 - 3.11 E-2 8.11 E+2 2.03E+3 3.72E-2 Tc-101 6.51 E-5 8.23E-5 - 9.79E-4 5.84E+2 8.44E+2 8.12E-4 2-23 REV. 9 12/02/2005

Table 2.7 RI Inhalation Pathway Dose Factors - INFANT (mrem/yr per glCVm 3 )

Nuclide Bone Liver Thyroid Kidney Lung Gl-LLI T.Body Ru-1 03 2.02E+3 - - 4.24E+3 5.52E+5 1.61 E+4 6.79E+2 Ru-105 1.22E+0 - 8.99E-1 1.57E+4 4.84E+4 4.10E-1 Ru-106 8.68E+4 - - 1.07E+5 1.16E+7 1.64E+5 1.09E+4 Rh-103m - - .

Rh-106 - - - - - -

Ag-11Om 9.98E+3 7.22E+3 - 1.09E+4 3.67E+6 3.30E+4 5.OOE+3 Sb-1 24 3.79E+4 5.56E+2 1.01 E+2 - 2.65E+6 5.91 E+4 1.20E+4 St-125 5.17E+4 4.77E+2 6.23E+1 - 1.64E+6 1.47E+4 1.09E+4 Te-125m 4.76E+3 1.99E+3 1.62E+3 - 4.47E+5 1.29E+4 6.58E+2 Te-127m 1.67E+4 6.90E+3 4.87E+3 3.75E+4 1.31 E+6 2.73E+4 2.07E+3 Te-1 27 2.23E+0 9.53E-1 1.85E+0 4.86E+0 1.03E+4 2.44E+4 4.89E-1 Te-129m 1.41 E+4 6.09E+3 5.47E+3 3.18E+4 1.68E+6 6.90E+4 2.23E+3 Te-129 7.88E-2 3.47E-2 6.75E-2 1.75E-1 3.OOE+3 2.63E+4 1.88E-2 Te-131m 1.07E+2 5.50E+1 8.93E+1 2.65E+2 1.99E+5 1.19E+5 3.63E+1 Te-1 31 1.74E-2 8.22E-3 1.58E-2 3.99E-2 2.06E+3 8.22E+3 5.OOE-3 Te-1 32 3.72E+2 2.37E+2 2.79E+2 1.03E+3 3.40E+5 4.41 E+4 1.76E+2 1-130 6.36E+3 1.39E+4 1.60E+6 1.53E+4 - 1.99E+3 5.57E+3 1-131 3.79E+4 4.44E+4 1.48E+7 5.18E+4 - 1.06E+3 1.96E+4 1-132 1.69E+3 3.54E+3 1.69E+5 3.95E+3 - 1.90E+3 1.26E+3 1-133 1.32E+4 1.92E+4 3.56E+6 2.24E+4 - 2.16E+3 5.60E+3 1-134 9.21 E+2 1.88E+3 4.45E+4 2.09E+3 - 1.29E+3 6.65E+2 1-135 3.86E+3 7.60E+3 6.96E+5 8.47E+3 - 1.83E+3 2.77E+3 Cs-134 3.96E+5 7.03E+5 - 1.90E+5 7.97E+4 1.33E+3 7.45E+4 Cs-136 4.83E+4 1.35E+5 - 5.64E+4 1.1 8E+4 1.43E+3 5.29E+4 Cs-137 5.49E+5 6.12E+5 - 1.72E+5 7.13E+4 1.33E+3 4.55E+4 Cs-138 5.05E+2 7.81 E+2 - 4.1OE+2 6.54E+1 8.76E+2 3.98E+2 Ba-139 1.48E+0 9.84E-4 - 5.92E-4 5.95E+3 5.1OE+4 4.30E-2 Ba-140 5.60E+4 5.60E+1 - 1.34E+1 1.60E+6 3.84E+4 2.90E+3 Ba-141 1.57E-1 1.08E-4 - 6.50E-5 2.97E+3 4.75E+3 4.97E-3 Ba-142 3.98E-2 3.30E-5 - 1.90E-5 1.55E+3 6.93E+2 1.96E-3 La-140 5.05E+2 2.OOE+2 - - 1.68E+5 8.48E+4 5.15E+1 La-142 1.03E+0 3.77E-1 - - 8.22E+3 5.95E+4 9.04E-2 Ce-141 2.77E+4 1.67E+4 - 5.25E+3 5.17E+5 2.16E+4 1.99E+3 Ce-1 43 2.93E+2 1.93E+2 - 5.64E+1 1.16E+5 4.97E+4 2.21 E+1 Ce-144 3.19E+6 1.21 E+6 - 5.38E+5 9.84E+6 1.48E+5 1.76E+5 Pr-143 1.40E+4 5.24E+3 - 1.97E+3 4.33E+5 3.72E+4 6.99E+2 Pr-144 4.79E-2 1.85E-2 - 6.72E-3 1.61 E+3 4.28E+3 2.41 E-3 Nd-147 7.94E+3 8.13E+3 - 3.15E+3 3.22E+5 3.12E+4 5.OOE+2 W-187 1.30E+1 9.02E+0 - - 3.96E+4 3.56E+4 3.12E+O Np-239 3.71 E+2 3.32E+1 - 6.62E+1 5.95E+4 2.49E+4 1.88E+1 2-24 REV. 9 12/02/2005

Table 2.8 RI Vegetation Pathway Dose Factors - ADULT (mremlyr per +/-CI/m3) for H-3 and C-14 (M2 x mrem/yr gCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 C-14 8.97E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 Na-24 2.76E+5 2.76E+5 2.76E+5 2.76E+5 2.76E+5 2.76E+5 2.76E+5 P-32 1.40E+9 8.73E+7 - - - 1.58E+8 5.42E+7 Cr-51 - 2.79E+4 1.03E+4 6.19E+4 1.17E+7 4.66E+4 Mn-54 . 3.11 E+8 - 9.27E+7 - 9.54E+8 5.94E+7 Mn-56 1.61 E+1 - 2.04E+1 - 5.13E+2 2.85E+0 Fe-55 2.09E+8 1.45E+8 - - 8.06E+7 8.29E+7 3.37E+7 Fe-59 1.27E+8 2.99E+8 - - 8.35E+7 9.96E+8 1.14E+8 Co-57 - 1.17E+7 - - - 2.97E+8 1.95E+7 Co-58 - 3.09E+7 - - 6.26E+8 6.92E+7 Co-60 1.67E+8 - - 3.14E+9 3.69E+8 Ni-63 1.04E+10 7.21 E+8 - - - 1.50E+8 3.49E+8 Ni-65 6.15E+1 7.99E+0 - - - 2.03E+2 3.65E+0 Cu-64 - 9.27E+3 - 2.34E+4 - 7.90E+5 4.35E+3 Zn-65 3.17E+8 1.01 E+9 - 6.75E+8 - 6.36E+8 4.56E+8 Zn-69 8.75E-6 1.67E-5 - 1.09E-5 - 2.51 E-6 1.16E-6 Br-82 - - - - 1.73E+6 1.51 E+6 Br-83 - - - - - 4.63E+O 3.21 E+0 Br-84 -

Br-85 Rb-86 - 2.19E+8 - - - 4.32E+7 1.02E+8 Rb-88 - - - - - - -

Rb-89 Sr-89 9.96E+9 - - - - 1.60E+9 2.86E+8 Sr-90 6.05E+11 - - - - 1.75E+10 1.48E+1 1 Sr-91 3.20E+5 - - - - 1.52E+6 1.29E+4 Sr-92 4.27E+2 - - - - 8.46E+3 1.85E+1 Y-90 1.33E+4 - - - - 1.41 E+8 3.56E+2 Y-91 m 5.83E-9 - - - - 1.71 E-8 Y-91 5.13E+6 - - - - 2.82E+9 1.37E+5 Y-92 9.01 E-1 - - - - 1.58E+4 2.63E-2 Y-93 1.74E+2 - - - - 5.52E+6 4.80E+0 Zr-95 1.19E+6 3.81E+5 - 5.97E+5 - 1.21 E+9 2.58E+5 Zr-97 3.33E+2 6.73E+1 - 1.02E+2 - 2.08E+7 3.08E+1 Nb-95 1.42E+5 7.91 E+4 - 7.81 E+4 - 4.80E+8 4.25E+4 Nb-97 2.90E-6 7.34E-7 - 8.56E-7 - 2.71 E-3 2.68E-7 Mo-99 - 6.25E+6 - 1.41E+7 - 1.45E+7 1.19E+6 Tc-99m 3.06E+0 8.66E+0 - 1.32E+2 4.24E+0 5.12E+3 1.10E+2 Tc-101 - - - - - - -

2-25 REV. 9 12/02/2005

Table 2.8 RI Vegetation Pathway Dose Factors - ADULT (mremlyr per gCfm 3

) for H-3 and C-14 (m2 x mremlyr iCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-103 4.80E+6 - - 1.83E+7 - 5.61E+8

• 2.07E+6 Ru-105 5.39E+1 - - 6.96E+2 - 3.30E+4 2.13E+1 Ru-106 1.93E+8 - - 3.72E+8 - 1.25E+10 2.44E+7 Rh-103m Rh-106 - -

Ag-1I Om 1.06E+7 .9.76E+6 - 1.92E+7 - 3.98E+9 5.80E+6 Sb-124 1.04E+8 1.96E+6 2.52E+5 - 8.08E+7 2.95E+9 4.11 E+7 Sb-125 1.36E+8 1.52E+6 1.39E+5 - 1.05E+8 1.50E+9 3.25E+7 Te-125m 9.66E+7 3.50E+7 2.90E+7 3.93E+3 - 3.86E+8 1.29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1.42E+9 - 1.17E+9 4.26E+7 Te-127 5.76E+3 2.07E+3 4.27E+3 2.35E+4 - 4.54E+5 1.25E+3 Te-129m 2.55E+8 9.50E+7 8.75E+7 1.06E+9 - 1.28E+9 4.03E+7 Te-129 6.65E-4 2.50E-4 5.1OE-4 2.79E-3 - 5.02E-4 1.62E-4 Te-131m 9.12E+5 4.46Ei5 7.06E+5 4.52E+6 - 4.43E+7 3.72E+5 Te-131 - - - - - -

Te-1 32 4.29E+6 2.77E+6 3.06E+6 2.67E+7 - 1.31 E+8 2.60E+6 1-130 3.96E+5 1.17E+6 9.90E+7 1.82E+6 - 1.01 E+6 4.61 E+5 1-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 - 3.05E+7 6.63E+7 1-132 5.74E+1 1.54E+2 5.38E+3 2.45E+2 - 2.89E1+1 5.38E+1 1-133 2.12E+6 3.69E+6 5.42E+8 6.44E+6 - 3.31 E+6 1.12E+6 1-134 1.06E-4 2.88E-4 5.OOE-3 4.59E-4 - 2.51 E-7 1.03E-4 1-135 4.08E+4 1.07E+5 7.04E+6 1.71E+5 - 1.21E+5 3.94E+4 Cs-1 34 4.66E+9 1.11 E+10 - 3.59E+9 1.19E+9 1.94E+8 9.07E+9 Cs-1 36 4.20E+7 1.66E+8 - 9.24E+7 1.27E+7 1.89E+7 1.19E+8 Cs-1 37 6.36E+9 8.70E+9 - 2.95E+9 9.81 E+8 1.68E+8 5.70E+9 Cs-1 38 - - - - -

Ba-1 39 2.95E-2 2.1 OE-5 - 1.96E-5 1.19E-5 5.23E-2 8.64E-4 Ba-i 40 1.29E+8 1.62E+5 - 5.49E+4 9.25E+4 2.65E+8 8.43E+6 Ba-141 Ba-142 La-1 40 1.97E+3 9.92E+2 - - - 7.28E+7 2.62E+2 La-1 42 1.40E-4 6.35E-5 - - - 4.64E-1 1.58E-5 Ce-141 1.96E+5 1.33E+5 - 6.17E+4 - 5.08E+8 1.51 E+4 Ce- 43 1.OOE+3 7.42E+5 3.26E+2 - 2.77E+7 8.21 E+1 Ce- 44 3.29E+7 1.38E+7 8.16E+6 - 1.11E+10 1.77E+6 Pr-143 6.34E+4 2.54E+4 - 1.47E+4 - 2.78E+8 3.14E+3 Pr-14 - - -4 Nd-147 3.34E+4 3.86E+4 - 2.25E+4 - 1.85E+8 2.31 E+3 W-187 3.82E+4 3.19E+4 - - - 1.05E+7 1.12E+4 Np-239 1.42E+3 1.40E+2 - 4.37E+2 - 2.87E+7 7.72E+1 2-26 REV. 9 12102/2005

Table 2.9 RI Vegetation Pathway Dose Factors - TEEN (mremlyr per +/-Cims) for H-3 and C-14 (M2 x mrem/yr gilCsec) for others Nucilde Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2:59E+3 C-14 1.45E+6 2.91 E+5 2.91 E+5 2.91 E+5 2.91 E+5 2.91 E+5 2.91 E+5 Na-24 2.45E+5 2.45E+5 2.45E+5 2.45E+5 2.45E+5 2.45E+5 2.45E+5 P-32 1.61 E+9 9.96E+7 - - 1.35E+8 6.23E+7 Cr-51 - - 3.44E+4 1.36E+4 8.85E+4 1.04E+7 6.20E+4 Mn-54 4.52E+8 1.35E+8 - 9.27E+8 8.97E+7 Mn-56 - 1.45E+1 - 1.83E+1 - 9.54E+2 2.58E+O Fe-55 3.25E+8 2.31 E+8 - - 1.46E+8 9.98E+7 5.38E+7 Fe-59 1.81 E+8 4.22E+8 - - 1.33E+8 9.98E+8 1.63E+8 Co-57 1.79E+7 - - 3.34E+8 3.OOE+7 Co-58 - 4.38E+7 - - - 6.04E+8 1.01 E+8 Co-60 - 2.49E+8 - - - 3.24E+9 5.60E+8 Ni-63 1.61E+10 1.13E+9 - - - 1.81E+8 5.45E+8 Ni-65 5.73E+1 7.32E+O - - 3.97E+2 3.33E+0 Cu-64 - 8.40E+3 - 2.12E+4 - 6.51 E+5 3.95E+3 Zn-65 4.24E+8 1.47E+9 - 9.41 E+8 - 6.23E+8 6.86E+8 Zn-69 8.19E-6 1.56E-5 - 1.02E-5 - 2.88E-5 1.09E-6 Br-82 - - - - - - 1.33E+6 Br-83 - - - - - - 3.01 E+0 Br-84 -

Br-85 Rb-86 - 2.73E+8 - - - 4.05E+7 1.28E+8 Rb-88 - -

Rb-89 -

Sr-89 1.51 E+1 0 - - - - 1.80E+9 4.33E+8 Sr-90 7.51 E+11 - - - - 2.11E+10 1.85E+11 Sr-91 2.99E+5 - - - - 1.36E+6 1.19E+4 Sr-92 3.97E+2 - - - - 1.01 E+4 1.69E+1 Y-90 1.24E+4 - - - 1.02E+8 3.34E+2 Y-91 m 5.43E-9 - - - 2.56E-7 Y-91 7.87E+6 - - - - 3.23E+9 2.11 E+5 Y-92 8.47E-1 - - - 2.32E+4 2.45E-2 Y-93 1.63E+2 - - - - 4.98E+6 4.47E+0 Zr-95 1.74E+6 5.49E+5 - 8.07E+5 - 1.27E+9 3.78E+5 Zr-97 3.09E+2 6.11 E+1 - 9.26E+1 - 1.65E+7 2.81 E+1 Nb-95 1.92E+5 1.06E+5 - 1.03E+5 - 4.55E+8 5.86E+4 Nb-97 2.69E-6 6.67E-7 - 7.80E-7 - 1.59E-2 2.44E-7 Mo-99 - 5.74E+6 - 1.31 E+7 - 1.03E+7 1.09E+6 Tc-99m 2.70E+0 7.54E+O - 1.12E+2 4.19E+0 4.95E+3 9.77E+1 Tc-101 2-27 REV. 9 12/02/2005

Table 2.9 RI Vegetation Pathway Dose Factors - TEEN (mrem/yr per gCim 3 ) for H-3 and C-14 (M2 x mremlyr pCIsec) for others Nucilde Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-1 03 6.87E+6 - - 2.42E+7 - 5.74E+8 2.94E+6 Ru-1 05 5.OOE+1 - - 6.31 E+2 - 4.04E+4 1.94E+1 Ru-106 3.09E+8 - - 5.97E+8 - 1.48E+10 3.90E+7 Rh-103m - - - - - -

Rh-100- - - - - -

Ag-11 Om 1.52E+7 1.44E+7 - 2.74E+7 - 4.04E+9 8.74E+6 Sb-124 1.55E+8 2.85E+6 3.51 E+5 - 1.35E+8 3.11 E+9 6.03E+7 Sb-125 2.14E+8 2.34E+6 2.04E+5 - 1.88E+8 1.66E+9 5.OOE+7 Te-125m 1.48E+8 5.34E+7 4.14E+7 - - 4.37E+8 1.98E+7 To-127m 5.51 E+8 1.96E+8 1.31 E+8 2.24E+9 - 1.37E+9 6.56E+7 Te-127 5.43E+3 1.92E+3 3.74E+3 2.20E+4 - 4.19E+5 1.17E+3 Te-129m 3.67E+8 1.36E+8 1.18E+8 1.54E+9 - 1.38E+9 5.81 E+7 Te-129 6.22E-4 2.32E-4 4.45E-4 2.61 E-3 - 3.40E-3 1.51 E-4 Te-1 31 m 8.44E+5 4.05E+5 6.09E+5 4.22E+6 - 3.25E+7 3.38E+5 Te-131 - - . . . .

Te-1 32 3.90E+6 2.47E+6 2.60E+6 2.37E+7 - 7.82E+7 2.32E+6 1-130 3.54E+5 1.02E+6 8.35E+7 1.58E+6 - 7.87E+5 4.09E+5 1-131 7.70E+7 1.08E+8 3.14E+10 1.85E+8 - 2.13E+7 5.79E+7 1-132 5.18E+1 1.36E+2 4.57E+3 2.14E+2 - 5.91E+1 4.87E+1 1-133 1.97E+6 3.34E+6 4.66E+8 5.86E+6 - 2.53E+6 1.02E+6 1-134 9.59E-5 2.54E-4 4.24E-3 4.01 E-4 - 3.35E-6 9.13E-5 1-135 3.68E+4 9.48E+4 6.1 OE+6 1.50E+5 - 1.05E+5 3.52E+4 Cs-1 34 7.09E+9 1.67E+1 0 - 5.30E+9 2.02E+9 2.08E+B 7.74E+9 Cs-136 4.29E+7 1.69E+8 - 9.19E+7 1.45E+7 1.36E+7 1.13E+8 Cs-137 1.01 E+10 1.35E+10 - 4.59E+9 1.78E+9 1.92E+8 4.69E+9 Cs-1 38 - - - - - -

Ba-1 39 2.77E-2 1.95E-5 - 1.84E-5 1.34E-5 2.47E-1 8.08E-4 Ba-140 1.38E+8 1.69E+5 - 5.75E+4 1.14E+5 2.13E+8 8.91 E+6 Ba-141 - - . -

Ba-142 - - - - -

La-1 40 1.80E+3 8.84E+2 - - - 5.08E+7 2.35E+2 La-1 42 1.28E-4 5.69E-5 - - - 1.73E+0 1.42E-5 Ce-1 41 2.82E+5 1.88E+5 - 8.86E+4 - 5.38E+8 2.16E+4 Ceo1 43 9.37E+2 6.82E+5 - 3.06E+2 - 2.05E+7 7.62E+1 Ce-144 5.27E+7 2.18E+7 - 1.30E+7 - 1.33E+10 2.83E+6 Pr-1 43 7.12E+4 2.84E+4 - 1.65E+4 - 2.34E+8 3.55E+3 Pr-1444 - - -

Nd-147 3.63E+4 3.94E+4 - 2.32E+4 - 1.42E+8 2.36E+3 W-1 87 3.55E+4 2.90E+4 - - - 7.84E+6 1.02E+4 Np-239 1.38E+3 1.30E+2 - 4.09E+2 - 2.1 OE+7 7.24E+1 2-28 REV. 9 12/02/2005

Table 2.10 RI Vegetation Pathway Dose Factors - CHILD (mremlyr per uCilm 3) for H-3 and C-1 4 (M2 x mreinlyr pLCilsec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 4.01 E+3 4.01 E+3 4.01 E+3 4.01 E+3 4.01 E+3 4.01 E+3 C-14 3.50E+6 7.01 E+5 7.01 E+5 7.01 E+5 7.01 E+5 7.01 E+5 7.01 E+5 Na-24 3.83E+5 3.83E+5 3.83E+5 3.83E+5 3.83E+5 3.83E+5 3.83E+5 P-32 3.37E+9 1.58E+8 - - - 9.30E+7 1.30E+8 Cr-51 - - 6.54E+4 1.79E+4 1.19E+5 6.25E+6 1.18E+5 Mn-54 - 6.61 E+8 - 1.85E+8 - 5.55E+8 1.76E+8 Mn-56 - 1.90E+1 - 2.29E+1 - 2.75E+3 4.28E+0 Fe-55 8.OOE+8 4.24E+8 - - 2.40E+8 7.86E+7 1.31 E+8 Fe-59 4.01 E+8 6.49E+8 - - 1.88E+8 6.76E+8 3.23E+8 Co-57 - 2.99E+7 - - - 2.45E+8 6.04E+7 Co-58 - 6.47E+7 - - - 3.77E+8 1.98E+8 Co-60 - 3.78E+8 - - - 2.10E+9 1.12E+9 Ni-63 3.95E+10 2.11E+9 - - - 1.42E+8 1.34E+9 Ni-65 1.05E+2 9.89E+O - - - 1.21 E+3 5.77E+O Cu-64 - 1.11 E+4 - 2.68E+4 - 5.20E+5 6.69E+3 Zn-65 8.12E+8 2.16E+9 - 1.36E+9 - 3.80E+8 1.35E+9 Zn-69 1.51 E-5 2.18E-5 - 1.32E-5 - 1.38E-3 2.02E-6 Br-82 - - - - - - 2.04E+6 Br-83 - - - - - - 5.55E+O Br-84 Br-85 -

Rb-86 - 4.52E+8 - - - 2.91 E+7 2.78E+8 Rb-88 - - - - - - -

Rb-89 -

Sr-89 3.59E+10 - - - 1.39E+9 1.03E+9 Sr-GO 1.24E+12 - - - 1.67E+10 3.15E+11 Sr-91 5.50E+5 - - - - 1.21 E+6 2.08E+4 Sr-92 7.28E+2 - - - - 1.38E+4 2.92E+1 Y-90 2.30E+4 - - - 6.56E+7 6.17E+2 Y-91m 9.94E-9 - - - - 1.95E-5 -

Y-91 1.87E+7 - - - - 2.49E+9 5.01 E+5 Y-92 1.56E+0 - - - - 4.51 E+4 4.46E-2 Y-93 3.01 E+2 - - - 4.48E+6 8.25E+0 Zr-95 3.90E+6 8.58E+5 - 1.23E+6 - 8.95E+8 7.64E+5 Zr-97 5.64E+2 8.15E+1 - 1.17E+2 - 1.23E+7 4.81E+1 Nb-95 4.10E+5 1.59E+5 - 1.50E+5 - 2.95E+8 1.14E+5 Nb-97 4.90E-6 8.85E-7 - 9.82E-7 - 2.73E-1 4.13E-7 Mo-99 - 7.83E+6 - 1.67E+7 - 6.48E+6 1.94E+6 Tc-99m 4.65E+0 9.12E+0 - 1.33E+2 4.63E+O 5.19E+3 1.51 E+2 Tc-101 - - - - - -

2-29 REV. 9 12/02/2005

Table 2.10 RI Vegetation Pathway Dose Factors - CHILD (mrernlyr per uClm 3) for H-3 and C-14 (M2 x mremlyr XCI/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-1 03 1.55E+7 - - 3.89E+7 - 3.99E+8 5.94E+6 Ru- 05 9.17E+1 - - 8.06E+2 - 5.98E+4 3.33E+1 Ru-106 7.45E+8 - - 1.01 E+9 - 1.16E+10 9.30E+7 Rh-103m -

Rh-106 - - - - -

Ag-11 Om 3.22E+7 2.17E+7 - 4.05E+7 2.58E+9 1.74E+7 Sb-1 24 3.52E+8 4.57E+6 7.78E+5 1.96E+8 2.20E+9 1.23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 - 2.78E+8 1.19E+9 1.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 - - 3.38E+8 4.67E+7 Te-127m 1.32E+9 3.56E+8 3.16E+8 3.77E+9 - 1.07E+9 1.57E+8 Te-1 27 1.OOE+4 2.70E+3 6.93E+3 2.85E+4 - 3.91 E+5 2.1 5E+3 Te-129m 8.54E+8 2.39E+8 2.75E+8 2.51 E+9 - 1.04E+9 1.33E+8 Te-129 1.15E-3 3.22E-4 8.22E-4 3.37E-3 - 7.17E-2 2.74E-4 Te-131m 1.54E+6 5.33E+5 1.10E+6 5.16E+6 - 2.16E+7 5.68E+5 Te-131 -

Te-1 32 6.98E+6 3.09E+6 4.50E+6 2.87E+7 - 3.11 E+7 3.73E+6 1-130 6.21 E+5 1.26E+6 1.38E+8 1.88E+6 - 5.87E+5 6.47E+5 1-131 1.43E+8 1.44E+B 4.76E+1 0 2.36E+8 - 1.28E+7 8.1 8E+7 1-132 9.20E+1 1.69E+2 7.84E+3 2.59E+2 - 1.99E+2 7.77E+1 1-133 3.59E+6 4.44E+6 8.25E+8 7.40E+6 - 1.79E+6 1.68E+6 1-134 1.70E-4 3.16E-4 7.28E-3 4.84E-4 - 2.10E-4 1.46E-4 1-135 6.54E+4 1.18E+5 1.04E+7 1.81 E+5 - 8.98E+4 5.57E+4 Cs-1 34 1.60E+10 2.63E+10 - 8.14E+9 2.92E+9 1.42E+8 5.54E+9 Cs-1 36 8.06E+7 2.22E+8 - 1.18E+8 1.76E+7 7.79E+6 1.43E+8 Cs-137 2.39E+10 2.29E+10 - 7.46E+9 2.68E+9 1.43E+8 3.38E+9 Cs-138 - - - - - -

Ba-139 5.11 E-2 2.73E-5 - 2.38E-5 1.61 E-5 2.95E+0 1.48E-3 Ba-140 2.77E+8 2.43E+5 - 7.90E+4 1.45E+5 1.40E+8 1.62E+7 Ba-141 Ba-142 - - - -

La-140 3.23E+3 1.13E+3 - - - 3.15E+7 3.81E+2 La-142 2.32E-4 7.40E-5 - - - 1.47E+1 2.32E-5 Ce-141 6.35E+5 3.26E+5 - 1.43E+5 - 4.07E+8 4.84E+4 Ce-143 1.73E+3 9.36E+5 - 3.93E+2 - 1.37E+7 1.36E+2 Ce-144 1.27E+8 3.98E+7 - 2.21 E+7 - 1.04E+10 6.78E+6 Pr-143 1.48E+5 4.46E+4 - 2.41 E+4 - 1.60E+8 7.37E+3 Pr-144 - - - -

Nd-147 7.16E+4 5.802+4 - 3.18E+4 - 9.18E+7 4.49E+3 W-187 6.47E+4 3.83E+4 - - - 5.38E+6 1.72E+4 Np-239 2.55E+3 1.83E+2 - 5.30E+2 - 1.36E+7 1.29E+2 2-30 REV. 9 12/02/2005

Table 2.11 RI Grass-Cow-Milk Pathway Dose Factors - ADULT (mrem/yr per )iClm 3 ) for H-3 and C-14 (M2 x mremlyr 1Ci/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 C-14 3.63E+5 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 Na-24 2.54E+6 2.54E+6 2.54E+6 2.54E+6 2.54E+6 2.54E+6 2.54E+6 P-32 1.71 E+10 1.06E+9 - - 1.92E+9 6.60E+8 Cr-51 - - 1.71 E+4 6.30E+3 3.80E+4 7.20E+6 2.86E+4 Mn-54 - 8.40E+6 - 2.50E+6 2.57E+7 1.60E+6 Mn-56 - 4.23E-3 - 5.38E-3 - 1.35E-1 7.51 E-4 Fe-55 2.51 E+7 1.73E+7 - - 9.67E+6 9.95E+6 4.04E+6 Fe-59 2.98E+7 7.OOE+7 - - 1.95E+7 2.33E+8 2.68E+7 Co-57 - 1.28E+6 - - - 3.25E+7 2.13E+6 Co-58 - 4.72E+6 - - - 9.57E+7 1.06E+7 Co-60 1.64E+7 - - - 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 - - 9.73E+7 2.26E+8 Ni-65 3.70E-1 4.81E-2 - - - 1.22E+O 2.19E-2 Cu-64 - 2.41 E+4 - 6.08E+4 - 2.05E+6 1.13E+4 Zn-65 1.37E+9 4.36E+9 - 2.92E+9 - 2.75E+9 1.97E+9 Zn-69 - - - - - - -

Br-82 - - - - 3.72E+7 3.25E+7 Br-83 - - - - 1.49E-1 1.03E-1 Br-84 - - -

Br-85 Rb-86 - 2.59E+9 - - - 5.11 E+8 1.21 E+9 Rb-88 -

Rb-89 Sr-89 1.45E+9 - - - - 2.33E+8 4.16E+7 Sr-90 4.68E+10 - - - - 1.35E+9 1.15E+10 Sr-91 3.13E+4 - - - - 1.49E+5 1.27E+3 Sr-92 4.89E-1 - - - - 9.68E+O 2.11 E-2 Y-90 7.07E+1 - - - - 7.50E+5 1.90E+O Y-91m Y-91 8.60E+3 - - - - 4.73E+6 2.30E+2 Y-92 5.42E-5 - - - - 9.49E-1 1.58E-6 Y-93 2.33E-1 - - - - 7.39E+3 6.43E-3 Zr-95 9.46E+2 3.03E+2 - 4.76E+2 - 9.62E+5 2.05E+2 Zr-97 4.26E-1 8.59E-2 - 1.30E-1 - 2.66E+4 3.93E-2 Nb-95 8.25E+4 4.59E+4 - 4.54E+4 - 2.79E+8 2.47E+4 Nb-97 - - - - - 5.47E-9 Mo-99 - 2.52E+7 - 5.72E+7 - 5.85E+7 4.80E+6 Tc-99m 3.25E+O 9.19E+O - 1.40E+2 4.50E+O 5.44E+3 1.17E+2 Tc-101 - - - -

2-31 REV. 9 12/02/2005

Table 2.11 RI Grass-Cow-Milk Pathway Dose Factors - ADULT (mrem/yr per ILCVm 3 ) for H-3 and C-14 (M2 x mremlyr gClsec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LU T.Body Ru-103 1.02E+3 - - 3.89E+3 - 1.19E+5 4.39E+2 Ru-105 8.57E-4 - - 1.11 E-2 - 5.24E-1 3.38E-4 Ru-106 2.04E+4 - - 3.94E+4 - 1.32E+6 2.58E+3 Rh-103m Rh-106 - - - -

Ag-11 Om 5.83E+7 5.39E+7 - 1.06E+8 - 2.20E+1 0 3.20E+7 Sb-124 2.57E+7 4.86E+5 6.24E+4 - 2.OOE+7 7.31 E+8 1.02E+7 Sb-125 2.04E+7 2.28E+5 2.08E+4 1.58E+7 2.25E+8 4.86E+6 Te-125m 1.63E+7 5.90E+6 4.90E+6 6.63E+7 - 6.50E+7 2.18E+6 Te-127m 4.58E+7 1.64E+7 1.17E+7 1.86E+8 - 1.54E+8 5.58E+6 Te-127 6.72E+2 2.41 E+2 4.98E+2 2.74E+3 - 5.30E+4 1.45E+2 Te-129m 6.04E+7 2.25E+7 2.08E+7 2.52E+8 - 3.04E+8 9.57E+6 Te-129 - - - - - - -

Te-131m 3.61 E+5 1.77E+5 2.80E+5 1.79E+6 - 1.75E+7 1.47E+5 Te-131 -

Te-1 32 2.39E+6 1.55E+6 1.71 E+6 1.49E+7 - 7.32E+7 1.45E+6 1-130 4.26E+5 1.26E+6 1.07E+8 1.96E+6 - 1.08E+6 4.96E+5 1-131 2.96E+8 4.24E+8 1.39E+11 7.27E+8 - 1.12E+8 2.43E+8 1-132 1.64E-1 4.37E-1 1.53E+1 6.97E-1 - 8.22E-2 1.53E-1 1-133 3.97E+6 6.90E+6 1.01 E+9 1.20E+7 - 6.20E+6 2.10E+6 1-134 . . . . -

1-135 1.39E+4 3.63E+4 2.40E+6 5.83E+4 - 4.1OE+4 1.34E+4 Cs-134 5.65E+9 1.34E+10 - 4.35E+9 1.44E+9 2.35E+8 1.10E+10 Cs-1 36 2.61 E+8 1.03E+9 - 5.74E+8 7.87E+7 1.17E+8 7.42E+8 Cs-137 7.38E+9 1.01 E+10 - 3.43E+9 1.14E+9 1.95E+8 6.61 E+9 Cs-138 - - - - .

Ba-1 39 4.70E-8 - - - 8.34E-8 1.38E-9 Ba-140 2.69E+7 3.38E+4 1.15E+4 1.93E+4 5.54E+7 1.76E+6 Ba-141 Ba-142 - - - - - - -

La-140 4.49E+0 2.26E+0 - - - 1.66E+5 5.97E-1 La-142 - - - - 3.03E-8 Ce-141 4.84E+3 3.27E+3 - 1.52E+3 - 1.25E+7 3.71 E+2 Ca-143 4.19E+1 3.09E+4 - 1.36E+1 - 1.16E+6 3.42E+O Ce-144 3.58E+5 1.50E+5 - 8.87E+4 - 1.21 E+8 1.92E+4 Pr-143 1.59E+2 6.37E+1 - 3.68E+1 - 6.96E+5 7.88E+O Pr-144 Nd-147 9.42E+1 1.09E+2 - 6.37E+1 - 5.23E+5 6.52E+O W-187 6.56E+3 5.48E+3 - - - 1.80E+6 1.92E+3 Np-239 3.66E+0 3.60E-1 - 1.12E+0 - 7.39E+4 1.98E-1 2-32 REV. 9 12/02/2005

Table 2.12 RI Grass-Cow-Milk Pathway Dose Factors - TEEN (mremlyr per ,iCLm 3) for H-3 and C-14 (M2 x mrem/yr pICI/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 C-14 6.70E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 Na-24 4.44E+6 4.44E+6 4.44E+6 4.44E+6 4.44E+6 4.44E+6 4.44E+6 P-32 3.15E+1 0 1.95E+9 - 2.65E+9 1.22E+9 Cr-51 - - 2.78E+4 1.10E+4 7.13E+4 8.40E+6 5.OOE+4 Mn-54 - 1.40E+7 4.17E+6 - 2.87E+7 2.78E+6 Mn-56 - 7.51 E-3 - 9.50E-3 - 4.94E-1 1.33E-3 Fe-55 4.45E+7 3.16E+7 - - 2.OOE+7 1.37E+7 7.36E+6 Fe-59 5.20E+7 1.21 E+8 - 3.82E+7 2.87E+8 4.68E+7 Co-57 - 2.25E+6 - - 4.19E+7 3.76E+6 Co-58 - 7.95E+6 - 1.1OE+8 1.83E+7 Co-60 - 2.78E+7 - - 3.62E+8 6.26E+7 Ni-63 1.18E+10 8.35E+8 - - - 1.33E+8 4.01 E+8 Ni-65 6.78E-1 8.66E-2 - - 4.70E+0 3.94E-2 Cu-64 - 4.29E+4 - 1.09E+5 - 3.33E+6 2.02E+4 Zn-65 2.11 E+9 7.31 E+9 4.68E+9 - 3.1 OE+9 3.41 E+9 Zn-69 Br-82 - - - - - 5.64E+7 Br-83 - - - - - - 1.91 E-1 Br-84 -

Br-85 -

Rb-86 - 4.73E+9 - - - 7.OOE+8 2.22E+9 Rb-88 - - - - 2 38E R-189 - - 1 Sr-89 2.67E+9 - - - - 3.18E+8 7.66E+7 Sr-90 6.61E+10 - - - - 1.86E+9 1.63E+10 Sr-91 5.75E+4 - - - 2.61 E+5 2.29E+3 Sr-92 8.95E-1 - - - - 2.28E+1 3.81E-2 Y-90 1.30E+2 - - - - 1.07E+6 3.50E+O Y-91 m Y-91 1.58E+4 -8 - - - 6.4E+6 4.24E+2 Y-92 1.00E-4 ---- 2.75E+0 2.90E-6 Y-93 4.30E-1 ---- 1.31 E+4 1.18E-2 Zr-95 1.65E+3 5.22E+2 - 7.67E+2 - 1.20E+6 3.59E+2 Zr-97 7.75E-1 1.53E-1 - 2.32E-1 - 4.15E+4 7.06E-2 Nb-95 1.41 E+5 7.80E+4 - 7.57E+4 - 3.34E+8 4.30E+4 Nb-97 - - - - - 6.34E-8 Mo-99 - 4.56E+7 - 1.04E+8 - 8.16E+7 8.69E+6 Tc-99m 5.64E+0 1.57E+1 - 2.34E+2 8.73E+O 1.03E+4 2.04E+2 Tc-101 - - - - - -

2-33 REV. 9 12/02/2005

Table 2.12 RI Grass-Cow-Milk Pathway Dose Factors - TEEN (mremlyr per ljCVm 3 ) for H-3 and C-14 (M2 x mremlyr pCI/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-1 03 1.81 E+3 - - 6.40E+3 - 1.52E+5 7.75E+2 Ru-1 05 1.57E-3 - - 1.97E-2 - 1.26E+0 6.08E-4 Ru-106 3.75E+4 - - 7.23E+4 - 1.80E+B 4.73E+3 Rh-103m Rh-106 - -

Ag-i 1 Om 9.63E+7 9.11 E+7 - 1.74E+8 - 2.56E+10 5.54E+7 Sb-124 4.59E+7 8.46E+5 1.04E+5 - 4.01 E+7 9.25E+8 1.79E+7 Sb-125 3.65E+7 3.99E+5 3.49E+4 - 3.21 E+7 2.84E+8 8.54E+6 To-125m 3.OOE+7 1.08E+7 8.39E+6 - - 8.86E+7 4.02E+6 Te-127m 8.44E+7 2.99E+7 2.01 E+7 3.42E+8 - 2.1OE+8 1.OOE+7 Te-127 1.24E+3 4.41 E+2 8.59E+2 5.04E+3 - 9.61 E+4 2.68E+2 Ts-129m 1.11E+8 4.10E+7 3.57E+7 4.62E+8 - 4.15E+8 1.75E+7 Te-129 - - - 1.67E-9 - 2.18E-9 Ts-131m 6.57E+5 3.15E+5 4.74E+5 3.29E+6 - 2.53E+7 2.63E+5 Te-13 - - - - - -3 To-1 32 4.28E+6 2.71 E+6 2.86E+6 2.60E+7 - 8.58E+7 2.55E+6 1-130 7.49E+5 2.17E+6 1.77E+8 3.34E+6 - 1.67E+6 8.66E+5 1-131 5.38E+8 7.53E+8 2.20E+11 1.30E+9 - 1.49E+8 4.04E+8 1-132 2.90E-1 7.59E-1 2.56E+1 1.20E+0 - 3.31 E-1 2.72E-1 1-133 7.24E+6 1.23E+7 1.72E+9 2.15E+7 - 9.30E+6 3.75E+6 1-134 - - - - - -

1-135 2.47E+4 6.35E+4 4.08E+6 1.OOE+5 7.03E+4 2.35E+4 Cs-1 34 9.81 E+9 2.31 E+1 0 - 7.34E+9 2.80E+9 2.87E+8 1.07E+1 0 Cs-136 4.45E+8 1.75E+9 - 9.53E+8 1.50E+8 1.41 E+8 1.18E+9 Cs-137 1.34E+10 1.78E+10 - 6.06E+9 2.35E+9 2.53E+B 6.20E+9 Cs-1 38 Ba-139 8.69E-8 - - - - 7.75E-7 2.53E-9 Ba-1 40 4.85E+7 5.95E+4 - 2.02E+4 4.OOE+4 7.49E+7 3.13E+6 Ba-141 Ba-142 La-140 8.06E+0 3.96E+O - - - 2.27E+5 1.05E+0 La-142 - - - - - 2.23E-7 -

Co-141 8.87E+3 5.92E+3 - 2.79E+3 - 1.69E+7 6.81 E+2 Ce-143 7.69E+1 5.60E+4 - 2.51 E+1 - 1.68E+6 6.25E+0 Ce-144 6.58E+5 2.72E+5 - 1.63E+5 - 1.66E+8 3.54E+4 Pr-143 2.92E+2 1.17E+2 - 6.77E+1 - 9.61E+5 1.45E+1 Pr-144 Nd-147 1.81E+2 1.97E+2 - 1.16E+2 - 7.11E+5 1.18E+1 W-1 87 1.20E+4 9.78E+3 - - - 2.65E+6 3.43E+3 Np-239 6.99E+0 6.59E-1 - 2.07E+0 - 1.06E+5 3.66E-1 2-34 REV. 9 12/02/2005

Table 2.13 RI Grass-Cow-Milk Pathway Dose Factors - CHILD (mremlyr per gCVm 3 ) for H-3 and C-14 (M2 x mrem/yr IIC/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 C-14 1.65E+6 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 Na-24 9.23E+6 9.23E+6 9.23E+6 9.23E+6 9.23E+6 9.23E+6 9.23E+6 P-32 7.77E+10 3.64E+9 - - - 2.15E+9 3.OOE+9 Cr-51 - - 5.66E+4 1.55E+4 1.03E+5 5.41 E+6 1.02E+5 Mn-54 - 2.09E+7 - 5.87E+6 - 1.76E+7 5.58E+6 Mn-56 - 1.31 E-2 - 1.58E-2 - 1.90E+0 2.95E-3 Fe-55 1.12E+8 5.93E+7 - - 3.35E+7 1.1 OE+7 1.84E+7 Fe-59 1.20E+8 1.95E+8 - - 5.65E+7 2.03E+8 9.71 E+7 Co-57 - 3.84E+6 - - - 3.14E+7 7.77E+6 Co-58 - 1.21 E+7 - - - 7.08E+7 3.72E+7 Co-60 - 4.32E+7 - - - 2.39E+8 1.27E+8 Ni-63 2.96E+10 1.59E+9 - - - 1.07E+8 1.01E+9 Ni-65 1.66E+0 1.56E-1 - - - 1.91E+1 9.11E-2 Cu-64 - 7.55E+4 - 1.82E+5 - 3.54E+6 4.56E+4 Zn-65 4.13E+9 1.1 OE+10 - 6.94E+9 - 1.93E+9 6.85E+9 Zn-69 - - - - - 2.14E-9 Br-82 - - - - - - 1.15E+8 Br-83 - - - - - - 4.69E-1 Br-84 Br-85 Rb-86 - 8.77E+9 - - - 5.64E+8 5.39E+9 Rb-88 Rb-89 - -

Sr-89 6.62E+9 - - - 2.56E+8 1.89E+8 Sr-9O 1.12E+11 - - - - 1.51E+9 2.83E+10 Sr-91 1.41 E+5 - - - - 3.12E+5 5.33E+3 Sr-92 2.19E+0 - - - 4.14E+1 8.76E-2 Y-90 3.22E+2 - - - - 9.15E+5 8.61 E+O Y-9lm -

Y-91 3.91 E+4 - - - 5.21 E+6 1.04E+3 Y-92 2.46E-4 - - - 7.10E+O 7.03E-6 Y-93 1.06E+0 - - - - 1.57E+4 2.90E-2 Zr-95 3.84E+3 8.45E+2 - 1.21 E+3 - 8.81 E+5 7.52E+2 Zr-97 1.89E+0 2.72E-1 3.91 E-1 - 4.13E+4 1.61 E-1 Nb-95 3.18E+5 1.24E.5 1.16E+5 - 2.29E_8 8.84E+4 Nb-97 - - - - 1.45E-6 -

Mo-99 - 8.29E+7 - 1.77E+8 - 6.86E+7 2.05E+7 Tc-99m 1.29E+1 2.54E+1 - 3.68E+2 1.29E+1 1.44E+4 4.20E+2 Tc-101 - -

2-35 REV. 9 12/02/2005

Table 2.13 RI Grass-Cow-Milk Pathway Dose Factors - CHILD (mremlyr per gCim 3 ) for H-3 and C-14 (M2 x mremlyr gCIsec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body -

Ru-103 4.29E+3 - - 1.08E+4 - 1.11E+5 1.65E+3 Ru-105 3.82E-3 - 3.36E-2 - 2.49E+0 1.39E-3 Ru-1 06 9.24E+4 - - 1.25E+5 - 1.44E+6 1.15E+4 Rh-103m - - - - - - -

Rh-106 - - - - - -

-

Ag-1i1m 2.09E+8 1.41E+8 - 2.63E+8 - 1.68E+10 1.13E+8 Sb-1 24 1.09E+8 1.41 E+8 2.40E+5 - 6.03E+7 6.79E+8 3.81 E+7 Sb-125 8.70E+7 1.41 E+6 8.06E+4 - 4.85E+7 2.08E+8 1.82E+7 Te-125m 7.38E+7 2.OOE+7 2.07E+7 - - 7.12E+7 9.84E+6 Te-127m 2.08E+8 5.60E+7 4.97E+7 5.93E+8 - 1.68E+8 2.47E+7 TG- 27 3.06E+3 8.25E+2 2.12E+3 8.71 E+3 - 1.20E+5 6.56E+2 Te-129m 2.72E+8 7.61 E+7 8.78E+7 8.OOE+8 - 3.32E+8 4.23E+7 Ta-1 29 - - - 2.87E-9 - 6.12E-8 Te-131m 1.60E+6 5.53E+5 1.14E+6 5.35E+6 - 2.24E+7 5.89E+5 Te-1 31 - - - - - - -

Te-1 32 1.02E+7 4.52E+6 6.58E+6 4.20E+7 - 4.55E+7 5.46E+6 1-130 1.75E+6 3.54E+6 3.90E+8 5.29E+6 - 1.66E+6 1.82E+6 1-131 1.30E+9 1.31E+9 4.34E+11 2.15E+9 - 1.17E+8 7.46E+8 1-132 6.86E-1 1.26E+0 5.85E+1 1.93E+O - 1.48E+0 5.80E-1 1-133 1.76E+7 2.18E+7 4.04E+9 3.63E+7 - 8.77E+6 8.23E+6 1-134 -

1-135 5.84E+4 1.05E+5 9.30E+6 1.61 E+5 - 8.OOE+4 4.97E+4 Cs-134 2.26E+10 3.71E+10 - 1.15E+10 4.13E+9 2.OOE+8 7.83E+9 Cs-1 36 1.OOE+9 2.76E+9 - 1.47E+9 2.19E+8 9.70E+7 1.79E+9 CsO137 3.22E+10 3.09E+10 - 1.01E+10 3.62E+9 1.93E+8 4.55E+9 Cs-138 Ba-139 2.14E-7 - - - - 1.23E-5 6.19E-9 Ba-140 1.17E+8 1.03E+5 - 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ba-141 Ba-142 La-140 1.93E+1 6.74E+0 - - - 1.88E+5 2.27E+O La-142 - - - - - 2.51 E-6 Ce-141 2.19E+4 1.09E+4 - 4.78E+3 - 1.36E+7 1.62E+3 Ce-143 1.89E+2 1.02E+5 - 4.29E+1 - 1.50E+6 1.48E+1 Ca-144 1.62E+6 5.09E+5 - 2.82E+5 - 1.33E+8 8.66E+4 Pr-143 7.23E+2 2.17E+2 - 1.17E+2 - 7.80E+5 3.59E+1 Pr-144 Nd-147 4.45E+2 3.60E+2 - 1.98E+2 - 5.71 E+5 2.79E+1 W-187 2.91 E+4 1.72E+4 - - - 2.42E+6 7.73E+3 Np-239 1.72E+1 1.23E+0 - 3.57E+0 - 9.14E+4 8.68E-1 2-36 REV. 9 12/02/2005

Table 2.14 RI Grass-Cow-Milk Pathway Dose Factors - INFANT (mremnyr per pC/m 3 ) for H-3 and C-14 (M2 x mrem/yr gCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 - 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 C-14 3.23E+6 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 Na-24 1.61 E+7 1.61 E+7 1.61 E+7 1.61 E+7 1.61 E+7 1.61 E+7 1.61 E+7 P-32 1.60E+11 9.42E+9 - - - 2.17E+9 6.21 E+9 Cr-51 - - 1.05E+5 2.30E+4 2.05E+5 4.71 E+6 1.61 E+5 Mn-54 - 3.89E+7 8.63E+6 - 1.43E+7 8.83E+6 Mn-56 - 3.21 E-2 - 2.76E-2 - 2.91 E+O 5.53E-3 Fe-55 1.35E+8 8.72E+7 - - 4.27E+7 1.11 E+7 2.33E+7 Fe-59 2.25E+8 3.93E+8 - - 1.16E+8 1.88E+8 1.55E+8 Co-57 8.95E+6 - - 3.05E+7 1.46E+7 Co-58 2.43E+7 - - 6.05E+7 6.06E+7 Co-60 - 8.81 E+7 - - - 2.1OE+8 2.08E+8 Ni-63 3.49E+10 2.16E+9 - - - 1.07E+8 1.21 E+9 Ni-65 3.51 E+O 3.97E-1 - - 3.02E+1 1.81 E-1 Cu-64 - 1.88E+5 - 3.17E+5 - 3.85E+6 8.69E+4 Zn-65 5.55E+9 1.90E+1 0 - 9.23E+9 - 1.61 E+1 0 8.78E+9 Zn-69 - - - - - 7.36E-9 Br-82 - - - - - 1.94E+8 Br-83 - - - - - 9.95E-1 Br-84 Br-85 Rb-86 - 2.22E+10 - - - 5.69E+8 1.10E+10 Rb-88 - - - - - -

Rb-89 - - - - -

Sr-89 1.26E+1 0 - - - - 2.59E+8 3.61 E+8 Sr-90 1.22E+11 - - - - 1.52E+9 3.1OE+10 Sr-91 2.94E+5 - - - 3.48E+5 1.06E+4 Sr-92 4.65E+O - - - - 5.01 E+1 1.73E-1 Y-90 6.80E+2 - - - - 9.39E+5 1.82E+1 Y-91m Y-91 7.33E+4 - - - 5.26E+6 1.95E+3 Y-92 5.22E-4 - - - - 9.97E+0 1.47E-5 Y-93 2.25E+0 - - - - 1.78E+4 6.13E-2 Zr-95 6.83E+3 1.66E+3 - 1.79E+3 - 8.28E+5 1.18E+3 Zr-97 3.99E+O 6.85E-1 6.91 E-1 - 4.37E+4 3.13E-1 Nb-95 5.93E+5 2.44E+5 - 1.75E+5 - 2.06E+8 1.41 E+5 Nb-97 - - - - 3.70E-6 -

Mo-99 - 2.12E+8 - 3.17E+8 - 6.98E+7 4.13E+7 Tc-99m 2.69E+1 5.55E+1 - 5.97E+2 2.90E+1 1.61 E+4 7.15E+2 Tc-101 - - - - - -

2-37 REV. 9 12/02/2005

Table 2.14 R, Grass-Cow-Milk Pathway Dose Factors - INFANT (mremlyr per p.CUm3 ) for H-3 and C-1 4 (m 2 x mremlyr I+/-CI/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body Ru-103 8.69E+3 - 1.81 E+4 - 1.06E+5 2.91 E+3 Ru-1 05 8.06E-3 - - 5.92E-2 - 3.21 E+O 2.71 E-3 Ru-106 1.90E+5 - - 2.25E+5 - 1.44E+6 2.38E+4 Rh-103m Rh-1060 - - -

Ag-11 Om 3.86E+8 2.82E+8 - 4.03E+8 - 1.46E+1 0 1.86E+8 Sb-124 2.09E+8 3.08E+6 5.56E+5 - 1.31 E+8 6.46E+8 6.49E+7 Sb-125 1.49E+8 1.45E+6 1.87E+5 - 9.38E+7 1.99E+8 3.07E+7 Te-125m 1.51 E+8 5.04E+7 5.07E+7 - - 7.18E+7 2.04E+7 Te-127m 4.21 E+8 1.40E+8 1.22E+8 1.04E+9 - 1.70E+8 5.10E+7 Te-127 6.50E+3 2.18E+3 5.29E+3 1.59E+4 - 1.36E+5 1.40E+3 Te-129m 5.59E+8 1.92E+8 2.15E+8 1.40E+9 - 3.34E+8 8.62E+7 Te-129 2.08E-9 - 1.75E-9 5.18E-9 - 1.66E-7 Te-131m 3.38E+6 1.36E+6 2.76E+6 9.35E+6 - 2.29E+7 1.12E+6 Te-131 - - - - - -

Te-1 32 2.10E+7 1.04E+7 1.54E+7 6.51 E+7 - 3.85E+7 9.72E+6 1-130 3.60E+6 7.92E+6 8.88E+8 8.70E+6 - 1.70E+6 3.18E+6 1-131 2.72E+9 3.21 E+9 1.05E+12 3.75E+9 - 1.15E+8 1.41 E+9 1-132 1.42E+0 2.89E+0 1.35E+2 3.22E+0 2.34E+0 1.03E+O 1-133 3.72E+7 5.410E+7 9.84E+9 6.36E+7 - 9.16E+6 1.58E+7 1-134 - - 1.01 E-9 1-135 1.21E+5 2.41 E+5 2.16E+7 2.69E+5 1 8.74E+4 8.80E+4 Cs-134 3.65E+10 6.80E+1 - 1.75E+10 7.18E+9 1.-8E+8 6.87E+9 Cs-136 1.96E+9 5.77E+9 - 2.30E+9 4.70E+8 8.76E+7 2.15E+9 Cs-137 5.15E+10 6.02E+10 1.62E+10 6.55E+9 1.88E+8 4.27E+9 Cs-1 38 - - - - - -

Ba-139 4.55E-7 - - - - 2.88E-i 1.32E-8 Ba-1 40 2.41 E+8 2.41 E+5 - 5.73E+4 1.48E+5 5.92E+7 1.24E+7 Ba-141 Ba-142 La-140 4.03E+1 1.59E+1 - - - 1.87E+5 4.09E+O La-142 - - - - - 5.21E-6 Ce-141 4.33E+4 2.64E+4 - 8.15E+3 - 1.37E+7 3.11E+3 Ce-143 4.00E+2 2.65E+5 - 7.72E+1 - 1.55E+6 3.02E+1 Ce-144 2.33E+6 9.52E+5 - 3.85E+5 - 1.33E+8 1.30E+5 Pr-1 43 1.49E+3 5.59E+2 - 2.08E+2 - 7.89E+5 7.41 E+1 Pr-1 44 Nd-147 8.82E+2 9.06E+2 - 3.49E+2 - 5.74E+5 5.i5E+1 W-187 6.12E+4 4.26E+4 - - - 2.50E+6 1.47E+4 Np-239 3.64E+1 3.25E+0 - 6.49E+0 - 9.40E+4 1.84E+0 2-38 REV. 9 12/02/2005

Table 2.15 RI Ground Plane Pathway Dose Factors (m 2 x mremfyr per gLCisec)

Nuclide Any Organ H-3 C-14 Na-24 1.21 E+7 P-32 Cr-51 4.68E+6 Mn-54 1.34E+9 Mn-56 _9.05E+5 Fe-55 Fe-59 2.75E+8 Co-57 4.37E+8 Co-58 3.82E+8 Co-60 2.16E+10 Ni-63 Ni-65 2.97E+5 Cu-64 6.09E+5 Zn-65 7.45E+8 Zn-69 Br-82 4.57E+7 Br-83 4.89E+3 Br-84 2.03E+5 Br-85 Rb-86 8.98E+6 Rb-88 3.29E+4 Rb-89 1.21 E+5 Sr-89 2.16E+4 Sr-90 Sr-91 2.19E+6 Sr-92 7.77E+5 Y-90 4.48E+3 Y-91m 1.01 E+5 Y-91 1.08E+6 Y-92 1.80E+5 Y-93 1.85E+5 Zr-95 2.48E+8 Zr-97 2.94E+6 Nb-95 1.36E+8 Nb-97 2.28E+6 Mo-99 4.05E+6 Tc-99m 1.83E+5 Tc-101 2.04E+4 Ru-103 1.09E+8 2-39 REV. 9 12/02/2005

Table 2.15 RI Ground Plane Pathway Dose Factors (m2 x mrem/yr per gCisec)

Nuclide l Any Organ Ru-105 6.36E+5 Ru-106 4.21 E+8 Rh-1 03m _

Rh-1 06 Ag-110m 3.47E+9 Sb-124 2.87E+9 Sb-1 25 6.49E+9 Te-125m 1.55E+6 Te-127m 9.17E+4 Te-127 3.OOE+3 Te-129m 2.OOE+7 Te-129 2.60E+4 Te-131m 8.03E+6 Te-1 31 2.93E+4 Te-1 32 4.22E+6 1-130 5.53E+6 1-131 1.72E+7 1-132 1.24E+6 1-133 2.47E+6 1-134 4.49E+5 1-135 2.56E+6 Cs-134 6.75E+9 Cs-1 36 1.49E+8 Cs-1 37 1.04E+10 Cs-138 3.59E+5 Ba-1 39 1.06E+5 Ba-140 2.05E+7 Ba-141 4.18E+4 Ba-142 4.49E+4 La-140 1.91E+7 La-142 7.36E+5 Ce-1 41 1.36E+7 Ca-143 2.32E+6 Ce-1 44 6.95E+7 Pr-143 Pr-144 1.83E+3 Nd-147 8.40E+6 W-187 2.36E+6 Np-239 1.71 E+6 2-40 REV. 9 12/02/2005

'4 RADIOLOGICAL EFFLUENT SPECIFICATIONS AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY AND SURVEILLANCE REQUIREMENTS SPECIFICATIONS 3.0.1. Compliance with the specifications contained in the succeeding text is required during the conditions specified therein; except that upon failure to meet the specifications, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a Specification shall exist when its requirements and associated ACTION requirements are not met within the specified time intervals. If the Specification is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required.

3.0.3 When a Specification is not met, except as provided in the associated ACTION requirements, reporting pursuant to TS 6.9.b.3 will be initiated.

SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the conditions specified for individual Specifications unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Specification. Exceptions to these requirements are stated in the individual Specification. Surveillance Requirements do not have to be performed on inoperable equipment.

3-1 REV. 9 12/02/2005

3/4.1 RADIOACTIVE LIQUID EFFLUENTMONITORING INSTRUMENTATION SPECIFICATIONS 3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 3.3.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the methodology in Section 1.0 of the OFF-SITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY During release via the monitored pathway.

ACTION

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.1.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS 4.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.1.

BASIS Radioactive Liquid Effluent Monitoring Instrumentation - The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding ten (10) times the values of 10 CFR Part 20, Appendix B, Table 2, Column 2. The operability and use of this instrumentation is consistent with the appropriate requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3-2 REV. 9 12/0212005

3/4.2 RADIOACTIVE GASEOUS EFFLUENTMONITORING INSTRUMENTATION SPECIFICATIONS 3.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.2 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 3.4.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the methodology in section 2.0 of the ODCM.

APPLICABILITY As shown in Table 3.2.

ACTION

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.2.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS 4.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.2.

BASIS Radioactive Gaseous Effluent Monitoring Instrumentation - The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip will occur prior to exceeding the dose rate limits of ODCM Specification 3.4.1. The operability and use of this instrumentation is consistent with the appropriate requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3-3 REV. 9 12/02/2005

3/4.3 LIQUID EFFLUENTS CONCENTRATION SPECIFICATIONS 3.3.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 104 ;Ci/ml total activity.

APPLICABILITY During release via the monitored pathway.

ACTION With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.3.

4.3.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of ODCM Specification 3.3. 1.

BASIS Concentration - This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR Part 20, Appendix B. Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix L 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its concentration limit in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP)

Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

34 REV. 9 12/02/2005

DOSE SPECIFICATIONS 3.3.2 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 rnrem to the total body and to less than or equal to 5 nerem to any organ, and

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

APPLICABILITY At all times.

ACTION With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification (TS) 6.9.b.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the release and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 4.3.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and

,_ parameters in the ODCM once per 31 days.

BASIS Dose - This specification is provided to implement the requirements of Sections II.A, Ill.A and IV.A of Appendix L 10 CFR 50. The Limiting Condition for Operation implements the guides set forth in Section Il.A of Appendix L. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in the ODCM implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113. "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

3-5 REV. 9 12/02/2005

LIQUID RADWASTE TREATMENT SYSTEM SPECIFICATIONS 3.3.3 The liquid radwaste treatment system as described in the ODCM shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses, due to the liquid effluent, to UNRESTRICTED AREAS would exceed 0.18 mrem to the total body or 0.62 mrem to any organ in a calendar quarter.

APPLICABILITY At all times.

ACTION

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission-within 30 days pursuant to TS 6.9.b.3, a Special Report that includes the following information:

I. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action(s) taken to prevent a recurrence.

SURVEILLANCE REQUIREMENTS 4.3.3 Doses due to liquid releases from the unit to UNRESTRICTED AREAS shall be projected once per 31 days in accordance with the methodology and parameters in the ODCM.

BASIS Liquid Radwaste Treatment System - The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

3-6 REV. 9 12/0212005

3/4.4 GASEOUS EFFLUENTS DOSE RATE SPECIFICATIONS 3.4.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY At all times.

ACTION

a. With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 4.4.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4.4.1.2 The dose rate due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.4.

3-7 REV. 9 12/02/2005

BASIS Dose Rate - This specification is provided to ensure that the dose rates at any time to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY are less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin. This also restricts releases, at all times, for the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/yr.

These dose rate limits provide additional assurance that radioactive material discharged in gaseous effluents will be maintained ALARA, and coupled with the requirements of ODCM Specification 3.4.2, ensure that the exposures of MEMBERS OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, will not exceed the annual average concentrations specified in Appendix B, Table 2, Column 1 of 10 CFR 20. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, LA., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3-8 REV. 9 12/02/2005

DOSE - NOBLE GASES SPECIFICATIONS 3.4.2 The air dose due to noble gases released in gaseous effluents, to areas at and 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 mriad for beta radiation and,

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

APPLICABILTY At all times.

ACTION

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to TS 6.9.b.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 4.4.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM once per 31 days.

BASIS Dose - Noble Gases - This specification is provided to implement the requirements of Sections II.B, H.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section I1.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix L" Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,"

Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

3-9 REV. 9 12/02/2005

DOSE - IODINE-131, IODINE-133, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM SPECIFICATIONS 3.4.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY shall be limited to the following:

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

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

APPLICABILITY At all times.

ACTION

a. With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to TS 6.9.b.3, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 4.4.3 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM once per 31 days.

BASIS Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form -

This specification is provided to implement the requirements of Sections ll.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section fl.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."

3-10 REV. 9 12/02/2005

The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculationaJ methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1,July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3-11 REV. 9 12/02/2005

GASEOUS RADWASTE TREATMENT SYSTEM SPECIFICATIONS 3.4.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY would exceed 0.62 mrad for gamma radiation and 1.25 mrad for beta radiation in a calendar quarter. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, to areas at and beyond the SITE BOUNDARY would exceed 0.94 mrem to any organ in a calendar quarter.

APPLICABILITY At all times.

ACTION

a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to TS 6.9.b.3, a Special Report that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action(s) taken to prevent a recurrence.

SURVEILLANCE REQUIREMENTS 4.4.4 Doses due to gaseous releases from areas at and beyond the SITE BOUNDARY shall be projected once per 31 days in accordance with the methodology and parameters in the ODCM.

3-12 REV. 9 12/02/2005

BASIS Gaseous Radwaste Treatment System - The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections ll.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

3-13 REV. 9 12/02/2005

3/4.5 TOTAL DOSE SPECIFICATIONS 3.5 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less 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.

APPLICABILITY At all times.

ACTION

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of ODCM Specification 3.3.2.a, 3.3.2.b, 3.4.2.a, 3.4.2.b, 3.4.3.a, or 3.4.3.b, calculations should be made including direct radiation contributions from the reactor unit to determine whether the above limits have been exceeded. If such is the case in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to TS 6.9.b.3, a special report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This special report as defined in 10 CFR 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the special report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS 4.5.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillance Requirements 4.3.2, 4.4.2, and 4.4.3 in accordance with the methodology and parameters in the ODCM.

4.5.2 Cumulative dose contributions from direct radiation from the reactor unit shall be determined in accordance with the methodology and parameters in the ODCM.

This requirement is applicable only under conditions set forth in ODCM Specification 3.5.a.

3-14 REV. 9 12/02/2005

BASIS Total Dose - This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the reactor remains within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in ODCM Specifications 3.3.1 and 3.4.1. An I individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

3-15 REV. 9 12/02/2005

314.6 REPORTING REQUIREMENTS 3/4.6.1 Radioactive Effluent Release Report The Radioactive Effluent Release Report shall include the following:

a. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit following the format of Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974.

b. An annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability 4. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit during the previous calendar year. The assumptions used in making these assessment, i.e., specific activity, exposure time and location shall be included in these reports. The assessment of radiation doses shall be performed based on the calculational guidance, as presented in the ODCM.

c. An assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

d. A list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

e. Any changes made during the reporting period to the ODCM.

4 In lieu of submission with the annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

3-16 REV. 9 12102/2005

TABLE 3.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Action Operable

1. Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line (R-18) 1 1

b. Steam Generator Blowdown Effluent Line (R-19) 1 2

2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release

a. Service Water System Effluent Line (Component cooling, R-20) 1 3

b. Service Water System Effluent Line (Containment fan cooling, R-16) 1 3 Action 1 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Surveillance Requirement 4.3.1.1 and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

Action 2 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of 1.OE-6 uCi/ml:

a. At least once per week with no indication of primary-to-secondary leakage; or

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with identified primary-to-secondary leakage (with secondary side activity > 1.OE-05 uCi/ml)

Action 3 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of 1.OE-6 uCi/ml. (Note: Failure to complete sampling and analysis prior to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the monitor is declared O.O.S. is a violation of this specification).

3-17 REV. 9 12/02/2005

TABLE 3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Page 1 of 2)

Minimum Instrument Channels Applicability Action Operable

1. Noble Gas Activity Monitor

a. R-13 or R-14

- Waste Gas Holdup System (auto-isolation)

- Auxiliary Building Ventilation System 1 *

- Containment Purge 2" line 4 (auto-isolation) 5

b. R-12 or R-21 6

- Containment purge 36" duct 1 *

(auto-isolation) 6

c. R-15 1 *

- Condenser Evacuation System 5

2. Radioiodine & Particulate Samplers

a. Containment Building Vent (R-21)

b. Auxiliary Building Vent (R-13 or 1

• 7 R-14)1 7

3. Sampler Flow Rate Measuring Devices

a. Containment Building Vent Sampler (R-21)

b. Auxiliary Building Vent Sampler 1

• 8 (R-13 orR-14) 1

• 8

• At all times 3-18 REV. 9 12/02/2005

TABLE 3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Page 2 of 2)

TABLE NOTATIONS Action 4 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

Action 5 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Action 6 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

Action 7 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.4.

Action 8 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3-19 REV. 9 12/02/2005

TABLE 4.0 FREQUENCY NOTATION Notation l Frequency 5 S Once per shift St Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W Once per 7 days M Once per 31 days 0 Once per 92 days

-

SA Once per 184 days R Once per refueling cycle, not to exceed 18 months P Prior to each reactor startup if not done previous week PR Completed prior to each release NA Not applicable fA maximum extension not to exceed 25% of the surveillance interval.

3-20 REV. 9 12/02/2005

TABLE 4.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REREMENTSr Channel Channel Source Channel Functional Instrument Check Check Calibration Test Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line (R-18) D PR R Q

b. Steam Generator Blowdown Effluent D M R Q Line (R-19)

2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release

a. Service Water System Effluent Line D M R Q (Component cooling, R-20)

b. Service Water System Effluent Line D M R Q (Containment fan cooling, R-1 6) 3-21 REV. 9 12/02/2005

TABLE 4.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes In Which Channel Source Channel Functional Surveillance Instrument Check Check Calibration Test Required

1. Noble Gas Activity Monitor

a. R-13 orR-14 Waste Gas Holdup System PR PR R Q *

(auto-isolation)

Auxiliary Building Ventilation D M R Q

• System Containment Purge 2" line D M R Q *

(auto-isolation)

b. R-12 or R-21 Containment purge 36" duct D PR R Q *

(auto-isolation)

c. R-15 Condenser Evacuation System D M R Q *

2. Radioiodine Particulate Samplers

a. Containment Building vent W NA NA NA *

(R-21)

b. Auxiliary Building vent (R-13 W NA NA NA

• or R-14)

3. Sampler Flow Rate Measuring Devices

a. Containment Building vent D NA R Q

• sampler (R-21)

b. Auxiliary Building vent sampler D NA R Q *

(R-13 orR-14)

• At all times other than when the line is valved out and tagged.

3-22 REV. 9 12/02/2005

TABLE 43 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Page 1 of 2 Minimum Lower Limit of Sampling Analysis Type of Activity Detection (LLD)r Liquid Release Type Frequency Frequency Analysis (ACitml) 6 A. Batch Waste Release PR PR Principal Gamma Ix10 Tanksb Each Batch Each Batch Emitters' I-131 .lx106 PR M H-3 lx10-5 Each Batch Composited Gross Alpha 5x10 '

PR Q Sr-89, Sr-90 5x10-'

Each Batch Composited Fe-55 1x10 B. Continuous Releases' W W Principal Gamma (SG Blowdown) Grab Sample Grab Sample Emitters' 5xl0 7 (TB Sumps) 1-131 1x10 4 W M H-3 1x10 5 Grab Sample Compositef Gross Alpha 5xI0-7 W Q Sr-89, Sr-90 5x10 4 Grab Sample Compositer Fe-55 Ix10 4 3-23 REV. 9 12/02/2005

TABLE 4.3 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Page 2 of 2 Table Notations 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:

LLD =(-\t E *V

• 2. 22 x 106 *Y*exp(Aft)

Where:

• LLD is the a oriof lower limit of detection as defined above, as H~i per unit mass or volume,

• sb is the standard deviation of the background counting rate or of the counting 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 106 is the number of disintegrations per minute per microcurie,

• Y is the fractional radiochemical yield, when applicable,

• Ais the radioactive decay constant for the particular radionuclide, and

• At for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

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

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 nosteniori (after the fact) limit for a particular measurement.

b A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, each batch shall be located, and then thoroughly mixed to ensure representative sampling.

c The principal gamma emitters for which the LLD specification applies exclusively are 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 be analyzed and reported in the Radioactive Effluent Release Report pursuant to TS 6.9.b.2.

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.

e 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.

f As a minimum, the monthly and quarterly composite samples shall be comprised of weekly grab samples.

g During periods of identified primary-to-secondary leakage (with the secondary activity > 1.OE-05 pCi/ml),

grab samples are collected daily and analyzed by gamma spectroscopy.

3-24 REV. 9 12/02/2005

TABLE 4.4 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Page 1 of 2 Lower Limit of Minimum Detection Sampling Analysis Type of Activity (LLD)'

Gaseous Release Type Frequency Frequency Analysis (UCi/mI)

A. Waste Gas Storage PR PR Principal Gamma lxl04 Tank Each Tank Each Tank Emittersb Grab

- Sample B. Containment PURGE PR PR Principal Gamma lx104 Each Each Purge Emittersb PURGE Grab Sample C. Auxiliary Building and M M Principal Gamma Containment Building Grab Emittersb lx 104 Vent Sample W I-131 3xl1-0 2 Continuousc Charcoal Sample W Principal Gamma 1x10"-

Continuousc Particulate Emitterb Sample (a-131,others)

M Gross Alpha lx10-"

Continuousc Composite Particulate Sample Q SR-89, SR-90 1x10 "

Continuousc Composite Patriculate Sample lx104 Noble Gas Noble Gases Continuousc Monitor Gross Beta or Gamma 3-25 REV. 9 12/02/2005

TABLE 4A RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Page 2 of 2 I. Table Notations 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:

LLD= 4.66

• sb, E

• V

• 2.22 Xlo6 *Y

• exp(os)

Where:

• LLD is the a priori lower limit of detection as defined above, as 1Ci per unit mass or volume,

• sb is the standard deviation of the background counting rate or of the counting 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 106 is the number of disintegrations per minute per microcurie,

• Y is the fractional radiochemical yield, when applicable,

• X 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 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 The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to TS 6.9.b.2.

c The ratio of the sample flow rate to the sampled flow stream flow rate shall be known (based on sampler and ventilation system flow measuring devices or periodic flow estimates) for the time period covered by each dose or dose rate calculation made in accordance with ODCM Specifications 3.4.1, 3.4.2, and 3.4.3.

3-26 REV. 9 12/02/2005

APPENDIX A TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS - LIQUID RADIOACTIVE EFFLUENTS A-1 REV. 9 12102/2005

Technical Basis for Effective Dose Factors -

Liquid Effluent Releases To verify that the current approach to determining environmental doses using a simplified method has remained consistent since the previous analysis (performed using effluent data from 1981-1983), a similar evaluation was performed using the liquid effluent release data from 2000-2002. From the effluent data, the dose contribution of the radionuclide mixture can be obtained to provide a simplified method of determining compliance with the dose limits of ODCM Specification 3.3.2. For the radionuclide distribution of effluents from the Kewaunee Power Station, the controlling organ is either the GI-LLI or the liver. The calculated GI-LLI dose is almost exclusively dictated by the Nb-95 releases; the liver dose is mostly a function of the Cs-134 and Fe-55 releases. The radionuclides, Fe-55, Co-58, Co-60, Sr-90, and Cs-137 contribute essentially all of the calculated total body dose. The results of this evaluation are presented in Table A-1. The individual nuclide doses used in the dose comparisons of Table A-1 were calculated using the total curies released via batch and continuous releases as reported in the Annual Radioactive Effluent Release Report, weighted by the appropriate dose factors.

Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is negligible. From 2000-2002, the maximum tritium release from the Kewaunee Nuclear Plant to Lake Michigan was 270 curies.

The calculated total body dose from such a release is 1.36E-02 mrem/yr via the fish ingestion and drinking water pathways. This amounts to 0.07% of the design objective dose of 3 mrem/yr.

Furthermore, the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation.

For purposes of simplifying the details of the dose calculational process, it is conservative to identify a controlling, dose significant radionuclide and limit the calculational process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculational method that is simplified while also being conservative.

While not present in the 2000-2002 liquid effluent releases, it still remains conservative to use the Cs-134 dose conversion factor (7.09E+05 mrem/hr per ,uCi/ml, liver) to evaluate the maximum organ dose. Only the reactor-generated radionuclide Nb-95 has a higher dose conversion factor (1.5 1E06 mrem/hr per gCi/ml, GI-LLI). However, since Nb-95 releases are typically less than 5% of the total releases, it is conservative to use the Cs-134 factor. By this approach, the maximum organ dose will be routinely overestimated. For 2000, using this simplified conservative method (CW value of 2.00E+05 gpm) would overestimate the maximum organ dose as reported in the Annual Radioactive Effluent Release Report by a factor of 234; for 2001, the conservatism is a factor of 109; and for 2002, a factor of 730. This comparison is shown in Table A-2.

A-2 REV. 9 12/02/2005

For the total body calculation, the Cs-134 dose factor (5.79E+05 mrem/hr per gCi/ml, total body) is again used since it is higher than the identified dominant nuclides. For 2000, using this simplified conservative dose calculational method would overestimate the total body dose by a factor of 253; for 2001, the conservatism is a factor of 105; and for 2002, a factor of 601.

For evaluating compliance with the dose limits of ODCM Specification 3.3.2 the following simplified equations may be used:

Total Body Dt= 1.67E -02x VOL xAC. -,34. TB xEC (A. 1) where:

Dtb = dose to the total body (mrem)

Acs-134,TB = 5.79E+05, total body ingestion dose conversion factor for Cs-134 (mrem/hr per ICi/ml)

VOL = volume of liquid effluent released (gal)

IQj = total concentration of all radionuclides (gCi/ml)

CW = average circulating water discharge rate during release period (gal/min) 1.67E-02 = conversion factor (hr/min)

Substituting the value for the Cs-134 total body dose conversion factor, the equation simplifies to:

Dtb XECi (A.2)

Maximum Organ Dr= = 1.6 7E-02xVOLxAcs -134, L X E C (A.3)

CW where:

D~x = maximum organ dose (mrem)

ACS-134,L = 7.09E+05, liver ingestion dose conversion factor for Cs-134 (mrem/hr per pCi/mR)

A-3 REV. 9 12/02/2005

Substituting the value for Acs-134,uver the equation simplifies to:

DX = 1.18E+04xVOL EC (A.4)

CW Only the total body dose need be evaluated by this simplified method since it represents the more limiting (compared with the maximum organ dose) for demonstrating compliance with ODCM Specification 3.3.2.

A4 REV. 9 12/02/2005

I [ r r I r I r Table A-1 Adult Dose Contributions Fish and Drinking Water Pathways

. 2000 2001. 2002 Radio- Release TB GI-LLI Liver Release TB GI-LLI Liver Release TB GI-LLI Liver nuclide (Ci) Dose Dose Dose (Ci) Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac.

Fe-55 4.8 IE- 0.03 0.02 0.l0 4.85E- 0.04 0.03 0.13 3.69E- 0.19 0.02 0.84

__ _ _ 02 _ _ __02 _ _ __ 02_ _ _ _

Co-58 8.07E- 0.01 0.03 _ 4.09E- 0.01 0.02 _ 4.94E- 0.05 0.02 0.02

__ _ _ 03 _ _ _ _ _ _ _ 03 _ _ _ 03 Fe-59 2.77E- 0.01 0.02 04 __ 04 ___04 __

Co-60 4.71E 0.02 0.04 0.01 4.31E- 0.02 0.05 0.01 2.07E- 0.06 0.02 0.03 03 _ _ _ __ _ _ _ _ _ _ 03 __ _ _ _ _ _03 _ _

Br-82 4.94E- 0.01 1.44E- N/D 04 _ _ _ _ _ _ _ _ 04 _ _ _

Sr-90 2.25E- 0.18 0.01

• 2.50E- 0.25 0.01 9.76E- 0.63 04 _ _ _ _ 04 _ _ _ 05_ _ _ _

Nb-95 3.41E-

• 0.89 - 2.39E- 0 0.86 . 2.45E- _ 0.91 _

04 2.74E- 0.68 0.01 0.85 3.04E- 0.0.

C-3 370E4 0.75 0.01 0.88 0406 O0 .5 06 0 .8

• Less than 0.01 NID = not detected A-5 REV. 9 12/02/2005

Table A-2 'I.

Table A-2 Adult Liver and Total Body Dose Assessment Dose Via the Simplified Method Versus the Actual Calculated Dose l 2000 2001 2002 i Simplified liver Dose (mRem)* 1.16E+00 9.87E-01 7.88E-01 Actual liver Dose (mRem)** 4.97E-03 9.02E-03 l.08E-03 Simplified divided by Actual 234 109 730 Simplified Total Body Dose I (mRem)

• 9.53E-01 8.09E-01 6.46E-01 l Actual Total Body Dose (mRem)I

• • 3.77E-03 7.73E-03 1.07E-03 Simplified divided by Actual 253 105 601 l

• Assuming 2.OOE+05 gpm circulating water flow

• • From the Annual Radioactive Effluent Release Report A-6 REV. 9 12/02/2005

APPENDIX B TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS -

GASEOUS RADIOACTIVE EFFLUENTS B-1 REV. 9 12/02/2005

APPENDIX B Technical Basis for Effective Dose Factors -

Gaseous Radioactive Effluents Overview The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose factors, which are radionuclide specific. These effective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication (Kcff, Mwff or Neff) times the total quantity of radioactive material released would be needed). This approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique.

Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:

• Keff= (K x fi) (B. 1) where:

Kff = the effective total body dose factor due to gamma emissions from all noble gases released Ki = the total body dose factor due to gamma emissions from each noble gas radionuclide "i" released fi = the fractional abundance of noble gas radionuclide "i" relative to the total noble gas activity (L~llM~tf

[(i +~lM~fi](B.2)

=E where:

(L + 1.1 M)eff = the effective skin dose factor due to beta and gamma emissions from all noble gases released (L1 + 1.1 M1) = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide "i" released Meff =2 (M xfi) (B.3) where:

MKff = the effective air dose factor due to gamma emissions from all noble gases released B-2 REV. 9 12/02/2005

Mi = the air dose factor due to gamma emissions from each noble gas radionuclide "i" released Neff =E(Ni x fi) (B.4) where:

Neff = the effective air dose factor due to beta emissions from all noble gases released Ni = the air dose factor due to beta emissions from each noble gas radionuclide "i" released Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors. However, the noble gas releases from Kewaunee have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult. For the years of 2000, 2001 and 2002, the total noble gas releases have been limited to 2.54E-04 Ci for 2000, 1.37E-01 Ci for 2001, and 1.91E-02 Ci for 2002. Therefore, in order to provide a reasonable basis for the derivation of the effective noble gas dose factors, the primary coolant source term from ANSI N237-1976/ANS-18.1, "Source Term Specifications," has been used as representing a typical distribution. The effective dose factors as derived are presented in Table B-i.

Auplication To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose calculational process when the effective dose transfer factor is used. This conservatism provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment.

For evaluating compliance with the dose limits of ODCM Specification 3.4.2, the following simplified equations may be used:

7 EO-08 xyQxNxff xQi D= 3 .10.50 (B.5)

Dp = 3 .*7 E -08xXQxNeffxZQi (B.6) 0.50 where:

Dy = air dose due to gamma emissions for the cumulative release of all noble gases (mrad)

Dpi = air dose due to beta emissions for the cumulative release of all noble gases (mrad)

VQ = atmospheric dispersion to the controlling site boundary (sec/M 3 )

B-3 REV. 9 12/02/2005

Meff = 5.3E+02, effective gamma-air dose factor (mrad/yr per tLCi/m 3)

Neff = 1.1E+03, effective beta-air dose factor (nirad/yr per pCi/m3)

I Qi = cumulative release for all noble gas radionuclides (pCi) 3.17E-08 = conversion factor (yr/sec)

-

0.50 = conservatism factor to account for the variability in the effluent data Combining the constants, the dose calculational equations simplify to:

D= 3.5E-O5xx/QXXQ (B.7) and Dœ= 7.OE-05xxIQxZQ (B.8)

The effective dose factors are used on a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods of computer malfunction where a detailed dose assessment may be unavailable. Dose assessments using the detailed, radionuclide dependent calculation are performed at least annually for preparation of the Radioactive Effluent Reports. Comparisons can be performed at this time to assure that the use of the effective dose factors does not substantially underestimate actual doses.

B4 REV. 9 12/02/2005

Table B-I Effective Dose Factors - Noble Gases Total Body Effective Skin Effective Dose Factor Dose Factor Kerf (L+1.1 M)efr Radionuclide fI (mrem/yr per jtCVm3) (mremlyr per ILCi/m 3 )

Noble Gases - Total Body and Skin Kr-85 0.01 1.4E+01 Kr-88 0.01 1.5E+02 1.9E+02 Xe-133m 0.01 2.5E+00 1.4E+01 Xe-133 0.9 3.OE+02 6.6E+02 Xe-135 0.02 3.6E+01 7.9E+01 TOTAL 4.8E1+02 9.6E+02 Noble Gases - Air Gamma Air Effective Beta Air Effective Dose Factor Dose Factor Mefr Neff Radionuclide fi (mrad/yr per tCVim 3 ) (mrad/yr per ipCi/m 3 )

Kr-85 0.01 -- 2.0E+01 Kr-88 0.01 1.5E+02 2.9E1+01 Xe-133m 0.01 3.3E+00 1.5E+01 Xe-133 0.95 3.4E+02 1.OE-+03 Xe-135 0.02 3.8E+01 4.9E+01 TOTAL 5.3E+02 1.IE+03 B-5 REV. 9 12/02/2005

APPENDIX C EVALUATION OF CONSERVATIVE, DEFAULT EFFECTIVE EC VALUE I FOR LIQUID EFFLUENTS

-

C-1 REV. 9 12/02/2005

Appendix C Evaluation of Conservative, Default Effective EC Value for Liquid Effluents In accordance with the requirements of ODCM Specification 3.1 the radioactive liquid effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive material at the discharge point does not exceed 10 times the value of 10 CFR 20, Appendix B, Table 2, Column 2 for all radionuclides other than noble gases and a value of 2X10 4 pCi/ml for noble gases. The determination of allowable radionuclide concentration and corresponding alarm setpoint is a function of the individual radionuclide distribution and corresponding EC values.

In order to limit the need for routinely having to reestablish the alarm setpoints as a function of changing radionuclide distributions, a default alarm setpoint can be established. This default setpoint can be conservatively based on an evaluation of the radionuclide distribution of the liquid effluents from Kewaunee and the ECe value for this distribution.

The effective EC value for a radionuclide distribution can be calculated by the equation:

EC,=

XCi (C.1)

Ci EC where:

ECe = an effective EC value for a mixture of radionuclide (MCi/mi)

C; = concentration of radionuclide "i" in the mixture EC, = the 10 CFR 20, Appendix B, Table 2, Column 2 EC value for radionuclide "i" (pCi/ml)

Based on the above equation and the radionuclide distribution in the effluents for past years from Kewaunee, an EC, value can be determined. Effluent release data from 2000-2002 was used to generate the results presented in Table C-1. The most limiting effective EC (for gamma emitting radionuclides) was for the calendar year 2001, with a calculated value of 5.98E-06 pCi/ml. For conservatism in establishing the alarm setpoints, a default effective EC value of 1.OE-06 pCi/ml was selected. The overall conservatism of this value is reaffirmed for future releases considering that 1.OE-06 pCi/ml is as or more restrictive than the individual EC values for the principal fission and activation products of Co-58, Co-60 and Cs-137. Overall, use of this effective EC value provides a factor of six (6) conservatism based on the 2000-2002 radionuclide distribution for gamma emitters.

C-2 REV. 9 12/02/2005

Being a non-gamma emitter, tritium is not detected by the effluent monitor. While tritium accounts for nearly all of the activity, it is not a significant contributor when determining the alarm setpoint for release rate evaluations. Examining releases over the years 2000-2002, the average, diluted H-3 contribution to its limiting concentration (i.e., fraction of concentration limit - 10 x EC) in liquid effluents was 0.004%. This contribution is not expected to change significantly over time, since the concentration of H-3 in effluents can be expected to remain fairly consistent in effluent releases regardless of fuel conditions, activation product releases, and waste processing.

Based on relative abundances, other non-gamma emitting radionuclides (Fe-55 and Sr-89/90) contributed up to 30% of the concentration limit (30% for CY 2001). It is reasonable to assume that the abundances of these non-gammas will remain the same relative to other fission and/or activation products under varying conditions. Therefore, under conditions of elevated effluent radionuclide levels, the gamma-emitting radionuclides can be expected to be the main contributors to limiting conditions on liquid effluent concentrations, as established in Technical Specification 6.16.b.1.B and ODCM Specifications 3.3.1. Note that including the non-gammas (excluding tritium) in the evaluation results in a higher effective EC value.

Therefore, under conditions of elevated effluent levels, the main contributor to the limiting conditions of the liquid effluent concentration would be the gamma-emitting radionuclides. The factor of six (6) conservatism in the effective EC determination (discussed above) provides adequate consideration for the contribution from non-gamma emitting radionuclides, and provides a conservative basis for establishing an alarm setpoint consistent with the requirements of Technical Specification 6.16.b.1.B and ODCM Specifications 3.3.1.

C-3 REV. 9 12/02/2005

I r r I I r r r r Ir r r r r r Table C-i Calculation of Effective EC (ECe)

_ 2000

• 2001

• 2002 Nuclide '(Ci/ml) Release Release Release I (C;) Cj/ECj Frac. (C,) Ci/ECi Frac. (CQ) Ci/EC . Frac.

Na-24 5.OOE-05 1.03E-03 2.06E+01 4.89E-03 2.18E-04 4.35E+00 ' 1.27E-03 O.OOE+00 O.OOE+0 O.OOE+00 Cr-51 5.OOE-04, 1.44E-03 2.89E+00 6.85E-04 8.26E-04 1.65E+00 4.83E-04 O.OOE+00 O.OOE+00 O.OOE+00 Mn-54 3.OOE-05 1.49E-04 4.97E+00 1.18E-03 3.30E-04 .IOE+-01 3.22E-03 6.41E-05 2.14E+00 9.83E-04 Fe-55 1.OOE-04 4.81E-02 4.81E+02 1.14E-01 4.85E-02 4.85E+02 1.42E-01 3.69E-02 3.69E+02 1.70E-01 Co-57 6.OOE-05 O.OOE+O0 O.OOE+00 O.OOE+O0 2.42E-05 4.03E-0I 1.18E-04 O.OOE+00 O.OOE+00 0.OOE+00 Co-58 2.OOE-05 8.07E-03 4.04E+02 9.59E-02 4.09E-03 2.05E+02 5.99E-02 4.94E-03 2.47E+02 ' 1.14E-01 Fe-59 1.OOE-05. 2.77E-04 2.77E+I01 6.57E-03 2.44E-04 2.44E+01 7.14E-03 1.65E-04 1.65E+01 '7.61E-03 Co-60 3.OOE-06 4.71E-03 1.57E+03 3.73E-01 4.31E-03 1.44E+03 4.21E-01 2.07E-03 6.89E+02 3.17E-01 Br-82 4.00E-05 4.94E-04 1.23E+01 2.93E-03, 1.44E-04 3.59E+00 1.05E-03 O.OOE+00 O.OOE+00: O.OOE+00 Sr-89 8.OOE-06 3.42E-04 4.27E+01 . 1.OIE-02 2.59E-04. 3.24E+01 9.48E-03 5.98E-04 7.48E+0 I 3.44E-02 Sr-90 5.OOE-07 2.25E-04 4.50E+02, 1.07E-01 2.50E-04 5.00E+02 1.46E-01 9.76E-05 1.95E+02 8.98E-02 Zr-95 2.OOE-05 1.16E-04 5.79E+00 1.38E-03 7.18E-05 3.59E+00 1.05E-03 5.24E-05 2.62E+00 1.20E-03 Nb-95 3.OOE-05 3.41E-04 1.14E+01 2.70E-03 2.39E-04 7.95E+00 2.33E-03 2.45E-04 8.17E+00 3.76E-03 Ag-lOim 6.00E-06 2.85E-03 4.74E+02 1.13E-01 1.63E-03 2.72E+02 7.97E-02 2.86E-03 '4.76E+02, 2.19E-01 Sn-113 3.OOE-05 9.65E-05 3.22E+00 7.64E-04 5.08E-05 1.69E+00 4.95E-04 7.06E-05 2.35E+00 1.08E-03 Sb-124 7.OOE-06 5.61E-04 8.01E+01 1.90E-02 1.81E-04 2.59E+O1 I 7.59E-03 4.34E-05 6.20E+00 2.85E-03 Sb-125 3.00E-05 4.86E-03 1.62E+02 3.85E-02 1.02E-03 '3.41E+01 9.99E-03 2.46E-03 8.18E+O I 3.76E-02 1-132 1.OOE-04 O.OOE+00 O.OOE+00 O.OOE+00 7.75E-08 7.75E-04 2.27E-07 O.OOE+O0 O.OOE+00 O.OOE+00 1-133 7.OOE-06 6.16E-04 8.80E+O1 2.09E-02 6.32E-04 9.03E+01 2.65E-02 O.OOE+0O O.OOE+0O O.OOE+O0 1-135 3.OOE-05 O.OOE+0O 0.OOE+00 O.OOE+-OO 4.61E-05 1.54E+00 4.50E-04 O.OOE+0O O.OOE+O0 0.OOE+O0 Cs-137 I.OOE-06 3.70E-04 3.70E+02 8.78E-02 2.74E-04 2.74E+02 8.02E-02 3.04E-06 ' 3.04E+00 1.40E-03

,' Total 7.46E-02 4.2 1E+03, 1.00E+00 6.34E-02 3.42E+03, l.OOE+00 5.06E-02 2.17E+03 L.OO1E+00 Non-Gamma Fraction 0.23 0.30 0.29 Gamma Fraction _ _ ' 0.77

• 0.70 0.71 ECe (pCi/ml, total) 1.77E-05 1.86E-05 2.33E-05 E C, (aCl/mi, Lammas) 8.03E-06 5.98E-06 8.44E-06 C-4 REV. 9 12/02/2005

APPENDIX D Site Maps D-1 REV. 9 12/02/2005

Appendix D Site Maps Plant drawing A408, "Radiological Survey Site Map" depicts the site area by illustrating the site boundary and the restricted areas. Plant drawing A-449, "Plan of Plant Area, Fence, lighting, and CCTV Support Structure" shows the layout of the site buildings. Members of the public are restricted from access to all areas of the Owner Controlled Area (OCA).

Figure D-1 presents the locations and elevations of radioactive effluent release points at the plant. The plant drawings referenced above are not included as part of the ODCM but can be found in the plant drawing system.

D-2 REV. 9 12/02/2005

KEWAUNEE COUNTY a -- 1-----

\ / A = CONTAINMENT BUILDINGl VENT ELEVATION 775' _______

B =AUXILARY BUILDING VENT ELEVATION 675' C =FLUENT LIOUID DISCHARGE l ELEVATION 580' D-3 REV. 9 12/02/2005 I I (((( ( l 1 ( (l(

APPENDIX E On-site Disposal of Low-Level Radioactively Contaminated Waste Streams E-1 REV. 9 12/02/2005

Appendix E consists of hard copies of the following reference documents:

DESCRIPTION DATE DOCKET NUMBER Operating License DPR-43 Kewaunee Nuclear Power Plant October 17, 1991 NRC-91-148 Disposal of Low Level Radioactive 50-305 Material Proposed Disposal of Low Level Radioactive Waste Sludge Onsite at the June 17, 1992 K92-119 I Kewaunee Nuclear Power Plant (TAC No. 50-305 M75047)

Safety Evaluation For An Amendment To An Approved 10 CFR 20.302 Application September 14, 1994 K-94-195 For The Kewaunee Nuclear Plant (TAC 50-305 No. M89719)

Alternate Disposal Of Contaminated Sewage Treatment Plant Sludge In November 13, 1995 K-95-172 Accordance With 10 CFR 20.2002 (TAC 50-305 No. M93844)

Onsite Disposal Of Contaminated Sludge K-97-64 Pursuant To 10 CFR 20.2002 (TAC No. April 9, 1997 50-305 M9741 1) _

I Adapted from N E-2 REV. 9 12/02/2005

WPSC 14 41433- 1538 MR 91-149 TELECOPIER (4'4) 43:,5544 EASYLAK 5z-;993 WISCOSiN PUBUC SIRVICE CoRPoaAMoN

-^ %rE ACar-S C So. '9002 Gret 6t.v.'. C2

-IM7 ;=2 bcc - K M Barlow. MrilE J NJMorison D2 A I Ruegc D2 N E Boys, %WPL J R Mueller D2 C A Schrock KNP larry Nilisen. ANFC D S Nalepka lCNP C S Smoker KNP D R Ber KNP L A Nuduis D2 (NSRAC) C R Steidardt D2 D A Bottom G6 R P Pulec D2 J I Wallace KNP R E Drahcim KNP I S Richmon D2 K H Weinhauer KNP K H Even D2 D I Ristu D2 S F Wczk D2

.M L Marchi KNP D I RocM KX QA Vault KNP D L Mank KNP T7X kvC4& tAY'P October 17, 1991 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 /*

Gendtemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Disposal of w Level Radioactive Material

References:

I) Letter frm; K.H.Evers to Document Control Desk dated September 12. 1989

2) Letter from M.I.Davis to K.H.Evers dated February 13, 1990

3) Letter from L.Sridharon (WDNR) to M.Vmndenbusch dated June 13, 1991 In reference 1, pursuant to the regulation of 10 CFR 20.302. Wisconsin Public Servicc Corporation (WPSC) requested authorization for the alternative disposal of very-low-level radioactive materials from the Kewaunee Nuclear Power Plant. In reference 2. the US NRC identified additional questions that needed to be addressed in order to complete their review.

Auachment I provides our response to the questions.

WPSC requested the State of Wisconsin Department of Natural Resources (WDNR) Lo revtew the disposal options for the service water pretreatment lagoon sludges. In reference 3. the WDNR completed a review of the most appropriate on site disposal methods for the slightly contaminated service water pretreatment lagoon sludges. The two proposed methods that the WDNR evaluated included in-situ capping of the sludge in the wastewater treatment lagoon and on site landsprcading. In Attachment 1, Appendix A, WPSC evaluated the on site landspreading E-3 REV. 9 12/02/2005

Document Control Desk October 17, 1991 Page 2 application which is our preferred disposal method. WPSC does not intend to utilize the in-situ capping of the sludge in the lagoon at this time. However, in the letter the WDNR agreed that either disposal method was acceptable provided:

if the material is to be left in the lagoon, it would be capped in accordance with Wisconsin State statutes.

- if the on site landspreading option is utilized, the material would be spread by either disking into the soil or by spiking into the ground.

WPSC will abide by the WDNR landspreading requirements which include locational and performance standards. Should there be any additional questions please feel free to contact a member of my staff.

Sincerely, --

cla .A A4.

C. A. Schrock Manager - Nuclear Engineering DJM/jms Attach.

cc- US NRC- Rcgion II Mr. Patrick Castleman, US NRC LlCDJIM\ N492 Em4 REV. 9 12/02/2005

-

ATTACHMENT 1 To Desk (NRC)

Letter from K. H. Evers (WPSC) to Document Control Dated October 17, 1991 E-5 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page I References 1) Letter from K. H. Evers to Document Control Desk dated September 1, 1989.

NRC Question #1 On page 4 of your submittal, the average input to the Sewage Treatment System is approximately 11,000 gallons per day. In the Final Environmental Statement, this system is to be operated below its design capacity of 9,000 gallons per day. Discuss this deviation from the design capacity, and provide information to justify the higher output for this system.

WPSC Response -

The original Sewage Treatment System installed at the Kewaunee Nuclear Power Plant (KNPP) was replaced in 1986 with a higher capacity system. The original system was designed for an onsite work force of around 150 people.

It was a limited capacity aerobic treatment system which included the onsite lagoon for additional retention. Because of his limited capacity and more stringent conditions on system effluent to Lake Michigan, an aerobic digester system was installed, which has a higher capacity, and uses current technology.

The estimated input volume to the Sewage Treatment System used in the September 12, 1989 application was 11,000 gallons per day. This value was based on past operating data. The increase in influent from the original design basis included in the Final Environmcntal Statement is due mainly to an increase in (he ntzmber of individualt and facilities (e.g., training and simulator building) located onsite. Design changes to the system were required to accommodate these new facilities.

E-6 REV. 9 12/02/2005

km Document Control Desk October 17, 1991 Attachment 1, Page 2 The current volumes of sewage sludge were used as the basis for the potential dose analysis and corresponding radionuclide concentration limits. This increase has no significant effect on the dose modeling. (Refer to the response Io NRC Question #2, below.)

NRC Question #2 Provide information regarding how the disposal plan assures that the annual dose to any exposed individual will be kept below 1 mrem per year.

WPSC Response The dose pathway modeling used for determining the radioactive material concentration limits was based on NRC modeling. The computer code IMPACTS-BRC was used as the basis for calculating the potential doses from the alternative disposal methods. This modeling includes reasonable conservative exposure pathway scenarios for the various disposal methods.

Administrative controls will be established to ensure that the actual disposal of any slightly contaminated materials from KNPP are within the bounds of the evaluation. Samples from each of the waste streams will be collected and analyzed by gamma spectroscopy prior to release fur disposal. A system lower limit of detection (LLD) of SE-07 pCi/ml for the principal gamma emitting radionuclides will be required. This LLD ensures the identification of any contaminated materials at a fraction of the allowable concentration limits for the alternative disposal.

The results of these analyses will be used to ensure that any detectable levels of radioactive material are within the limits for alternative disposal. Any materials with levels of radioactive material above the concentration limits E-7 REV. 9 12/02/2005

Document Control Desk Octobe 17, 1991 , Page 3 (and of plant origin) will be treated as a radioactive waste and appropriately controlled.

Records will be maintained to ensure that the cumulative disposal of any contaminated materials ae maintained within the bounds of the evaluation. In addition to a comparison of the individual radionuclide concentration limits, a record of the total amount of radioactive material disposed of will be maintained. Cumulative totals will be maintained to ensure that the total activity does not exceed the quantity assumed in the derivation of the limits.

In developing the concentration limits presented in Table I of reference I, it was assumed the total annual design basis volume of 27,000 ft 3 would be contaminated at the derived limit. Tbe dose commitmnent from each radionuclide was individually evaluated as if it were the only radioactive material present. To determine if a mixture of radionuclides meets the limit, the sum-of-the-fractions rule should be applied (i.e., the sum of each radionuclide's concentration divided by its limiting concentration must be less than one).

The concentration limits of Table I of reference 1 also have an implied total activity limit. This limit is determined by multiplying the individual radionuclide concentration limit by the total estimated waste volume of 27,000 ft 3 . These total activity limits are presented in Table A of this response, for each radionuclide individually. For a mixture of radionuclides, a total annual activity limit may be determined by normalizing the concentrations so that the sum-of-thefractions for the mixture equals one (1). These resultant adjusted concentrations may be multiplied by the 27,000 ft3 waste volume to determine the corresponding total activity limit of the mixture.

E-8 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 4 A Disposal Log will be maintained on a calendar year basis for all disposals of any very-low-level radioactive materials. The log will contain as a minimum the following information:

• Disposal location

• Description of waste

• ShipmentLdisposal date

. Waste volume

• Radionuclide concentrations (gamma emitters)

• Year-to-date radionuclide activity.

Year-to-date waste volume In addition to the above Disposal Log, a record file will be kept for cach individual disposal. This file will contain, as a minimum, the following information:

• Waste identification

• Sample gamma spectroscopy results

• Identified radionuclide concentrations and total activity NRC Question #3 Revise Appendix B, Section A of your submittal, Rdt Exposure During Transport,w by adding the cumulative dose to the exposed population per reactor year for both the transportation worker and the general public (onlookers along route).

WPSC Response The potential exposure to the general public (onlookers along route) is modeled by the IMPACTS-BRC code. As addressed in NUREGICR-3585, this modeling is based on an integration of the source strength, an assumed E-9 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 5 population density along route and vehicular speed. For a conservative valuation of the potential exposure to the general public from the transport of 2

the KNPP waste, a population density of 610 persons/mi was assumed. This value is conservative for the KNPP site area where the average population density is less than 53 persons/mi2 . A ransport distance of 45 miles was assumed. The IMPACTS-BRC modeling assumes five (5) tons of material are transported per shipment. For the assumed KNPP waste volume, this shipment weight tnslates into a total of 167 shipments per year. With a vehicular speed of 20 miles per hour, the resultant total population exposure time is 375 person-hours per year. At the concentration limits established for the alternative disposal, the potential onlooker doses during transport will be less than 0.01 person-rem-per year. For the modeling of the exposure to the transport worker, the IMPACTS-BRC model assumes two drivers per vehicle.

As presented in the September 12, 1989 submittal, the maximum dose to the driver is less than 1 mrem per year (<0.001 remlyr). Therefore, the total collective dose to the transport workers will be twice the individual dose, i.e.,

Icss than 0.002 person-rem. Including the population dose of <0.01 person-rem per year, the total collective dose to both the transport workers and the population is less than 0.02 person-rem (0.002 person-rem + 0.01 person-rem

< 0.02 person-rem).

For the disposal of the existing 15,000 ft 3 of contaminated sludges, the population dose due to the transportation of the waste is calculated to be 0.0002 person-rem. The estimated collective exposure to the transport worker is 0.00007 person-rem. The total collective dose due to transport of the waste is 0.00027 person-rem.

E-10 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 6 Additional Potential Disposal Method The Wisconsin Department of Natural Resources has requested Wisconsin Public Service to examine the feasibility of land application of the lagoon sludges in lieu of disposal in the Kewaunee County Landfill. Land application is also an option for the disposal of the sewage sludges. Therefore, WPS requests that the option for onsite disposal at the KNPP site by land application be included in the alternative disposal methods which was determined to be acceptable in our September 12, 1989 submittal.

The potential pathways of exposure as evaluated in the September 12, 1989 submittal conservatively bound any additional pathways of exposure that would result from onsite land spreading of the waste. Attachment A to this response provides an overview of the land spreading disposal method. Also, the pathways of exposure applicable to the onsite land application are evaluated; and a comparison to the controlling pathways and radionuclide concentrations as presented in the September 12, 1989 submittal are discussed. From a modeling standpoint, the two exposure scenarios, 'Radiation Exposure During Transportw and "Radiation Exposure to Landfill Operator," appropriately characterize any potential exposure to workers involved with the land spreading of the waste. The other post-disposal exposure scenaros, lIntruder Scenario", 'Intruder Well", and 'Exposed Waste Scenario," as described in NUREGICR-3585 (and as discussed in Appendix C of the submittal) reasonably bound any potential exposures from either ground waste migration or post-release from the Kewaunee site. In no case is there a higher potential for exposure from land application than the pathways and potential exposures that were used for the derivation of the limits for alternative disposal.

Therefore, no revisions are needed to the radionuclide concentration limits proposed in the September 12, 1989 submittal to include the option for disposal by onsite land spreading of the waste.

E-11 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 7 Table A

- Radionuclide Quantity Limits for Alternative Disposal Limiting Limiting Annual Nuclide Concentration Quantity (U01Mi~) (Ci)

H-3 9.65E-04 0.7382 C-14 4.55E-05 0.0348 Cr-51 3.13E-04 0.2394 Mn-54 1.14E-05 0.0087 Fe-55 1.OOE-02 7.6500 Fi-59 7.90E-06 0.0060 Co-58 1.166E-05 0.0089 Co-60 3.74E-06 0.0029 Ni-63 I .OOE-02 7.6500 Sr-90 3.45E-03 2.6393 Zr-95 6.28E-06 0.0048 Nb-95 1.23E-05 0.0094 Mo-99 6.73E-05 0.0515 Tc-99 2.70E-04. 0.2066 1-129 2.50E-06 0.0019 1-131 2.68E-05 0.0205 Cs- 134 6.16E 06 0.0047 Cs-137 I 1E-5 0.0131 Ba-140 5.52E-05 0.0422 L~a-140_ 4.17E6 0.0032 Transuranics

.TRU (T'a > 5 yrs) 8.91E-05 0.0682 Pu-241 2.85E-03 2.1803 Cm-242 1.OOE-02 7.6500 Assumes annual quantity of KNPP Wastes is 27,000 f 3 or 7.65E8 mls.

E-12 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 8 Appendix A Evaluation of Onsite Land Application for Alternative Disposal of Very-Low-level Contaminated Materials O verview LAnd spreading of lagoon sludges onsite at the Kewaunee Nuclear Power Plant has been as recommended by personnel from the Wisconsin Department of Natural Resources (DNR) a desirable alternative to the use of the Kewaunee County Landfill for disposal. This method of disposal is also a recommended practice for disposing of sewage treatment facility sludges.

for Therefore, WPS requests that this disposal method be included in the options available the alternative disposal of very-low-level radioactively contaminated materials from KNPP.

Description of Disposal Method The disposal of KNPP sludges will be performed by beneficial land application to a dedicated disposal area located onsite at the Kewaunee Nuclear Power Plant. Typical methods of land spreading will be employed. KNPP sludges will be loaded onto appropriate vehicles (e.g.,

tanker truck, sludge spreader, etc.) and applied to the dedicated disposal area. The dedicated disposal area will be periodically plowed to a depth of 6 inches.

Onsite disposal of water treatment and sewage sludges are allowed by EPA and State of Wisconsin Dcpartment of Natural Resources with the criteria and limits for land spreading being specified by the potential use of the land. The two land use criteria are 1) Agricultural land that covers any lands upon which food crops are grown or animals are grazed for human consumption, and 2) Non-Agricultural land that covers lands which do not represent ingestion pathways to inan. To be conservative, the Agricultural Land Application limits of sludge contaminants will be applied to the KNPP wastes even though the less restrictive Non-Agricultural Land Application sludge contamination limits are allowed. Therefore, no more than 50 metric tons of sludge per hectare will be applied to the dedicated disposal site. This limit will ensure that any land application will not exceed the bounds of the dose analysis as E-13 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 9 performed previously. In addition, other limitations as applied to land application by the State of Wisconsin Department of Natural Resources will be followed (e.g., control of runoff/erosion, proximity to wells/residences/surface water, etc.).

Applicable Pathways of Exposure The pathways of exposure applicable for land spreading are not appreciably different from the pathways evaluated for the disposal methods at the Kewaunee County Landfill or the Green Bay Metropolitan Sewerage District facilities. The major exposure pathways are discussed below:

Direct Exposure to Works Any potential exposures to workers involved in the removal, transport and land spreading of the sludges are reasonably bound by the evaluation of the exposure to the transport worker in the September 12, 1989 submittal. The transport worker has been assumed to be exposed for 460 hours0.00532 days <br />0.128 hours <br />7.60582e-4 weeks <br />1.7503e-4 months <br /> per year at one (1) meter from unshielded waste. For the land spreading of these wastes, it is estimated that the total exposure time for the removal and disposal of the lagoon sludges will require no longer than a three week period per year (i.e.,

120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />).

The potential exposure to a worker onsite after land spreading, has been estimated at no more that 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year. Such an individual would be involved in land maintenance activities, such as plowing and mowing. As modeled in the September 12, 1989 submittal, an exposure of 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year to the landfill operator has been assumed. For this exposure, the KNPP materials are mixed with other landfill waste: a 1:13 mixing of KNPP materials to other waste is assumed. This mixing is not significantly different from the type of mixing that will occur in the field with the sludges being E-14 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page 10 plowed into the soil to a depth of six (6) inches. With a land spreading of 50 metric tons per hectare per year, a mixing ratio of 1:30 will be achieved.

Therefore, the resultant dose to the exposed worker would be less than the 1 mrem per year dose to the transport worker as evaluated in the September 12, 1989 submittal.

Post Disposal Exposure - utruder Scenario The IMPACTS-BRC model, as applied to the disposal of the KNPP waste, assumes a loss of institutional controls 10 years after closure of the site (See Appendix B of the September 12, 1989 submittal). An individual is assumed to reside in a house built on the disposal area. This individual receives a direct exposure (from the uncovered waste), an inhaation exposure (from resuspension), and an ingestion exposure (from growing ' of his food crops). For modeling purposes, it is assumed that the waste is mixed at a ratio of 1:13 with other soils during the resident's construction process.

The onsite land application of KNPP waste will be limited by the Agricultural Land Application sludge concentrations evet dtugh the less restrictive Non-Agricultural Land Application sludge concentrations are applicable since a "dedicated land disposal silt will be used (i.c., no crops will be grown on the disposal site).

Therefore, provided the KNPP waste does not exceed the Non-Agricultural maximum sludge concentrations for heavy metal or organic chemicals, unlimited application of waste to the dedicated land disposal site is allowed. However, to be conservative the land application of KNPP wastes will be limited to 5 metric tons per hectare per year.

The intruder scenario as evaluated in the September 12, 1989 submittal conservatively bounds this exposure pathway for the on-site land spreading.

E-15 REV. 9 12/02/2005

Document Control Desk October 17, 1991 , Page II Post Disposal - Intruder Well The intruder well pathway for onsite land disposal is essentially the same as the intruder well pathway as evaluated by the IMPACTS-BRC model. It is conservatively assumed that the well is located at the edge of the disposal site. As modeled, locating the well at the disposal site edge in downstream flow' direction maximizes the calculated hypothetical dose. (Additional discussion of this modeling is presented in NUREGICR-3585, Volume 2).

The potential dose for the intruder well scenario for the land spreading disposal would be less than 0.001 mrem per year. The modeling as presented in the September 12, 1989 submittal reasonably bounds any hypothetical well water exposure pathway.

In summary, the modeling of the exposure scenarios, as presented in the September 12, 1989 submittal, conservatively bounds the hypothetically exposures for the on-site land spreading. In no case is it likely that any individual, either on-site or off-site, will receive a dose in excess of 1 mrem per year from the disposal of the slightly contaminated materials.

E-16 REV. 9 12/02/2005

UNITED STATES NUCLEAR REGULATORY COMMISSION C-1,

.I X ; WASHINGTON. 0 C 27M June 17, 1992 Docket No. 50-305 Mr. C. A. Schrock Manager - Nuclear Engineering Wisconsin Public Service Corporation P. 0. Box 19002 Green Bay, Wisconsin 54037-9002

SUBJECT:

PROPOSED DISPOSAL OF LOW LEVEL RADIOACTIVE WASTE SLUDGE ONSITE AT THE KEWAUNEE NUCLEAR POWER PLANT (TAC NO. M75047)

By letters dated September 12, 1989, and October 17, 1991, you submitted a request pursuant to 10 CFR 20.302 for the disposal of waste sludge onsite at the Kewaunee Nuclear Power Plant. We have completed our review of the request and find your procedures, including documented commitments, to be acceptable.

This approval is granted provided that the enclosed safety evaluation is permanently incorporated into your Offsite Dose Calculation Manual (00CM) as an Appendix, and that future modifications of these commitments are reported to the NRC.

Issuance of this safety evaluation completes all effort on TAC No. M75047.

Sincerely, Allen G. Hansen, Project Manager Project Directorate 111-3 Division of Reactor Projects 1II/IV/V Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/enclosure:

See next page NRC LrrER DISTRIBUSON T A HAAWD (MG&E) J P Gieder D2 C A Sdcrock D2 J D Loock (WFL) M L Muchi KNr C R Sweinhw D2 Lurry Nielsen IANtFC) D L Misarik KJNP T J Webb 1CNP

L Bdant (NSRAC) R P Pulec D2 (2) S F Wouniak DZ U A 5.,flom G6 D J Ristau D2 QA Vault KNP K 1i Evers KNP A J Ruegc D2 E-17 REV. 9 12/02/2005

Wiscor~sin Public Service Corpcration Kewaunee luclear Power Plant cc:

David Baker. Esquire Fcley and Ladrdn.r P.O. Box 2i93 OrlanKo, Florida 32382 Glen Kunesh, Chaiirrn Town of Carlnon Route I Kewaunee, Wisconsin 54215 Mr. larold Reckelberg, Chalrnam Kevaunea County Board Kewaunee County tourthouse Kewaunee, Wisconsin 54216 Cha4rman Public Service Connissi-n of isconsin Hill Farr.s State Office Bjilding Madison, 'isconsin S3732 Attorrney General 114 East, State Capitol Madison, Wisconsin 537C2 U.S. Nucloar Regulatory Conaission Resident Inspact3rs Office Route I1, Box 99)

Kowa-.nfn, Wiscunsin 54216 Regional Administrator - Reviuor III U.S. Ntclear Regulatory Comissio'n 7s9 RHicsevelt Road Glen Ellyn, Illincis 60137 Mr. Robert S. Cullen Chief Engireer Wisccnsin Ptblic Service Coruissicr P.O. Box 7554 Madison, Wisccnsin 53707 E-18 REV. 9 12/02/2005

LU, 0P UNITED STALTES LLjr r INUCLEAR REGULATORY COMMISSION S2 f WASSaCT!. D.C.

70.0 Fly EV--UK'0L !IYE.PFM-r "A5 OF r mxCd.%JR FPor:TI)?L-:ULATlo;!

RFIAT!hiS T-- PAM!ACTI'4EL' 1FDI4HtLOFLWL~

A!;TWL Rfl~'IJ T 5ilE DISM)AL 0E-WL-l6,E'.

LQI4TMIINATFD W1A1SE1L1Dr AT TPF KEWA4UNEE MUMLFAR PaFR PLANT 1'ISCONSIN PUBLiC 5FVi~CE CCRPORATION WISCOEiSIN POMER AND LIGHT COMPANY

?IADISON GAS AND ELECTRIClPMANf MOCKET NO. 50---C'

,3 1iTR01LCTION In reference 1, Wisconsin Public Service Corporation (dPSC) requested apprcval pursuant to Section 20.302 of Title 10 of the Code cf Federal Regulations (CFR) for the disposal of licensed material not previously consicered in tIe Kewaunee Fir3al Environmental Statement (FES) dated D)cernber 1?2. Addittcr.al related material frcm the licensee, froi the State of Wisconsin, usd frcn thQ

taft are cor-airetd n references 2 tlhrough 5.

The WPSC rpques! can.talns a detailed description ot- the licensed r.aterial i.c., contazm;rated sludge) subject to this 10 CFR 2C.3C2 reques*, based cn radioactivity absorbed frcm liquid discharges of licersed material. The l.C00 cubic feet :f 4coriaminated sludge identif rd ir tIe request contairs a

ctal radionujlide invertory of 0.17 rCi cF Cesijn-137 ard Cchalt 63.

irn its sibr.i-tal, the licensee addressed spectric -nformaticn reques-ed in accordar.ce with la CFq 2O.3C2(a), prov1dec a detailed description of the

!fcensed raterial, thoroughly aral!'zi~d anc evaluated the friftriation port;nent to Ine effects an the environimenrt nf the proposed diipasal of licensed naterial. and cu.riltted to follow specific procedures to minimize the risk of unexpected exposures.

2.3 :ESCRIPliol CF WAN!E During the normal operation of Kewaunee, the pctertiai exists Far ir-plant process streams which are not r.orMally radicact;*e to becou= ccrtaminated with very lcw levels of radioactive materials. Ttese Pvaste streaa.s are norily sqFarated from the radioactive streams. tcwever, due ma'nly to inFrequent, mlncr sviten leaks, ard anticipated cperationa' occucrrences, the potential exists for these systems to becore sl'ghtly cuntatir.ated. At Kewaunee, the setcndary SySt*.. damleralizer resins, the srviie wa:er pre-treatrent system sludges. the make-up water syster resins, an trne sewage treatrent plant sludges are waste streans that have the potetitfal to beccne contarinated at very 'cw leve's.

E-19 REV. 9 12/02/2005

During the yearly testirg of a batch of pre-trea:nent sluCge, it was found that approximately 15,000 cubic feet of sludge had oeen ccntavinated with Cs-137 and Co-S0.

3.0 PROPOSf1 DISPOSAL METH00 WPSC plar.s to dispose of the 15,OCO c;.bic feet of contaminated sludge onsite pursuant to 10 CFR 20.302. The sludge is currently contained In an cnsite lagoon at the KNPP sewage treatment facility. Tne disposal of the sludge s%1.

• be by land application to an area located onstte at KNPP. as shown in Figure 1. .;ne area will be periodically.plawod to a depth cf 6 inches.

Table I lists the principal rouclides identified In the sludge. The activity is based on reasurerents made in 1939. The radionucide hal f-lives, which are dominated by 30-year Cs-137. meet the staff's iO CF.2 ZC.3GZ guidelines (reference 6). whic apply to radionuciles with half-lives less than 35 years.

4ucllde Total Activity lnCil Co-60 0.076 Cs-l3' 0.094 0.170 4.0 RAD1OLO3t:AI IMPACTS The licensee has evaluated the following pctenrial exposure pathways to members of it'e general public fron tIe radionuclides Inthe sludge: (1) external exposLre caused by groundshir.e frcm the disposal site; (2)in:ernal exposure from irnalation of re-suspended radionucliees; and (3) internal exposure from ingesting ground water. The staff has revieW2d the licensee's calculatior.al methods and assumptions and finds that they are consistent %ith NRC Regulatory Guide 1.109, Calculation of Annual Dcses to Man from Rou:ire Releases of Reactor Effluents for the Purpose of Evaluating Compliance witl. 10 CFR Part 50, Appendix 1I,Revision 1,Octoter 1977. The staff finds t!.e assessment methodology acceptable.

Table 2 lists tha deses calculated by the licensee or the raxizally exposed member of the public based on a total activity nf C.170 TCi disposed of in the current year, as well as the c=:ulative irpact of similar disposals during subsequent years. For any repetitive disposals, the licensee rust reapply 'o the XRC %hen a particular disposal would exceed the following houndary conditions! (1)tha annual disposal rust ba less than a total activity of 0.2 mCi; (2)the whole tcdy dose to the hypothetical %axinally exposed irdividual must be less than 0.1 nrem/year; and (3)the disposal rust be dt the save site as descrited in Figure 1.

E-20 REV. 9 12/02/2005

'Ohole Body DWAe RWoitd by Maximially Exposed 'ndividual Pathway frrem/l'earl Grcundshtne 0.034 Inhalation 0.008 Groundwater Ir.gestton 0.07 IOThL O

As shown in Table 2, the annual dose is expected to be on the order of 0.1 mren or less. Such a dost is a small fraction of the 3DO trem received dnnuaily by menbers of the general public fron sources of natural bacegroand raciatton.

lie guidelines used by the NR' staff fcr onsite disposal of '.icer-sed material are presented inTable 3, along with the staff's evaluatton of how eac.h qj4.deline has been satisfied.

The licensee's procedures and coi.itrents as docurented In the subrmttal are acceptable, provided that they are permanently incorporated into the licensee's Offsite Csuse Calculation Manial (ODCM) as an Appendix, and that ruture moofticatimns b repcrted ta NRC ir. 3ccordance with the applhcatle OCCM change protocol.

Based un tne above findings, the staff finds the licensee's proposal :-

dispose of the lcw level radicactive waste sludge onsite in the ranner Apscribed in the UPSC letter dated September 12, 1989, to be acceptable. Tne State cf Wiscons n has also approved these procedures (refererce 5).

E-21 REV. 9 12/02/2005

fABLE 3 20.302 Guideline for Onsite Disposal Staff's Eva.9atW.g I. Thte radioactive material should 1. Due to the nat.re of the be disposed of in a manner that it disposed material, recycling to the Is unlikely tnat the material would general public is not considered be recycled. likely.

2. Doses to the total body and any 2. nis guideline 's dOdressed in body organ of a ;maximal y exposed Table 2.

Indlildual (a iuaber of the general public or a non-occupationally exposed worker) from the probable pathways of exposure to the disposed material should be less than I nrem/year.

3. Doses to the total body and any 3. Because the material will be body organ of an inadvertent land-spread, the staff considers the intruder from the probable pathways maximally exposed individual of exposure t1ould be less than scenario to also address the S uremiyear. intruder scenario.

4. Doses to :ne total body and try 4. Even it recycling were to cecur body organ of an individual fror after release from regulatory assumed recycling of the disposed control. the dose to the maximally material at the tire the disposal exposed iewber of the public is not site is released from regulatory expected to exceed I rram/yAar, control from all likely pathways of based on the exposure scenarins expcsire should be less than I rrem. considered tn this analysis.

E-22 REV. 9 12/02/2005

p.FEENC ES (1) 'PSC letter from K :1. tvers to hJZC 3ocument Ccntrol Osk, September 12, Iseg.

(2) Memorandum froma L. J. Cunnir.ghaqi, CREP, to J. 1. Harnar, 'Request For Additicnal Information," Decerber 11, 1989.

(3) NRC letter frcom M9.J. Davis to K. H. Lvers cft'PSC dated February 13, 1990.

(4) IWPSC letter !ror. K. H. Evers to NRC Document Control Dask, Detoter 17, (5) Letter frcn L. Srtdharon of the State of Wisconsin Departinent of Natiral Resources to '..Vandenbisch of WPSC, dated June 13, 1991.

(5) E. F. Branagan Jr. and F. J. rongel. Disposal of Contaminated Radioactive Wastes from Nuclear Power Plants,' presented at the Health PhyrSics Society's ir.dyear Syrposiu: on Health Physics Considerations in Decontaiination;0eownissioning, Kncxville, TN, February 1996 (CCNF-E60203).

?rincipal Contritutor; '. Yinns Date: June 1, l SS2 E-23 REV. 9 12/02/2005

- c -

Figure 1 Pacer Prant Site Area Map Kewaunee %lc'ear E-24 REV. 9 12/02/2005

'IL K-94. k95

- UNiED STATES -/21I94 NUCLEAR REGULATORY COMMISSION WASHINGTON, DVX. 2M--1 September 14, 1994 Mr. C. A. Schrock Manager - Nuclear Engineering llisconsin Public Service Corporation Post Office Box 19002 Green By, VI 54307-9002

SUBJECT:

SAFETY EVALUATION FOR AN AMENUDMENT Io AN APPROVWD 10 CFR 20.302 APPLICATION FOR THE KWAJUNEE NUCLEAR PLANT (TAC NO. M89719)

Cear Mr. Schrock:

By letter dated June 23, 1994, as supplemented June 29, 1994, you requested approval to use another onsite area for the disposal of contaminated waste sludge in addition to the location approved by the NRC on June 17, 1992. The staff has completed its review of your request and finds that your proposal meets the radiological boundary conditions approved in the June 17. 1992, Safety Evaluation, and is therefore acceptable. The staff also finds that your proposal is in accordance with 10 CFR 20.2002 which replaced 20.302 on January 1, 1994.

This approval Isgranted provided that the enclosed Safety Evaluation Is permanently incorporated into your Offsite Dose Calculation Manual (00CM) as an Appendix, and that future modifications of these comitments are reported to the NRC.

Sincerely.

r4" J6 Richard J. Lalfer, Acting Project Manager Project Directorate I11-3 Division of Reactor Projects III/JY Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosure:

Safety Evaluation cc w/enclosure:

see next FAgo T A Ia~am Ka(4u&*4 K A a" no C 5S okei KNP X W Seitz(WPL) ML M ' CDR a tD2 1z NYdaE (ANPC D L Mum DW C A Seafty V:P DA Boa=6 JNmmTimDi TTW1b W D EOCob L ANU (NSRAC) S F Wos 2 K Eves KNp R r Pu 2(2) QAVansI XD I PGielrKWP C A dual E2 E-25 REV. 9 12/0212005

Wisconsin Public Service Corporation Kewtaunee Nuclear Power Plant cc:

Foley & Lardner Attention: Mr. Bradley 0. Jackson One South Pinckney Street P. 0. Box 1497 Madison, Wisconsin 53701-1497 Chairman Town of Carlton Route 1 Kewaunee, Wisconsin 54215 Mr. Harold Reckelberg, Chairnan Kewaunee County Board Kewaunee County Courthouse Kewaunee, Wisconsin 54216 Chairman Public Service Cocmission of Wisconsin Hill Farts State Office Buildfng IQ11son, Wisconsin 537OZ Attorney General 114 East. State Capitol Madison, Wisconsin 53702 U. S. Nuclear Regulatory Comission Resident Inspectors Office Route I1, Box 999 Kewauiee, WisconsiA 54216 Regional Administrator - Region III U. S. Nuclear Regulatory Commission 801 Warrenville Road Lisle. Illinois 60532-4531 Mr. Robert S. Cullen Chief Engineer Wisconsin Public Service Conaission P. 0. Box 7854 Madison, Wisconsin 53707 E-26 REV. 9 12/02/2005

UNITED STATES m)7 ra NUCLEAR REGULATORY COMMISSION b WASUINTON. 9.C. =O6-M SAFETY EVALUATION BY THE OFFICE OF WJCLFAR RFACTOR REGUIA.T ON RELATINO TO ONSITE DISPOSAL OF LOW-LEVEL RADIOACTIVELY COHNIANATED WASTE SLUDGE AT THE KEWAUNEE NUCLEAR POWLR iLAM WISCONSIN PUBLIC SERVICE CORPOPATION VISCONS INPWER AND LIGHT COMPANX MAI~SO N ELECTR CIMPANY AY3 DOCKET NO, 50-305 1.0 INIRODUCTION By letter dated June 23, 1994, and as supplemented an June 29. 1994, Wisconsin Public Service Corporation (the licensee) requested approval to use another onsite area for the disposal of contaminated waste sludge in addition to the location approved by the NRC on June 17, 1992.

2.0 rVAtUATtIH A Safety Evaluation (SE) dated June 1?. 1992. approved the I censee's request pursuant to 10 CFR 20.302 for the disposal of 15,000 cubic feet of contaminated waste sludge by land application at the Kewaunee Nuclear Power Plant (KNPP) at a specific cnsite locatton. The SE imposed the following boundary conditions:

I The annual disposal must be less than a total activity of 0.2 mCi.

2. The whole body dose to the hypothetical maximally exposed individual must be less than 0.1 inrem/year.

3. The disposal must be the same site.

The site designated In the SE was an unused area adjacent to the onsite lagoon at the K1PP sewage treatment facility. In 1993, approximately 7500 cubic feet of the original 15,000 cubic feet of contaminated sludge was spread on that location. The licensee his now proposed to dispose of the remaining contaminated sludge at another onsite location northwest of the plant (see Attachmatit). The licensee hit cCsittid that the new ditpesfl. location will meet all the radiological boundary conditions contained in the SE for the 10 CFR 20.302 application approved on June 17, 1992. Additionally, the licensee has stated that this-additional disposal site will Feet all applicable Wisconsin Department of Natural Resources (WDNR) application requirements (i.e.. sludge application rate and frequency of spreading rate),

in addition to WONR landspreadiag requirements regarding locatien and perfomance stindards that were required at the original disposal site.

E-27 REV. 9 12/02/2005

3.0 CONCLUS 10 The staff finds the licensee's proposal to dispose of the low-level radioactive waste sludge in the additional oassto location to be within the radiological boundary conditions approved Inthe June 17, 1992, SE and Is therefore acceptable. The staff also finds that your proposal is in accordance with 10 CFR 20.2002 which replaced 20.302 on January 1, 1994.

As stated In the NRC's June 17, 1992, approval of the licensee's 10 CFR 20.302 application, the licenses is required to peruanently incorporate this modification Into the Offsito Vase Calculation Manual as an Appendix, and that future modification of this commitment be reported to the NRC.

Principal Contributor: S. Klementawicz Date: September 14, 1994

Attachment:

KNPP Site Area Map E-28 REV. 9 12/02/2005

r r r C rr r r~ r r r [ r C r r r r KEWAU~ NULEARPoWea PL4ANT SITC AR1EA MA C-f

be E-29 REV.A9 12/02/2005

UNITED STATES NUCLEAR REGULATORY COMMISSION

• wASHINGTOM. D.C. 2m-opsg

• -

• t C'eiba-@ 13, 1995 Mr. M. L. Marchi Manager - Nuclear Busu mss Group Wisconsin Pjblic Service Corporation Post Office Box 19002 Green Bay, VI 54307-9002

SUBJECT:

ALTERNATE DISPOSAl Of COMTANIXATED SEWAGE TREATHENT PLANT SLONE IN ACCORDANCE VMs 10 C$R 20.2002 (TAC N0. M93844)

Dear Mr. Harchi:

By latter dated October 17, 1995, as supplemented on Kovember 3, 1995, you requested approval for the onsite disposal of contaminated sewage treatment sludge In accordance with 10 CIA 20.2002. This request wai similar to a previous dispo3s1 request that was approved by the NRC cn June 17, 1992.

The staff has capleted Its review of your request and finds that your proposal meets the radiological bowudary conditions approved In the June 17, 1992, Safety Evaluation, and Is therefore acceptable.

This approval Is granted provided that the enclosed safety evaluation is permanently incorporated Into you OffsMto Dose Calculation Manual (OlMii) as an Appendix. and that futur modifications of these comitments are reported to the NRC.

Sincerely, Richard J. Laufer, Prjt Manager Project Directorato 111-3 Division of Reactor Projects II /I1 Office of Nuclear Reactor Regulation Docket No. 5D-305

Enclosure:

Safety Evaluation cc: See next page c, to WPlC l DrIfl'fLyn A

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N W SUitz (WPL) M1L Marc DQ C A St!.nhar* D2 1i Niet= (ANt) I XC Jabla NSRAC) CA Stezky XNP(Lic)

D A Bo1omC G6 . RPPuloc (3) 5 F Womia D2 n E Da Di C A SckrockKNP IJ Douick XCP(m E-30 REV. 9 12/02/2005

Mr. N. L. Mirchi Service CorporitioA Kewaunee N(uclear Power Plant Wiscons1n Public cc:

Fole~y & Lardner Bradley 0. Jackson Attention: Mr.

One South Pinckney Street P. 0. Box 1497 Hadison, Wlsconsin 53701-1491 Chairman Town of Carlton Route I 54216 Kewaunee, Wisconsin Mr. Harold Retkelberg, Chairman Kewaunee County Board Kewaunee County Courthouse Kewaunee, Wisconsin 54216 Chairman Public Service Comlssion of Wisconsin Hill Farus State Office Building Hzdison, V1sconsln 53702 Attorney general 114 East, State Capitol Madison, Visconsin 4370Z U. S. Nuclear Regulatory Camission Resident Jnspectors Office Route fI, Box 999 Klewaunee, Wisconsin 54216 fit Regional Administrator - Regfon Coission U. S. Nuclear Regulatory 801 Warrenville Read Lisle, Illinois 605324-31 Mr. Robert S. Cullen Chief Enginetr Wisconsin Public Service Coisslon P. 0. Box 76S4 Madison, Misconsin 53707 REV. 9 E-31 12/02/2005

A 9UNITED STATES tNUCLEAR REGULATORY COMMISSION WASHINON. D.C. DV.WC.

S/

SAFETY EVALUATION 3YT.HE OFFICE (fCLEAR RMTOR REGULATION RELATING To ONSI! DISPOSAL OF -LEVEL RADUCT!VFLY CONTAMIPATEED SEW1E TREATMENT SL E AT THE KINNEE NUCLEMR POER PLA1 WISCONSIN PUBLIC SERVICE CORPORATION WISCONSIN POWER AND Lit COMPA tADISON GAM AHO FULCTRIC COI4PAN DOCKET NO. 50-305 1.0 1IITRODCT1 Q gy ltter dated 0tober l7, 1995, s supplented on Moveu(er 3, 1995, Wisconsin Publit Service Corporation (the license.) requested approval for the onsite disposal of contaminated sewage sludge stlllar to a previous disposal request that was approved by the HRC on June 17. 1992.

2.0 IBACKROUN In a letter dated September 12, 1989, the licensee requested authorization for the alternate disposal of very-low-level radioactive material. In a Safety Evaluation (SE) dated June 17, 1992, the HRC approved the licensee's request pursuant to 10 CFR 20.302 (new 10 CFR 20.2002) for the disposal of 15,000 cubic feet of contaminated waste sludge by land application at the Kewaunee Huclear Pover Plant (ltPP) location. lhe SE imposed the following boundary conditio ns:

1. The annual disposal must be less than a total activity of 0.2 mti.

2. The whole body dose to thu hypothetical maximally exposed individual must be less than 0.1 area/year.

3. The disposal must be at the same site.

The lIcensee completed the disposal of the contaminated waste sludge discussed in the SE dated June 17, 1992. The licensee Is now requesting authorization to dispose of additional contaminated waste sludge within the boundary conditions of the previously approved disposal.

3.0 EYAIUATION The licensee has proposed to dispose of approximately 50 gallons (800 cubic feet) of sewage sludge similar to the material approved for disposal In the SE dated June 17. 1992. The principal radionuclides identified In the waste sludge and their activity based on measurements in May 1995 are: Co-58, E-32 REV. 9 12/02/2005

-z -

0.0g09 CI; Co-60, C.0008 mCi; and Cr-S1, 0.0006 mCi. The total combined activity Is0.0023 MCI. This activity Is well below the boundary value of 0.2 .Ci. Additionally, Cr-51 with itshort half-life (27.7 day) wtill have undergone significant decay from Its initial value of 0.0006 Ci.

The licensee has comitted that the new disposal will meat all the radiological boundary conditions, on a cimulative basis, contained inthe SE for the 10 CFR 20.302 application approved on June 17, 1992. Additionally, the licensee has stated that all applicable permits for this disposal have been obtained from the Wisconsin Department of Natural Resources.

4.0 CONCLUS[08 The staff finds the licensee's proposal to dispose of the low-level radioactive waste sludge pursuant to 10 CFR 20.2002. on the licensee's site (see Attachment). Iswithin the radiological boundary conditions approved in the June 17, 1992, SER and Is therefore acceptable.

The licensee isrequired to permanently incorporate this odification Into the Offsite Case Calculation Manual as an Appendix, and to ensure that future modifications of these comitments are reported to the RRC.

Principal Contributor: S. Klementowicz Date: Novenber 13, 199S

Attachment:

KNPP Site Area Nap E-33 REV. 9 12/02/2005

-I E-34 REV. 9 12/02/2005 It([

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4. 0 UNITED STATES

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• A4 0 NUCLEAR REGULATORY COMMISSION S K WAMHNMCTO, D.C. mu

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• 0 April 9, 19S?'

Mr. M. L. Marchi Manager - Nuclear Business Group Wisconsin Public Service Corporation Post Office Box 19002 Green Bay, WI 54307-9002 SUBJ[CT: ONSITE DISPOSAL OF C0H1AMINAIED SLUDGE PURSUANT 10 10 CFR 20.2002 (TAC NO. M97411)

Oear Mr. Marchi:

By letter dated December 10, 1996, you requested that the U.S. Nuclear Regulatory Commission (NRC) review the applicability of a 10 CFR 20.203 (now 20.2002) application appreved on June 17, 1992, for additional disposals of a similar nature.

The staff has completed its review of your request and agrees with your deternination that the 10 CFR 20.203 application for onsite disposal of sludge contaminated with licensed radioactive material, which was approved an June 17, 1992, contains bounding conditions that are applicable for additional onsite disposals of a similar nature. A copy of the Safety Evaluation is enclosed.

Sincerely, Richard J. Larfe Proect Manager Project Directorate 111-3 Division of Reactor Projects 1II/IV Office of Nuclear Reactor Regulation Docket No. 50-305

Enclosure:

Safety Evaluation cc: See next page Xa Ev~s CN? CS$ X~KNIP M W Seitz (%'P) UL archj D2 H D Cd (SM C R ktcin1 D2 MM4 I'W PNSRAC)

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E-35 REV. 9 12/02/2005

Mr. M. L. Marchi Wisconsin Public Service Corporation Kewaunee Nuclear Power Plant cc, Foley & Lardner Attention: Mr. Bradley 0. Jackson One Sobth Pinckney Street P.0. Box 1497 Madison, Wisconsin 53701-1497 Chairman Town of Carlton Route 1 Kewauree, Wisconsin 54216 Mr. Harold Reckelberg. Chairman Kuwaunee County Board Kewaunee County Courthouse Kewaunee, Wisconsin 54216 Chailrmn Wisconsin Public Service Commission 610 N. Whitney Way Madison, Wisconsin 53705-2729 Attorney General 114 East, State Capitol Madison, Wisconsin 53702 U. S. Nuclear Regulatory Conmission Resident Inspectors Office Route t1, Box 999 Kewaunee, Wisconsin 54216 Regional Administrator - Region IIl U. 5. Nuclear Regulatory Conuisslon 801 Warrenavlle Road Lisle, Illinois 60532-4531 Mr. Robert S. Cullen Chief Engineer Wisconsin Public Service Cormission 610 N. Whitney Way Madison, Wisconsin 53705-2829 E-36 REV. 9 1210212005

UNITED STATES NUCLEAR REGULATORY COMMISSION WIIIF Wl$$INCON D.C . W-~

SAFETY EVALUATION BY THE OFFICE OF NOCLEAR REACTOR REGULATION RELATING TO ONSITE DISP-SAL OF CONTAMINATED RLUP AT THE kEUAUNFE NUCLEAR POWER PLANT WISCONSIN PUBLIC -SERYICE CORPORATION WISCONSIN POWER AND LIGHT COMPANY MADISON GAS AND ELECTRIC COMPANY QGCKET NO.50-30N

!.L INIRODUCTION By letter dated December 10, 1996, Wisconsin Public Service Corporation (the I icensee) requested that the U.S. Nuclear Regulatory Commission (NRC) review its determination that NRC approval, pursuant to 10 CFR 20.2002. for the ontite disposal of contaminated sludge at the Kewaunee Nuclear Power Plant (KNPP) is not required, provided such disposals are conducted within the limits and bounding conditions approved by the NRC In its June 17, 1592, Safety Evaluation (SE),

2.0 BACKGROUND

In a letter dated September 12, 1989, the licensee requested authorization for the alternate disposal of sludge contaminated with licensed radioactive material. In an SE dated June I7, 1992, the NRC approved the licensee's request pursuant to 10 CFR 20.302 (new 10 CFR 20.2002) for the disposal of 15,000 cubic feet of contaminated waste sludge by land application at the KNPP location. The SE imposed boundary conditions as follows:

1. The annual disposal must be less than a total activity of 0.2 mCi;

2. The whole body dose to the hypothetical maximally exposed individual must be less than 0.1 arem/year; and

3. The disposal must be at the same site.

The SE also stated that for any repetitive disposals, the licensee must reapply to the KRC when a particular disposal would exceed the boundary conditions.

'1.0 LALUAIIOH The licensee has determined that NRC approval for future onsite disposals of sludge contaminated with licensed radioactive material is not required provided the disposals comply with the limits and conditions of the SE issued on June 17, 1992. The licensee has also developed a sludge sampling and analysis procedure that implements the guidance contained in NRC Information E-37 REV. 9 12/02/2005

Notice 88-22. Specifically, the licensee's procedure will require the analysis of sludge samples using a detection system design and operating characteristics that yield a laer limit of detection for Co-58. Co-6O, Cs-134. and Cs-137 consistent with measurements of environmental samples. The licensee has provided a site map (attached) that specifies the acceptable onsite disposal areas for the contaminated sludge.

4.0 CONCLUSIOM The staff agrees with the licensees determination that additional onsite disposals of contaminated sludge, which are conducted within the bounding limits and conditions contained in the June 17, 1992, SE and within the areas specified in the attached site map, do not require specific NRC approval.

The licensee should permanently incorporate this Safety Evaluation into the OffsIte Dose Calculation Manual as an Appendix.

Principal Contributor: S. Kle~wentowicz Date: April 9, 1S99

Attachment:

KNPP Site Map E-38 REV. 9 12/02/2005

I I r I ' X r r r r r r r r r r r r I

E-39 REV. 9 12/02/2005

Appendix C Kewaunee Power Station Annual Radioactive Effluent Release Report January 1 - December 31, 2003 Revised Section 3.0 Liquid Effluents

(

_M Table of Contents Section Description

h. 3.0 Liquid Effluents ............................................. C4 3.1 Lower Limits of Detection (LLD) for Liquid Effluents .................................. C4 3.2 Liquid Batch Release Statistics ............................................. C6

- 3.3 Liquid Effluent Data ............................................. C6 Table 3.1 Liquid Effluents - Summation of all Releases ............................... C7 Table 3.2A Liquid Effluents - Batch Releases 1st Quarter ............................. C8 Table 3.2B Liquid Effluents - Batch Releases 2nd Quarter .......................... CIO Table 3.2C Liquid Effluents - Batch Releases 3rd Quarter ........... ............... C12 Table 3.2D Liquid Effluents - Batch Releases 4th Quarter ........... ............... C14 Table 3.3A Liquid Effluents - Continuous Releases 1st Quarter .................. C16 Table 3.3B Liquid Effluents - Continuous Releases 2nd Quarter ................C18 Table 3.3C Liquid Effluents - Continuous Releases 3rd Quarter ................. C20 Table 3.3D Liquid Effluents - Continuous Releases 4th Quarter ................. C22 Table 3.4 Dose From Liquid Effluents ................... .......................... C24 Page C2 of C25

0.0

SUMMARY

During 2003 all liquid radioactive effluents from the Kewaunee Nuclear Plant were well below regulatory limits. For liquid effluent stream, the quarterly limit most closely approached was:

LIQUID: Ingestion Pathway-Organ GI-LLI Quarterly Limit (mRems) 5 Actual Dose (mRems) 0.01252 (4 th Quarter)

% of Limit 0.25 Page C3 of C25

3.0 LIOUID EFFLUENTS 3.1 Lower Limits of Detection (LLD) for Liquid Effluents Liquid radioactive effluents are released as both batch releases and continuous releases. Each batch is sampled prior to release and analyzed for gamma emitters and tritium. A fraction of each sample is retained for a monthly proportional composite which is then analyzed for Gross Alpha, Strontium 89, Strontium 90 and Iron 55.

The LLD's for liquid batch release radioanalyses, as listed in Table 4.3 of the Kewaunee Nuclear Power Plant Off-Site Dose Calculation Manual, are:

Analysis LLD hLCi/ml Principal Gamma Emitters 1.00 E-06 Iodine 131 1.00 E-06 Tritium 1.00 E-05 Gross Alpha 5.00 E-07 Strontium 89, 90 5.00 E-08 Iron 55 1.00 E-06

_. The actual obtained "a priori" LLD values for batch releases are shown below.

I Average a I 6.. I Isotope 1 st Quarter 2nd Quarter 3rd Quarter 4th Quarter priori LLD I 6W J

i Mn-54 6.62E-10 6.62E-08 6.62E- 10 6.62E- 1O (J.LCi/ml) 1.70E-08 i

Fe-59 1.46E-09 1.46E-09 1.46E-09 1.46E-09 1.46E-09 i

Co-58 6.50E-08 6.50E-10 6.50E-10 6.50E-10 1.67E-08 I.. I Co-60 9.61 E-10 f9.61E-10 9.61E-08 9.61E-10 2.47E-08 Zn-65 1.64E-09 1.64E-09 1.64E-09 1.64E-09 1.64E-09 Mo-99 4.70E-09 4.69E-09 4.69E-09 4.69E-09 4.69E-09 Cs-134 5.22E- 10 5.22E- 10 5.22E- 10 I 5.22E- 1 0 5.22E-10 Cs- 137 6.44E- 10 6.44E-08 6.44E- 10 6.44E-I10 1.66E-08 Ce-141 7.62E-08 3.81E-10 4.21E-08 3.81E-10 2.98E-08 Ce- 144 1.711E-07 5.9OE-07 .71E-077"

.:...Fii

.,....,..... 4.86E-07 3.54E-07 I-131 4.06E- 10 4.06E- 10 4.06E-08 4.06E-08 2.05E-08 H-3 3.33E-06 3.82E-06 3.57E-06 3.76E-06 3.62E-06 Sr-89 1.63E-08 2.87E-08 1.60E-08 1.1OE-08 1.80E-08 Sr-90 6.73E-09 f 9.27E-09 6.60E-09 7.87E-09 7.62E-09 Gross Alpha 8.23E-09 7.23E-09 5.40E-09 5.83E-09 6.68E-09 Fe-55 7.93E-07 8.40E-07

...- . i 8.23E-07 9.43E-07 8.50E-07 Page C4 of C25

Continuous liquid releases are grab sampled weekly and analyzed for principal gamma emitters. A fraction of each weekly sample is retained for a monthly proportional composite which is then analyzed for Tritium, Gross Alpha, Strontium 89, Strontium 90 and Iron 55.

The LLD's for liquid continuous release radioanalyses, as listed in Table 4.3 of the Kewaunee Nuclear Power Plant Off-Site Dose Calculation Manual, are:

Analysis LLD (pCi/ml)

Principal Gamma Emitters 5.00 E-07 Iodine 131 1.00 E-06 Tritiumn 1.00 E-05 Gross Alpha 5.00 E-07 Strontium 89, 90 5.00 E-08 Iron 55 1.00 E-06 The actual obtained "a priori" LLD values for continuous releases are shown below.

C Average a I Isotope 1st Quarter 12nd Quarter l 3rd Quarter 4th Quarter i I priori LLD I

4 (ICi/ml)

Mn-54 2.73E-08 1.56E-08 1.10E-08 1.91E-08 1.83E- 08 Fe-59 2.43E-10 4.20E-08 2.43E-10 1 2.43E-10j 1.07E-08 Co-58 1.20E-08 1.35E-08 2.42E-08 1.53E-08 1.63E-08 Co-60 2.04E-08 3.23E-08 2.89E-08 1.44E-08 2.40E-08 Zn-65 2.73E-08 2.73E- 10 2.73E- 10 3.02E-08 1.45E-08 Mo-99 1.48E-07 2.84E-07 3.19E-07 2.17E-07 2.42E-07 Cs- 134 9.58E-09 8.69E- 11 1.98E-08 3.88E-08 1.71 E-08 Cs- 137 2.90E-08 1.07E- 10 3.51E-08 1.07E-10I 1.61E-08 Ce- 141 3.45E-08 1.56E-08 1.68E-08 1.68E-08 2.09E-08 Ce-144 7.55E-08 1.34E-07 1.67E-07 1.90E-07 1.41E-07 1-131 4.46E-08 1.33E-08 1.69E-08 6.76E-09 2.04E-08 H-3 3.33E-06 3.82E-06 3.57E-06 3.76E-06 3.62E-06 Sr-89 1.67E-08 2.70E-08 1.63E-08 1.26E-08 1.82E-08 Sr-90 7.55E-09 8.70E-09 7.68E 09( 7.78E-09 7.93E- 09 Gross Alpha 5.12E-09 4.77E-09 5.10E-09 4.82E-09 4.95E-09 Fe-55 7.9OE-07 8.38E-07 8.22E-07 9.18E-07 8.42E-07 Page C5 of C25

3.2 Liquid Batch Release Statistics The following is a summation of all liquid batch releases made during 2003.

Release Type Number Gallons Released A SGBT Monitor Tk. 10 93650.0 B SGBT Monitor Tk. 9 84704.0 A CVC Monitor 16 99365.0 B CVC Monitor 14 89215.0 Both WCTs 8 14480.0 Total time for all batch releases ............. 23632.0 Min.

Maximum time for a batch release .............. 837.0 Min.

_I- Minimum time for a batch release .............. 55.0 Min.

Average time for a batch release .............. 414.6 Min.

3f-6- 3.3 Liquid Effuent Data The following Table 3.1 presents a quarterly summation of the total activity released and average concentration for all liquid effluents. It also presents the gross alpha activity released, volume of waste released and volume of dilution water used. Tables 3.2 and 3.3 are monthly summations of the same information in Table 3.1. Table 3.2 contains the quantity of the individual isotopes released to the unrestricted area for batch releases. Table 3.3 presents a monthly summation of gross radioactivity, tritium, gross alpha and isotopic activity for the secondary blowdown and leakage releases. It also presents the monthly total volume for these releases and dilution volumes. Table 3.4 presents the doses from liquid effluents for each quarter and the calculated doses this year from liquid effluents.

Page C6 of C25

TABLE 3.1 Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Summation of all Releases 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission and Activation Products Total Release Excluding H3 and Dissolved Gases (Ci) 4.878E-003 3.860E-002 6.136E-003 3.116E-003 Average Concentration (ACi/ml) 4.596E-01 1 2.447E-010 2.981E-011 1.873E-01 1 Tritium Total Release (Ci) 2.163E+002 3.794E+001 1.518E+001 1.094E+001 Average Concentration (gCi/ml) 2.038E-006 2.405E-007 7.376E-008 6.573E-008

% of Tech. Spec.

Limit(3.OE-3 IiCi/ml) 6.793E-002 8.017E-003 2.459E-003 2.191E-003 Dissolved Gases Total Release (Ci) 1.553E-004 0.OOOE+000 0.OOOE+000 0.OOOE+000 Average Concentration (ipCi/ml) 1.464E-012 0.OOOE+000 0.OOOE+000 0.OOOE+000

% of Tech. Spec.

Limit(2.OE-4 liCi/ml) 7.318E-007 0.OOOE+000 0.OOOE+000 0.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 0.OOOE+000 Volume of Waste Released Batch (liters) 6.058E+005 5.012E+005 2.530E+005 8.374E+004 Continuous (liters) 2.315E+007 2.438E+007 1.991E+007 2.055E+007 Total (liters) 2.376E+007 2.488E+007 2.016E+007 2.064E+007 Volume of Dilution Water Batch (liters) 8.022E+009 8.386E+009 5.239E+009 1.761E+009 Continuous (liters) 9.812E+010 1.494E+01 1 2.006E+01 1 1.646E+01 1 Total (liters) 1.061E+0 1I 1.577E+011 2.058E+011 1.664E+01 1 Page C7 of C25

TABLE 3.2A Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases January February March Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 7.245E-004 1. 169E-003 1.595E-003 3.489E-003 Avg. Conc.

(p.Ci/ml) 3.73 1E-0 10 4.636E-010 4.483E-010 Tritium Total Release (Ci) 5.241E+001 8.403E+001 7.982E+001 2.163E+002 Avg. Conc.

(pCi/ml) 2.699E-005 3.332E-005 2.243E-005 Dissolved Gases Total Release (Ci) 2.895E-006 1.070E-005 1.418E-004 1.553E-004 Avg. Conc.

(gCi/ml) 1.491E-012 4.242E-012 3.984E-01 I Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(gCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Volume of Waste Released (liters) 1.455E+005 1.980E+005 2.623E+005 6.058E+005 Volume of Dilution Water (liters) 1.942E+009 2.522E+009 3.558E+009 8.022E+009 Page C8 of C25

TABLE 3.2A (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases Isotope (Ci) January February March Total H-3 5.241 E+001 8.403E+001 7.982E+001 2.163E+002 Fe-55 1.746E-004 5.940E-004 4.984E-004 1.267E-003 Co-58 5.353E-005 3.756E-005 1.949E-004 2.860E-004 Co-60 4.53 1E-005 6.684E-005 2.868E-004 3.989E-004 Sr-89 2.328E-007 O.OOOE+000 O.OOOE+000 2.328E-007 Sr-90 4.364E-008 1.980E-008 4.197E-007 4.831E-007 Ag-llOm 9.400E-005 1.401E-004 1.903E-004 4.244E-004 Sb-125 3.569E-004 3.232E-004 4.244E-004 1.104E-003 Xe-133 O.OOOE+000 1.070E-005 1.418E-004 1.525E-004 Xe-135 2.8951E-006 O.OOOE+000 O.OOOE+000 2.895E-006 Cs-137 O.OOOE+000 7.278E-006 O.OOOE+000 7.278E-006 Total 5.241E+001 8.403E+001 7.982E+001 2.163E+002 Page C9 of C25

TABLE 3.2B Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases April May June Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.575E-002 1.510E-002 4.692E-003 3.554E-002 Avg. Conc.

(OiCi/ml) 5.389E-009 1. 134E-008 1.135E-009 Tritium Total Release (Ci) 2.549E+OO1 2.987E+OOO 9.463E+OOO 3.794E+OO1 Avg. Conc.

(1ICi/ml) 8.723E-006 2.243E-006 2.290E-006 Dissolved Gases Total Release (Ci) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Avg. Conc.

(iLCi/ml) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Gross Alpha Activity Total Release (Ci) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Avg. Conc.

(gCi/ml) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Volume of Waste Released (liters) 2.413E+005 1. 102E+005 1.497E+005 5.012E+005 Volume of Dilution Water (liters) 2.922E+009 1.332E+009 4.132E+009 8.386E+009 Page C10 of C25

TABLE 3.2B (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases Isotope (Ci) April May June Total H-3 2.549E+001 2.987E+000 9.463E+000 3.794E+001 Cr-51 2.294E-004 3.848E-003 2.091E-004 4.287E-003 Mn-54 3.093E-004 9.073E-005 2.245E-006 4.023E-004 Fe-55 4.585E-003 2.093E-003 2.845E-003 9.524E-003 Co-58 1.249E-003 5.339E-003 1.382E-003 7.970E-003 Fe-59 O.OOOE+000 1.1 18E-003 1.254E-004 1.244E-003 Co-60 6.994E-003 8.813E-004 6.984E-005 7.945E-003 Sr-90 2. 100E-006 9.585E-007 1.303E-006 4.361E-006 Nb-95 O.OOOE+000 5.280E-004 2.308E-005 5.510E-004 Zr-95 O.OOOE+000 3.285E-004 O.OOOE+000 3.285E-004 Ag-i lOm 2.379E-003 8.169E-004 3.418E-005 3.230E-003 Sn- 113 O.OOOE+000 5.564E-005 O.OOOE+000 5.564E-005 Total 2.55 1E+001 3.002E+000 9.467E+000 3.798E+001 Page CI1 of C25

TABLE 3.2C Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases July August September Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.1 84E-003 3.319E-004 5.801E-004 2.096E-003 Avg. Conc.

(glCi/ml) 3.277E-010 7.164E-010 4.989E-010 Tritium Total Release (Ci) 1.032E+001 1.222E+000 3.642E+000 1.518E+001 Avg. Conc.

(11Ci/ml) 2.857E-006 2.637E-006 3.13 1E-006 Dissolved Gases Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 0.OOOE+000 Avg. Conc.

(RCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 0.OOOE+O000 Avg. Conc.

(pCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Volume of Waste Released (liters) 1.773E+005 3.541E+004 4.03 1E+004 2.530E+005 Volume of Dilution Water (liters) 3.613E+009 4.633E+008 1.163E+009 5 .239E+009 Page C12 of C25

TABLE 3.2C (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases Isotope (Ci) July August September Total H-3 1.032E+001 1.222E+000 3.642E+000 1.518E+001 Fe-55 4.078E-004 8.144E-005 9.271E-005 5.820E-004 Co-58 4.268E-004 7.769E-005 9.718E-006 5.142E-004 Co-60 8.535E-005 1.637E-005 2.274E-005 1.245E-004 Sr-90 3.724E-007 7.436E-008 8.465E-008 5.314E-007 Ag-110m 1.828E-004 6.880E-005 1.484E-004 4.OOOE-004 Sb-125 8.075E-005 8.757E-005 3.064E-004 4.747E-004 Total 1.032E+001 1.222E+000 3.642E+000 1.519E+001 Page C13 of C25

TABLE 3.2D Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases October November December Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 4.166E-005 1.145E-003 1.923E-003 3.11 OE-003 Avg. Conc.

(ACi/ml) 4.743E-010 9.757E-010 3.848E-009 Tritium Total Release (Ci) 7.655E-005 4.104E+000 6.831 E+000 1.094E+001 Avg. Conc.

(OlCi/ml) 8.717E-010 3.498E-006 1.367E-005 Dissolved Gases Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(ACi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 0.OOOE+000 O.OOOE+000 0.OOOE+000 Avg. Conc.

(1Ci/ml) O.OOOE+000 O.OOOE+000 0.000E+000 Volume of Waste Released (liters) 7.154E+003 4.063E+004 3.595E+004 8.374E+004 Volume of Dilution Water 6 (liters) 8.782E+007 1.173E+009 4.997E+008 1.761E+009 Page C14 of C25

TABLE 3.2D (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Batch Releases Isotope (Ci) October November December Total H-3 7.655E-005 4.104E+000 6.83 1E+000 1.094E+001 Fe-55 3.935E-005 2.235E-004 1.977E-004 4.606E-004 Co-58 2.277E-006 5.960E-005 1.175E-004 1.794E-004 Co-60 O.OOOE+000 2.789E-005 4.286E-005 7.075E-005 Sr-89 2.862E-008 1.625E-007 1.4388E-007 3.350E-007 Ag-110m 0.000E+000 4.873E-004 3.897E-004 8.771E-004 Sb-125 0.000E+000 3.466E-004 1.175E-003 1.522E-003 Total 1.182E-004 4.105E+000 6.833E+000 1.094E+001 Page C15 of C25

TABLE 3.3A Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases January February March Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.720E-005 9.236E-004 4.482E-004 1.389E-003 Avg. Conc.

(giCilml) 5.089E-013 3.026E-O1 l 1.326E-Ol 1 Tritium Total Release (Ci) 2.861E-002 O.OOOE+OOO O.OOOE+OOO 2.861E-002 Avg. Conc.

(iCilml) 8.465E-010 O.OOOE+OOO O.OOOE+OOO lo Dissolved Gases Total Release (Ci) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Avg. Conc.

(gLCi/ml) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Gross Alpha Activity Total Release (Ci) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Avg. Conc.

(pCi/ml) O.OOOE+OOO O.OOOE+OOO O.OOOE+OOO Volume of Waste Released (liters) 7.280E+006 7.41 lE+006 8.461E+006 2.31 5E+007 Volume of Dilution Water (liters) 3.380E+010 3.053E+010 3.380E+010 9.812E+010 Page C16 of C25

TABLE 3.3A (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases Isotope (Ci) January February March Total H-3 2.861E-002 O.OOOE+000 O.OOOE+000 2.861E-002 Fe-55 O.OOOE+000 8.637E-004 3.587E-004 1.222E-003 Sr-89 O.OOOE+000 5.974E-005 8.948E-005 1.492E-004 Sr-90 1.720E-005 2.353E-007 O.OOOE+000 1.744E-005 Total 2.862E-002 9.236E-004 4.482E-004 3.OOOE-002 Page C17 of C25

TABLE 3.3B Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases April May June Tcital Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.063E-003 1.020E-003 9.729E-004 3.056E-003 Avg. Conc.

(jiCi/ml) 3.047E-0 1I 2.079E-01 1 1.487E-01 1 Tritium Total Release (Ci) O.OOOE+000 0.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(JACi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Dissolved Gases Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(1.Ci/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(gCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Volume of Waste Released (liters) 8.030E+006 9.145E+006 7.207E+006 2.438E+007 Volume of Dilution Water (liters) 3.489E+010 4.906E+010 6.541E+010 1.494E+01 1 Page C18 of C25

TABLE 3.3B (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases Isotope (Ci) April May June Total Fe-55 1.021E-003 9.589E-004 9.369E-004 2.917E-003 Sr-89 4.157E-005 6.036E-005 3.603E-005 1.380E-004 Sr-90 8.150E-008 8.363E-007 O.OOOE+000 9.179E-007 Total 1.063E-003 1.020E-003 9.729E-004 3.056E-003 Page C19 of C25

TABLE 3.3C Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases July August September Total Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.224E-003 1.337E-003 1.480E-003 4.040E-003 Avg. Conc.

(giCi/ml) 1.81 lE-01 1 1.978E-011 2.262E-01 1 Tritium Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(,Ci/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Dissolved Gases Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(giCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 O.OOOE+000 O.OOOE+000 Avg. Conc.

(,uCi/ml) O.OOOE+000 O.OOOE+000 O.OOOE+000 Volume of Waste Released (liters) 5.684E+006 7.407E+006 6.818E+006 1.991E+007 Volume of Dilution Water (liters) 6.759E+010 6.759E+010 6.541E+010 2.006E+01 I Page C20 of C25

TABLE 3.3C (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases Isotope (Ci) July August September Total Fe-55 1.184E-003 1.291E-003 1.432E-003 3.907E-003 Sr-89 3.948E-005 4.305E-005 4.773E-005 1.302E-004 Sr-90 8.848E-008 2.516E-006 O.OOOE+000 2.604E-006 Total 1.224E-003 1.337E-003 1.480E-003 4.040E-003 Page C21 of C25

TABLE 3.3D Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases Octo ber November December To tal Gross Radioactivity Total Release Excluding H3 and Dissolved Gases (Ci) 1.716E-006 1.821E-006 2.629E-006 6.166E-006 Avg. Conc.

(ACi/ml) 2.538E-014 2.879E-014 7.780E-014 Tritium Total Release (Ci) 0.OOOE+000 0.OOOE+000 0.OOOE+000 0.OOOE+000 Avg. Conc.

(ACi/ml) O.OOOE+000 0.OOOE+000 0.OOOE+000 Dissolved Gases Total Release (Ci) 0.OOOE+000 0.OOOE+000 0.OOOE+000 0.OOOE+000 Avg. Conc.

(Oci/ml) 0.OOOE+000 0.OOOE+000 0.OOOE+000 Gross Alpha Activity Total Release (Ci) O.OOOE+000 O.OOOE+000 0.OOOE+000 0.OOOE+000 Avg. Conc.

(ALCi/ml) 0.OOOE+000 O.OOOE+000 O.OOOE+000 Volume of Waste Released (liters) 5.719E+006 6.069E+006 8.765E+006 2.055E+007 Volume of Dilution Water (liters) 6.759E+010 6.323E+010 3.380E+010 1.646E+01 1 Page C22 of C25

TABLE 3.3D (Con't)

Annual Radioactive Effluent Release Report 2003 Liquid Effluents - Continuous Releases Isotope (Ci) October November December Total Sr-90 1.716E-006 1.821E-006 2.629E-006 6.166E-006 Total 1.7166E-006 1.821E-006 2.629E-006 6.166E-006 Page C23 of C25

hi Table 3.4 Annual Radioactive Effluent Report 2003 Dose From Liquid Effluents The dose to a member of the public from total liquid radioactive releases for each quarter was below the ODCM limits of 1.5 mrems to the total body and less than or equal to 5 mrems to any organ.

Additionally, the dose to a member of the public from total liquid radioactive releases for the year was below the ODCM limits of 3 mrems to the total body and less than or equal to 10 mrems to any organ.

Instantaneous release concentrations are limited by the individual radionuclide concentrations established in 10 CFR 20, Appendix B, for unrestricted areas. During the report period, none of the isotopes released exceed the concentrations specified in Appendix B. The following offsite doses were calculated using equation 1.5 from the Kewaunee ODCM.

Organ Dose Quarterly Percent 1st Qtr Dose Total Limit of Limit mRem mRem Total Body 1.636E-003 1.5 0.11 Bone 8.578E-005 5.0 0.00

%N_ Liver 1.670E-003 5.0 0.03 Thyroid 1.571E-003 5.0 0.03 Kidney 1.599E-003 5.0 0.03

%. Lung 1.588E-003 5.0 0.03 GI-LLI 1.639E-003 5.0 0.03 6.-

Organ Dose Quarterly Percent 2nd Qtr Dose Total Limit of Limit mRem rnRem Total Body 4.295E-004 1.5 0.03 Bone 1.867E-004 5.0 0.00 Liver 4.772E-004 5.0 0.01 Thyroid 2.391E-004 5.0 0.00 Kidney 2.535E-004 5.0 0.01 Lung 2.995E-004 5.0 0.01 GI-LLI 1.920E-002 5.0 0.38 Page C24 of C25

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Table 3.4 (Con't)

Annual Radioactive Effluent Report 2003 Dose From Liquid Effluents Organ Dose Quarterly Percent 3rd Qtr Dose Total Limit of Limit mRem mRem Total Body 6.790E-005 1.5 0.00 Bone 8.386E-005 5.0 0.00 Liver 7.860E-005 5.0 0.00 Thyroid 5.515E-005 5.0 0.00 Kidney 5.516E-005 5.0 0.00 Lung 6.777E-005 5.0 0.00 GI-LLI 9.282E-005 5.0 0.00 Organ Dose Quarterly Percent 4th Qtr Dose Total Limit of Limit mRem mRem Total Body 7.922E-005 1.5 0.01 Bone 5.618E-005 5.0 0.00 Liver 6.847E-005 5.0 0.00 Thyroid 6.453E-005 5.0 0.00 Kidney 6.456E-005 5.0 0.00 Lung 6.652E-005 5.0 0.00 GI-LLI 8.738E-005 5.0 0.00 Calculated Dose This Year Organ Dose Annual Percent Total Limit of Limit mRem mRem Total Body 2.213E-003 3.0 0.07 Bone 4.125E-004 10.0 0.00 Liver 2.294E-003 10.0 0.02 Thyroid 1.930E-003 10.0 0.02 Kidney 1.973E-003 10.0 0.02 Lung 2.021E-003 10.0 0.02 GI-LLI 2.102E-002 10.0 0.21 Page C25 of C25