ML20138A366

Rev 15 to Procedure FNP-O-M-011, Offsite Dose Calculation Manual

ML20138A366
Person / Time
Site:	Farley
Issue date:	04/25/1996
From:	Hill R SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20138A332	List: ML20138A328 ML20138A350 ML20138A366
References
FNP-O-M-011, FNP-O-M-11, NUDOCS 9704280067
Download: ML20138A366 (253)

Text

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PROCEDURE REQUEST FORM - I FNP-0-AP-1 -  !

1. Procedure Numbsr 1

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Temporary Change One Time Only or Reg'd by Plant Conditions.TCN Dates this temporary changs is effective: From_- Through

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FARLEY WUCLEAR PLANT - 10 CFR 50.59 EVALUATION Ptg3 i of 7 RE: Of fsita Dos 3 Calculation Manual, Rivision 15 f~h k

SECTION A:

Unit Number: One (X) Two(X) Shared ( )

Document Title or Number: ODCM Revision or TCN Number: 15 SECTION B: 10 CFR 50.59 APPLICABILITY:

Does the document to which this evaluation applies represents:

1. [ ] Yes [X] No A change to the plant as described in the FSAR?

Basis for answer: No change to the FSAR is required, however, the Offsite Dose Calculation Manual (ODCM) is being revised to correct a typographical error and to delete the reference to the Environmental Protection Agency (EPA) cross-check program (see attached markups). This EPA service is no longer available (beginning January 1,1996) for satisfying the Interlaboratory Comparison Program (ICP) requirements. Instead, an independent laboratory will be used to provide this service.

2. [ ] Yes [X] No A change to procedures described in the FSAR?

, Basis for answer: Procedures associated with implementation of the EPA cross-(g check program are not described in the FSAR.

3. [ ] Yes [X] No A test or experiment not described in the FSAR?

Basis for answer: The ICP is integral to an acceptable environmental monitoring program and therefore is not considered to be a test or experiment.

4. [ ] Yes [X] No A change to the Technical Specifications?

Basis for answer Programmatic requirements for the ICP are contained in Technical Specification 6.8.3.f.iii. These changes to the ODCM are consistent with those requirements. PORC approval of these changes is required per Technical Specification 6.14.2.

If questions 1,2,3, or 4 in Section B are answered "YES", then PORC review of the safety evaluation (Section D) is reauired prior to implementation.

SECTION C:

Preparer:N/Y8fd Date:J/dks Reviewed By: Date:

ReviewerR d. MDateAMVr6 Reviewed B . _s Date-Reviewed By:N/ od%/ec Date:/1/-16 Approved By:_ 3bwuDate: A/H/#f, Reviewed By: /f Date: FNP Approved.W Date:_ f////96 Reviewed By: Date: PORC ReviewM Date#/N/96 C Reviewed By: Date: NORB Review: Date:

7 FAALEY _WUCLEAR PLANT - 10 CFR 50.59 EVALUATION P:ga 2 of 7 RE: Of fsite Dos) Calculation Manual, Revision 15 SECTION D: SAFETY EVALUATION I

1. [ ] Yes [X] No May the proposed activity increase the probability of occurrence of  !

an accident previously evaluated in the FSAR7 Basis for answer: See attached safety evaluation.

2. [ ] Yes [X] No May the proposed activity increase the consequences of an ..

accident previously evaluated in the FSAR7 l

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. Basis for answer: See attached safety evaluation. l

3. [ ] Yes [X] No May _the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously ,

evaluated in the FSAR7 l Basis for answer: See attached safety evaluation.

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4. [ ] Yes [X] No May the proposed activity increase the consequences of a ,

malfunction of equipment important to safety previously evaluated l in the FSAR7 l O Basis for answer: See attached safety evaluation.

5. [ ] Yes [X] No May the proposed activity create the possibility of occurrence of an accident of a different type than any previously evaluated in the r FSAR7 ,

Basis for answer: See attached safety evaluation

6. [ ] Yes [X] No May the proposed activity create the possibility of a malfunction of ,

equipment important to safety of a different type than any previously evaluated in the FSAR7-Basis for answer: See attached safety evaluation.

7. [ ] Yes [X] No Does the proposed activity reduce the margin of safety as defined  !

in the basis for any Technical Specification?

. Basis for answer: See attached safety evaluation.

If the answer to any of the seven questions in Section D is "YES", an unreviewed safety question may be indicated. Approval from the NRC is required before the O document / activity may be imolemented.

Intr: comp:ny Cerrespondence Ssuthern Nuclear Operating Company A l NEle96- 0107 I

DATE: March 26, 1996 RE:

Revision 15 to the FNP Offsite Dose Calculation Manual (ODCM)

FROM:

Mr.M.J. Ajiuni [

TO: Mr. W. R. Bayne

, Enclosed Licensing Services letter LS-96-019, dated 02/29/96, provides a safety evaluation, hand-marked 4

changes, and revised pages for revision 15 to the FNP ODCM. These changes are forwarded for Plant Farley Staff review and approval per item 2 of Technical Specification 6.14.2, which states that Licensee initiated J changes to the ODCM shall become effective after review and acceptance by the PORC and the approval of the Nuclear Plant General Manager.

DRC/maf: odem-rl5. doc I

Enclosure:

Licensing Services Letter LS-96-019 i

i cc: Mr. R. D. Hill . . . ... ....w/o l

Mr. W. H. Warren . . ...w/o O  :

a intracompany Correspondence SOuthem Nuclear Operating Company d 1

, Date. February 29,1996 LS-96-019 File: D.03.01 Re: Joseph M. Farley Nuclear Plant - Units 1 and 2 l Offsite Dose Calculation Manual - Revision 15 l

From: B. J. George I

To

M. J. Ajluni j l

Enclosed are proposed changes to the Farley Offsite Dose Calculation Manual and a supporting safety i evaluation related to the Interlaboratory Comparison Program (ICP) which were prepared in

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conjunction with Environmental Services. Spedfically, the proposed changes delete the reference to '

the Environmental Protection Agency (EPA) cross-check program. In the past, the EPA has provided an independent third party check of the contract laboratory used by the plant for performing '

measurements of radioactive material in environmental samples. However, beginning in 1996, this

service will no longer be provided by the EPA. Accordingly, the GPC Environmental Laboratory is currently in the process of making arrangements with a qualified independent laboratory that can

f. provide an equivalent ICP service. I

Regarding the 10 CFR 50.59 evaluation, note that the checklist screening criteria was answered with all "NO's." However, PORC approval of the ODCM changes will still be required in accordance with

Technical Specification 6.14.2. In addition to the hand-markups of the ODCM, you will also fmd

, enclosed the revised pages which can be used by plant documentation control for distributing the revision to holders of controlled copies of the manual.

3 Should you have any questions regarding any of this information, please call Tom Milton at j . extension 7867. .-

]

B. J. George i

BJG/TMM f Enclosure N

1. fc EO NEL cc: K. W. McCracken (w/ enclosure) p W. C. Carr (w/ enclosure)

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4 Farley Nuclear Plant I

ODCM Revision 15 Revised Pages l

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FARLEY NUCLFAR PLANT 10 CFR 50,59 EVALUATION Psg* 3 of 7 RE: Offsita Dosa Calculation Rmual, R* vision 15

Background

By letter dated August 10,1995 (Reference 1), the Nuclear Energy Institute (NEI) notified their utility administrative points of contact of a decision made by the  !

Environmental Protection Agency (EPA) to terminate Interlaboratory Companson Program (ICP) services. Under the EPA program, the licensees' contract laboratory was l obtaining intercomparison samples (e.g., water, milk, and paniculate filters) from the EPA for evaluation. Accordingly, the EPA provided a third party performance evaluation service to confirm the contract laboratory's capability to analyze the radionuclides in the samples. However, this service will no longer be provided by the EPA after December 31, 1995. Consequently, acceptable alternative arrangements for ICP services will have to be made. I References 1

1. NEIletter dated August 10,1995, from John F Schmitt to NEI Administrative Points I of Contact (attached).  !

2. Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 Offsite Dose Calculation Manual I (ODCM), Revision 14.

3. NRC letter dated December 26,1995, from Charles L. Miller to John F. Schmitt at NEI(attached).

4. Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment, Revision 1, dated February 1979 (attached).

5. Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 Technical Specifications.

6. Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 Final Safety Analysis Report Update. j Discussion Technical Specification 6.8.3.f.iii stipulates licensee participation in an ICP to ensure that independent (third party) checks are performed on the precision and accuracy of the measurements of radioactive material in environmental samples. This arrangement confirms a contract laboratory's capability to analyze the radionuclides in the samples, thereby satisfying the licensee's quality assurance program requirements for environmental monitoring, and also demonstrates compliance with 10 CFR 50, Appendix 1.

Accordingly, ODCM (Reference 2) Section 4.1.3 contains the following requirement:

In accordance with TechnicalSpecification 6.8.if(iii), analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the NRC. Analyses are required to be performed only in cases in which l the sample type andanalysis are the same as the sample type and i 4 analysisincludedin Table 4-1.

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FARLEY WUCLEAR PLANT - 10 CFR 50.59 EVALUATIgg p gs 4 of 7 j RE: Offsite Dos) Calculgtlon Manuel, R2 vision 15 j I

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As discussed in Reference 3, the NRC refers to Regulatory Guide 4.15 (Reference 4) j.

which describes an acceptable ICP. Section 6.3.2 of Regulatory Guide 4.15 indicates that  ;

licensees can pvticipate in the EPA's cross-check program m an equivalent program. l 4

Also stipulated is that participation should include all of the determinations (san ple l medium /radionuclide combinations) that are both offered by EPA and included in the  !

licensee's environmental monitoring program. The NRC further states in Reference 3 that j j they view the transition away from the EPA in this program as a change of contractor, '

l with no impact on the overall quality of the program. Consequently, the NRC concludes  ;

that a change should not constitute a departure from the program they originally .

j approved. Therefore, licensees would be expected to continue to maintain the same high-  !

, quality program previously experienced with the services provided by the EPA.  ;

l Accordingly, an equivalent ICP will be implemented at Farley beginning in 1996, which l

satisfies the requirements of Regulatory Guide 4.15. Therefore, Section 4.1.3 of the  !

ODCM is being revised to read as fo!!ows.

In accordance with Technical Specification 6.8.3.f(iii), analyses shall l l be performed on radioactive materials supplied aspart of an i Interlaboratory Comparison Program which satisfies the requirements i

ofRegulatory Guide 4.15, Revision 1, February 1979.

l The surveillance requirements associated with implementation of the ICP are contained in

} Section 4.1.3.3 of the ODCM. They read as follows:

i Either ,

a summary of the results obtained as part of the required .

l Interlaboratory Comparison Program shall be includedin the Annual l Radiological Environmental Operating Report, '

or l participants in the EPA cross-checkprogram shallprovide the EPA l

program code designationfor the plant in the AnnualRadiological  :

Environmental OperatingReport.

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Again, reference to the EPA cross-check program should be deleted. Accordingly, the surveillance requirements in Section 4.1.3.3 of the ODCM are being revised to read as follows:

A summary of the results obtainedaspart ofthe required Interlaboratory Comparison Program shall be includedin the Annual Radiological Environmental Operating Report.

Reference to the EPA cross-check program is also made in the programmatic information requirements contained in Sectica 7.1.2.3 of the ODCM which currently reads as follows:

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l FARLEY NUCtFAR PLANT - 10 CFR 50.59 EVALUATION Pter 5 of 7 RE: Offsit2 Dos: C lculttion Manual, Revision 15 l l

Also to be includedin each report are thefollowing: a summary (p)

'" description of the REMP; a map (s) of all sampling locations keyed to j a table giving distances and directionsfrom the centerpoint between the Unit 1 and Unit 2 plant vent stacks; the results ofland use '

censuses required by Section 4.1.2; and the results oflicensee \

participation in the Interlaboratory Comparison Program required by l Section 4.1.3. [The report shallinclude either a summary of the results obtainedaspart of the requiredInterlaboratory Comparison Program or,for licenseesparticipating in the EPA cross-check program, the EPA program code designationsfor the plant.]

The last sentence (bracketed information) is being deleted because the portion of the statement concerning summary of results to be included in the report is redundant to the statement in the previous sentence, and because it contains reference to the EPA cross-check program. Therefore, Section 7.1.2.3 will simply read as follows:

J Also to be includedin each report are thefollowing: a summary description of the REMP; a map (s) ofallsampling locations keyed to l a table giving distances and directionsfrom the centerpoint between j the Unit I and Unit 2 plant vent stacks; the results ofland use censuses required by Section 4.1.2; and the results oflicensee  ;

participation in the Interlaboratory Comparison Program required by Section 4.1.3.

(~h) b Included as part of this ODCM revision package is a correction to a typographical error in  !

the reference list on page viii. The NUREG reference in item I is incorrectly listed as NUREG-1033. The correct reference is NUREG-0133.

Since the proposed changes to the ODCM are administrative in nature and do not involve any changes to any plant system, structure, or component, or the manner in which they are operated, a determination has been made in accordance with Technical Specification 6.14.2 that the proposed changes will maintain the level of radioactive efIluent control required by 10 CFR 20.1302,40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

Unreviewed Safety Ouestion Determination The proposed changes to the ODCM to delete the reference to the EPA cross-check program and to correct a typographical error have been reviewed in accordance with 10 CFR 50.59, and the following determination has been made:

1. The proposed changes to the ODCM do not increase the probability of occurrence of

-, an accident previously evaluated in the FSAR because they are administrative in nature in that they delete a ICP service that will no longer be provided by the EPA,

FARLEY NUCLEAR P(ANT - 10 CFR 50.59 EVALUATION Peg) 6 of 7 RE: Of fsit3 Dos) C;lculation Manual, Revision 15

, and correct a typographical error. An acceptable alternative to the EPA cross-check l

(3) program will be implemented as currently allowed by the plant Technical l Specifications and the ODCM. Accordingly, the overall performance of any plant systems, structures, or components governed by the ODCM is unaffected by these proposed changes.

2. The proposed changes to the ODCM do not increase the consequences of an accident previously evaluated in the FSAR because they are administrative in nature and do not alter any of the conditions or assumptions in the FSAR accident analyses. Since the accident analyses remain bounding, the radiological consequences previously evaluated are not adversely affected by the proposed ODCM changes.

3. The proposed changes to the ODCM do not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FS AR because they are administrative in nature and do not affect the design or operation of any plant system, stmeture, or component important to safety.

4. The proposed changes to the ODCM do not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR because they are administrative in nature and do not affect the design or operation of any plant system, structure, or component important to safety and therefore do not adversely impact the radiological consequences presented in the FSAR accident analyses.

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5. The proposed changes to the ODCM do not create the possibility of an accident of a different type than any previously evaluated in the FSAR because they are administrative in nature and do not involve any change to the configuration or method of operation of any plant system, stmeture, or component. Accordingly, no new failure modes or new limiting single failures are created. A!so, there will be no change in types or increase in unounts of any effluents released offsite as a result of the ODCM changes.

4

6. The proposed changes to the ODCM do not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR because as stated previously, plant system, structure, or component performance will not be impacted since the ODCM changes are administrative in nature. Also, no new accident initiators or single failures have been identified as a result of the ODCM changes.

7. The proposed changes to the ODCM do not reduce the margin of safety as defined in the basis for any technical specification because an acceptable alternative to the EPA cross-check program will be implemented which satisfies the requirements of Regulatory Guide 4.15. It will continue to ensure that independen: enecks on the i precision and accuracy of the measurements of radioactive material in environmental l o samples are performed as part of the quality assurance program requirements for b

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. FARLEY NUCLEAR PLANT - 10 CFR 50.59 EVALUATION Pegt 7 of 7 l RE: Of fsita Dis) Calculation Manual, Rrsision 15

.rm environmental monitoring, and to demonstrate compliance with 10 CFR 50,

( Appendix L

Conclusion

Based on the preceding evaluation it can be concluded that the proposed changes to the j ODCM regarding deletion of the EPA cross-check program and correction of a j typographical error do not involve a change to the plant Technical Specifications, or an unreviewed safety question as defined by 10 CFR 50.59, and therefore can be implemented without prior approval from the NRC.

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{ Reference 1 NElletter dated August 10,1995 from John F. Schmitt to NEI Administrative Points of Contact O

N U C L E A R _E,N i t G Y IN Silt U T[

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y# a W Ja". %' " M d August 10,1995 9

TO: NEI Adminiatrative Points of Contact i

SUBJECT:

Termination of EPA Interlaboratory Comparison Program This letter is to advise you of the upcoming unavailability of the program many I licensees now use to satisfy their requirement to have the laboratory that analyzes their radiological environmental monitoring samples participate in an interlaboratory comparison program. Some alternative approaches to address this loss are discussed.

We recommend you provide this letter and enclosures to the individuals who manage your company's radiological environmental monitoring program for their action. Alternative intercomparison arrangements will be needed and f- in some cases a timely amendment to the technical specifications may be necessary.

4 The Environmental Protection Agency's National Exposure Research Laboratory in Las Vegas, Nevada has been conducting performance evaluation studies oflaboratories' capabilities to evaluate radionuclides in environmental mples for many years.

Licensees have been participating in this program to fdffS commitments to perform

! interlaboratory comparisons of environmental sample measurements, free of charge.

Many licensee technical specifications or Offsite Dose Calculation Manuals (ODCMs)

commit to participate in the EPA program. EPA notified NRC on May 18,1995 that they would no longer provide this service to NRC and its licensees as of September 1, 1995 (see enclosure 1). NEI wrote to EPA on June 26,1995 requesting EPA continue to provide services to NRC licensees through the end of the calendar year to permit licensees adequate time to establish alternative programs. EPA has agreed to extend the services until December 31,1995 (see enclosure 2).

Due to the unavailability of the EPA program, licensees with technical speciScations that specifically reference the EPA program without allowing for alternatives, i.e., do not state " EPA program or similar program," will need to make editorial modifications to their technical specifications. Other licensees with technical specifications that commit to the " EPA program or a similar program" or that simply commit to "an NRC approved program," will not need to make changes to their technical specifications. Licensees that have transferred environmental monitoring

-7 . 51.iiv so sun .co a. s-.~ c m s c acco m. .-o~e ::: y..ooo .. 2:2 res me 4,L .

l NEI Arbini=tr:tive Points cf Contact  !

! August 10,1995 Page 2 1-comnutments to the ODCM will need to determine if wording changes are needed. If j so, changes can be made without prior NRC approval.

l

, Licensees will likely have several options for the laboratory they utilize for environmental samples radionuclide analyses to participate in alternative e

interlaboratory comparison programs. Any third party organi= tion that has a j documented QA program and the capability to prepare QC materials in a manner that ensures traceability to the NationalInstitute of Standards and Technology (NIST) can j qualify. Options include, for example, radioactive source manufacturers, other utilities, ;

j or NIST. The nuclear industry already has a measurements assurance program j j through the NIST that could be expanded to function as an alternative interlaboratory ,

comparison program. A utility which employs a commercial environmental analysis  !

laboratory could amend its contract to stipulate the type and number of samples and  !

! performance requirements for successful participation in an interlaboratory comparison l l program. Efficiencies can be realized far those situations where several utilities use 3 the same contract laboratory. Those utilitic could get together and decide the total i i number ofinterlaboratory compansons and which nuclides and media should be l evaluated that year. By contrast, under EPA's program, each of these licensees was I sending all of the intercomparison samples they received from EPA to their contract i laboratory for evaluation. Analyzing several duplicates of the same EPA sample I

provided unnecessary redundant information regarding the laboratory's performance.

i NEI is discussing with NRC the timely availability of a clear definition of the latitude available to licensees in establishing parameters for interlaboratory analyses that meet j

the intent of commitments referencing the EPA program. NRC staff have stated they i

recognize the maturity oflicensees' programs for ensuring the quality of environmental sample measurements and plan to emphasize flembility in considering the objectives of alternative interlaboratory comparison programs established to fulfill commitments cc,ntained in the techical spacincations or ODCM. Enclosure 3, a letter from NEI to NRC, suggests an adequate program for NRC consideration and endorsement. We will provide NRC's response to our letter as soon as it is available. You may find this of benefit to guide you in deciding upon alternative interlaboratory comparison arrangements.

If we can be of assistance to you as you review this material, please do not hesitate to contact Lynnette Hendricks (202 739-8109) or me (202 739 8108).

Sincerely, '

John F. Schmitt '

Enclosures

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4 i.. ,g*,. ENCLOSURE 1

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i i UNITED STATa3 ENVIRONMENTAL. PROTECTION AoENCY jvn %. /

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, = n w..w .on .u m,. :: w a LAs vec As. =tvaoA .sssa use i.

k E I8 O l oumennuation ncsc Ancu ems ou 1

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! Charles L. Miller, Chief

} Daergency Preparedness and )

i Radiation Protection Branch j Division:of . Technical support 4 i

j U.S.. Nuclear Regulatory Commission )

11535 Rockville Pike, Mail Stop 09H15 i i

Rockville, Maryland 20855 i

j Dear Mr. M111ert

! on April 29, 1994, I wrote to Dr. Eric Beckjord to explain

! that, owing to increasing competition for dwindling resources, we s

would be forced to discontinue providing certain services to non-EPA clients, including the NRC. (For your convenience, I enclose i a copy of that letter).

i This would have included radionuclide j calibration standards and performance evaluation studies that we j

had been providing to NRC licensees and contractors. Presumably i

as a result of that letter, Tom Essig contacted i

NRC's desire to establish an interagency agreeme.us to express nt with our i

laboratory to. ensure the continued availability to NRC. That agreement would have established a mechanism for of our services j

j transferring funds from NHC to EPA with which we could finance

the services.,NRC needs, either through additional Federal staff i

or through a qualified contractor. Pending execution of such an agreement, we continued to distribute performance evaluation i- studies or standards to NRC licensees and contractors. However, on Thursday, April 13, Charles Ninson called Coorge Dilbeck, the l' i

Technical Lead for Radiation Quality Assurance, to tell him that NRC was no longer interested in entering into an agreement to t

cover these services.

i 1 i For the reasons I stated in my letter to Dr. Beckjord, we i

hereby NRC.

notify you of our intention to discontinue services to This is necessary because of staff reductions that severely

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3 14=4 our *hM (non-EPA) uy tas proulhyroducts or services to external organizations.

We are forced to redirect our limited Laboratory certification Program. such resources to satisfy EPA priorities as the Drinking Water i I

As of September 1, 1995, we j will no longer provide the following:

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1.

Performance evaluation studies for plutonium in I water, gross alpha and beta on air filters, and radionuclides in milk. . .

j 2. '

Services and materials previously provided to the HRC licenuses and contract laboratories, as detailed on the enclosed list. These include all'parformance j

evaluation and samples performance eva,luationradionuclide calibration standards' i

, study reports.

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I re i however, gret the need to end a long-standing service to NRC;

' I suspect you are fatailiar with the on-going reductions in Federal agency staffing that necessitate it. For your

{ possible use, I enciese a list of several dommercial vendors who j_

j may samples be able to supply that.you need. calibration standards and quality assurance 1

If you have any questions, please feel free to call ne at (702) 79s-2525,.or Dr. Dilbeck, at (702) 798-2104.

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Sincerely, i

W k +M&h N. Marichant Director -

Enclosures j cca Richard F.mch, NRC 4

Amira 4111, HRc -

i charles Rinson, NRC H. Eatthaw Bills, HERL Gary Foley, NERL i l

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=Arenunes.rna SUPPI.IERS l

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1. Amershas Corporation 312 593-6300 l

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! 2636 South Clearbrook Drive 800 313-6695 Arlington Heights, 1111 acts 60005 (U.S. Toll-Free)

2. Analycles, Inc.

404 351-8677 1380 Seaboard Industrial Boulevard

Atlanta, Georgia 30318 i

a

! 3. Du Pont NEN Research Products '800 551-1121 i 549 Albany Street.

Boston, Massachusetts 02118

! 4. Eberline Analytical Laboratory 505 345-3461 TNA/Eberline

. N.B.

) Albuquerque, New Mexico 7021 Pan American Eighway$7109

5. Isotope Products Laboratory 1800 North Keystone Street til 843-7000 Burbank, California 91504

6. National Institute of Standards and Technology 301 975-5531 .

. Radioactivity Croup L Building 245/C114 i l Gaithersburg, Maryland 20899 i

4 i

7. North American Scientific, Inc.

i 818 503-9201 1 7435 Greenbush Avenue '

l North Bollywood, California 91605-9823 i

i 8. U.S. Departseat of Energy 615 574-6984 1

1 c/o Martia Marietta Energy-8ystems, Inc.

Dek Ridge National Imboratory

} Isotope Distribution Office j P.O. Box 2009 Oak Ridge, Tennessee 37831-8044

9. Vestingbosse Banford Company Isotope Prograa Office 509 376-9619 F.O. som 1970 Mail stop 50-37 Richland, Vashington 99352 0

i ENCLOSURE 2 l i /% l 1-  !

1 3 {r O .)

(g/ UNITED STATES ENVIRONMENTAL PROTECTION AGENCY lO =?#!R'f#.'#!t!&a!!I'2?/#v"eiv

-veora::a' a sun-3 July 28,1995 enAnactanaenow nesaanen oms.o=

I j

Dear Colleagues:

i since IW9, the Environmental ".MT. Agency (EPA) and the U.S. Nuclear Regulatory l

Conunission 04RC) have been woddng jointly under a Memorandum of Understanding (MOU) that describes the wode the Radiation Quality Amw (RADQA) Program will  ;

pertoon for NRC. Under this MOU, an NRC licensees and watractors have been allowed to l participate in our Perfonnanos Evaluation (PE) Studies Program. 'this included access to att

{ PE studies and calibration standards that are provided to program participants.

i Por the past several years, funding for this wode dwindled. The EPA and the NRC discussed creating an Interagency Ap:  ; (IAG) through which NRC could provide funds to offset

{ some of thb fhading lost by the RADQA program. But in AprB of this year, the NRC  !

i decided it was no longer interested in providing ibnds through an IAQ mechanism. On M i 2

18 Wayne Marchant, the Director of the Charact*i~' Research Division of the National F-w_-e Research Latmi iy - Las Vegas, sent the attached letter to Mr. Chades Miller of l i the NRC.  !

Dr. Marchant's letter indicated our original intent was to eliminsee all support to NRC

! laborneories fmen our program by WM 1,1995. However, in respones to urgent appea j

from representatives of the Nuclear Energy Institute, we have agreed to delay this action un i December 31,1995. 'this should aBow time for cornpletion of contracts between the NRC  !

and the nuclear facilities that cod on that dats. 'this four-rnanth delay also wG1 allow the affected laboratories to find a suitable external quality assurance program, i

We have eg)oyed our amenaBy beneficial associadon with the NRC clients over the year However, our primary obligation must be to support EPA's regulatory mission

! accwidA.wra. We regret that .V=*-- in our funding and staKhave forced us to take this action. If we can be of any help in this transition process, please do not hesitate to

{ contact me at 702 798-2104, i

i k

George Dubeck Chemist Perfosmanos Evaluation Prograrn Radioenelysis Program (RSA RADQA)

s w. mas

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f ENCLOSURE 3 t

.Ch NUCLEAR ENERGY IN SiliU T E 4

John F. Schmitt, CHP couctom. m m neauno,.csMs.

E== .

1 August 10,1995 Mr. Charles L. Miller, Chief Emergency Preparedness and Radiation Protection Branch Division of Technical Support

Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 0001 i

Dear Mr. Miller.

i The Environmental Protection Agency (EPA) has notified the commercial nuclear O power industry ofits intent to discontinue providing performance evaluation studies in support of environmental monitoring programs conducted by the nuclear industry as of December 31,1995. I am writing to clarify the parameters of a program that will meet the intent of commitments made by licenaces in tech +al specifications that reference the " EPA program or equivalent" or that reference an NRC-approved program, or similar commitments made in Offsite Dose Calculation Manuals (ODCMs). To aid this clarification, questions and suggested responses are provided for your convenience in the enclosure.

A prompt reply would be very useful to licensees in establishing adequate interlaboratory comparison programs as alternatives to EPA's program. NEI will

communicate your reply promptly to our members.

Ifwe can be of any assistance, please do not hesitate to contact Lynnette Hendricks (202 739-8109) or me (202 739-8108).

, Sincerely,

,R, h L._ W__

John F. Schmitt p

C JFSlen Enclosures t 776 e startt. NW SUIT 4 800 W A$MiNO TON OC 20006-3708 PMONE 202.739 9000 FAN 202.78 5 40 t 9 4*

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i i O j V Clarification ofInterlaboratory Comparison Requirements

+

l. As licensees establish interlaboratory comparison alternatives to the

t terminated EPA program, what actions if any must licensees take to obtain l NRC approval of changes to wording regarding this requirement contained l in the technical specifications or ODCMT This answer assumes that the alternative comparison arrangements are generally in agreement *;.ith the clarifications provided later in this document.

i Licensees that have moved the environmental monitoring requirements from their i

technical specifications into their Offsite Dose Calculation Manual (ODCM)

}

pursuant to the guidance provided in Generic Letter 89-01 can make changes to their environmental monitoring programs, including changes that migh . < >

l appropriate in establishing alternative interlaboratory comparisons, by performing i a safety analysis and documenting that the changes do not result in an unreviewed safety question. A compilation of these changes are submitted to the NRC annually i for infa== tion.

4 i

Other licensees have chosen to retain environmental monitoring program commitments in their technical specifications. The wording in some of these j .

technical specifications regarding interlaboratory comparisons states the licensee I will participate in an " EPA program or similar" or "an NRC approved program."

Clarifications contained in this letter regarding alternatives to the EPA program, j

t once agreed to by NRC, constitute a "similar" program to EPA per wording in technical specifications, and therefore do not require changes to techniaal specifications and do not require further NRC approval.

However, if wording in the technical specification calls out the EPA program j without qualification, i.e., the wording does not state "or simdar program," an editorial amendment to the technical specifications would be required. Because

! these changes are editorialin nature, licensees would not be considered to be out of j compliance if a technical specification amendment request had been submitted to i

NRC by the beginning of the year when the EPA program is no longer extended to NRC licensees.

Changes made in the interlaboratory comparisons pursuant to these clarifications j

do not constitute significant departures from the original"NRC approved" program i specified by the tehnical specifications or the ODCM. Therefore, NRC approvalis i not needed to comply with the requirement for an "NRC approved" program.

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i What objective should guide licensees' efforts to respond to termination of EPA's interlaboratory comparison program?

j

' Section 6.3.2 Interlaboratory Analyses, of Reg. Guide 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the

{ Environment," states interlaboratory measurements provide a means to detect i errors that might not be detected by intralaboratory checks alone. The intent is to I have the utility participate in, or have their motract laboratory participate in, a third party single blind testing program with demonstrated traceability to NIST.

i This participation will help to ensure that no unobserved systematic errors or i biases exist in the laboratory's internal QC program of such a magnitude to lead to

! unacceptable errors in environmental measurements. Licensees should use this opportunity to thoughtfully evaluate the appropriate role ofinterlaboratory comparisons and design a program that is most effective and meaningful for their situation. A licensee employing an analysis program that has shown consistently l good performance and that is not undergoing major changes may be able to scale j

back from the EPA program (see answer to the final question below regarding i number of samples and media) focusing on radionuclides and media that are most j relevant to ensuring adequate performance of their environmental analysis l laboratory. There is no need to duplicate the EPA interlaboratory comparison i program.

Who can conduct an interlaboratory comparison?

Any organization that has a documented QA program and the capability to prepare j QC materials in a manner that ensures traceability to the National Institute on j Standards and Technology (NIST). This organization must provide samples to i

environmental analysis laboratories without identifying the concentrations of j specific nuclides. Utilities can obtain interlaboratory comparisons services from i

' any organization meeting these requirements, e.g., from radioactive source manufacturers, other utilities or directly from NIST.

What is the minimum number ofsamples and media that must be

} evaluated, and at what frequency?

4 r Evaluation of three samples per year, ona each of water, milk and particulate l

! filters, containing nuclides representative of those occurring in the licensee's  !

! effluent mod environmental samples could be adequate to verify that a laboratory is i l capable of continuing to make measurements that are within acceptable accuracy.

For laboratories that are just starting up or laboratories that are experiencing l j difficulties in meeting internal quality control goals for certain nuclides or media, a l licensee should expand .the scope of the interlaboratory comparison program in order to assist in addressing sources of error and to provide information regarding

. the laboratory's performance.

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l Reference 3 i l 3 NRC letter dated 1 i

j December 26,1995 l from Charles L. Miller to John F. Schmitt at NEI l

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g UNITED STATES i e NUCLEAR REGULATORY COMMIS810N i 5 wasmorou. o.c. sues.amn l

\,,..... December 26, 1995 I Mr. John F. Schmitt, CHP Director, Radiological Protection, Emergency Preparedness, i and Waste Regulation l Nuclear Energy Institute

! 1776 I Street, NW, Suite 400 '

i Washington, DC 20006-3708

Dear Mr. Schmitt:

! I am responding to your letter of August 10, 1995, in which you requested that j the Emergency 3reparedaess and Radiation Protectier. Branch clarify the

parameters of a program that meets the requirements of an interlaboratory

' comparison program that is equivalent to the one that has been conducted by the Environmental Protection Agency (EPA) Program. In your letter, you also

presented a position on what constitutes an acceptable minimum radiological l monitoring program. I believe it is premature to address the contents of an

acceptable minimum program until we have had the opportunity to evaluate the

! status of licensee programs during the next year of transition from the EPA to alternate suppliers of environmental standards.

iO j

The a " interlaboratory t> a r c < comparison r a diai -

Effluent Streams and the Environment." The gu pr3ran a is discussed 4*

interlaboratory comparison program is to be conducted with the " EPA's r4 'ide '- in Regulatory gr - (a i aaGuide r *' 4.15, ndicates that the i Environmental Radioactivity Laboratory Intercomparison Studies (Cross-check) i Program, or an equivalent program." The program requires the environmental laboratory of the licensee, or the licensee's contractor, to participate in a j third-party blind testing program. Thus, it has always been the licensee's

choice as to which laboratory (EPA or commercial) is used for the program.

! Nevertheless, regardless of which laboratory is used, an outside laboratory l j provides a means for independent checks on the precision and accuracy of the l

! measurements in the environmental monitoring program and can detect errors

that might not be detected by intralaboratory measurements alone.

l l In lieu of using the EPA, licensees must continue to maintain their existing interlaboratory comparison program, which includes all of the determinations (sample medium /radionuclide combinations) that were offered by the EPA and included.in the licensee's environmental monitoring program, as well as the l

frequency of the sample determinations. The independent laboratory and/or its measurement should be traceable to the National Institute of Standards and j Technology (NIST). Also, in order to be independent and avoid the possibility 4 of introducing unobserved systematic errors or biases, the laboratory should j not be the same one used for the environmental monitoring program.

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4 Decaster 26, 1995

! Mr. J. F. Schmitt i '

! In conclusion, we view the transition away from the EPA in this program as a i change of contractor, with no impact on the overall quality of the program '

! and one that does not constitute a departure from the original program that

' was approved by the Nuclear Regulatory Commission. Licensees are expected to i continue to maintain the same high-quality program they maintained with the I services of the ERA. -

! If you have any questions on this matter, please call Stephen Klementowicz at  ;

(301) 415-1084.

t l

l Sincerely, 0 -

, Charles L. Miller, Chief l Emergency Preparedness and Radiation Protection Branch 5 Division of Reactor Program Management j Office of Nuclear Reactor Regulation l

cc

J. Wiggins, RI i C. Hehl, RI i B. Mallett, RII i A. Gibson, RII

! G. Grant, RIII l C. Pederson, RIII R. Scarano, RIV l

i T. Gwynn, RIV i

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! Regulatory Guide 4.15 l Revision 1, February 1979 i

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2 rano R. vision 1 N o 7 wx U.S. NUCLEAR REGULATORY COMMISSION February 1979 QW,;i

n. . REGULATORY GUIDE OFFICE OF STANDARDS DEVFi '* MENT '}.'~C? I' ,

?

1 REGULATORY GUIDE 4.15 QUALITY ASSURANCE FOR RADIOLOGICAL MONITORING PROGRAMS i (NORMAL OPERATIONS) - EFFLUENT STREAMS AND THE ENVIRONMENT I 1

A. INTRODUCTION Part 20 as is reasonably achievable, taking into account the state of technology and the This guide describes a method acceptable to economics of improvements in relation to public the NRC staff for designing a program to as- health and safety and to the utilization of sure the quality of the results of measurements atomic energy in the public interest, of radioactive materials in the effluents and the i environment cutside of nuclear facilities during Section 30.34, " Terms and Conditions of Li-

normal operations, censes," of 10 CFR Part 30, " Rules of General Applicability to Licensing of Byproduct Mate-The NRC regulations that require the control rial," provides that the Commission may incor-of releases of radioactive materials from nuclear porate in any byproduct material license such facilities, that require the measurements of ra- terms and conditions as it deems appropriate or dioactive materials in the effluents and envi- necessary in order to protect health.

• ronment outside of these facilities, that require quality assurance programs and establish Section 40.41, " Terms 2nd Conditions of Li-quality assurance requirements for certain fa- censes," of 10 CFR Part 40, " Licensing of

' cilities, or that authorize license conditions not Source Material," provides that the Commission 1 4

otherwise authorized in the regulations are as may incorporate in any source material license

(,

4 follows: such terms and conditions as it deems appro- )

priate or necessary to protect health.

Section 20.106, " Radioactivity in Effluents to '

Unrestricted Areas ," of 10 CFR Part 20, Section 50.50, " Issuance of Licenses and

" Standards for Protection Against Radiation," Construction Permits," of 10 CFR Part 50, "Li-  :

(

provides that a licensee shall not release to an censing of Production and Utilization Facil-unrestricted area radioactive materials in con- l ities," provides that each operating license for I centrations that exceed limits specified in a nuclear power plant issued by the Nuclear 10 CFR Part 20 or as otherwise authorized in a Regulatory Commission will contain such condi-license issued by the Commission. Section tions and limitations as the Commission deems 20.201, " Surveys," of 10 CFR Part 20 further appropriate and necessary.

I

' requires that a licensee conduct surveys, including measurements of levels of radiation or Section 70.32, " Conditions of Licenses," of concentrations of radioactive materials , as 10 CFR Part 70, "Special Nuclear Material,"

necessary to demonstrate compliance with the provides that the Comniission may incorporate l

, regulations in 10 CFR Part 20. such terms and conditions as it deems appro-priate or necessary to protect health.

' Paragraph (c) of Section 20.1, " Purpose," of 10 CFR Part 20 states that every reasonable Section IV.B of Appendix I, " Numerical  ;

effort should be made by NRC licensees to Guides for Design Objectives and Limiting Con- i maintain radiation exposure, and releases of ditions for Operation to Meet the Criterion 'As radioactive materials in effluents to unrestrict. Low As is Reasonably Achievable' for Radio-ed areas, as far below the limits specified in active Material in Light-Water-Cooled Nuclear- 1 s

Power Reactor Effluents," to 10 CFR Part 50, l

" Licensing of Production and Utilization Facil-in.. ineesi. .ubstanuve enanses from previous :s.u.. ities " requires that licensees establish an ap-a USNRC REGULATORY CUlOES j * " , ' ' , " , " *'Q "Q Cf;;,,,,,j $8,,, ,,,

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[] propriate surveillance and monitoring program Mills ," all give some guidance on means for

( / to provide data on quantities of radioactive assuring the quality of the measurements of ra- l

" material released in liquid and gaseous efflu- dioactive materials in effluents and the envi- I ents and to provide data on measuruble levels ronment outside of nuclear facilities. More com- I of radiation and radioactive materials in the plete and extensive guidance on this subject is l environment.Section III.B of Appendix ! to provided in this document for nuclear power l 10 CFR Part 50 provides certain effluent anti reactor tscilities and for other facilities for environmental monitoring requirements with which radiological monitoring is required by respect to radioactive iodine if estimates of the NRC. This guidance does not identify exposure are made on the basis of existing separately the activities that are within the i conditions and if potential changes in land and scope of Appendix B to 10 CFR Part 50. How- l water usage and food pathways could result in ever, this guidance is intended to be consist-exposures in excess of the guidelines of ent with the requirements of Appendizes A and Appendix I to 10 CFR Part 50. B to 10 CFR Part 50 in that quality asw.'nce requirements should be cor.sistent 9ith the i General Design Criterion 60, " Control of re- importance cf the activity. For the wonitoring I leases of radioactive materials to the environ- of production and utilization facihties that is l ment," of Appendix A, " General Design Cri- within the scope of Appendix B to 10 CFR l teria for Nuclear Power Plants," to 10 CFR Part 50, other regulatory guides that provide I Part 50 requires that nuclear power plant guidance on meeting the quality assurance re- I designs provide means to control suitably the quirements of Appendix B to 10 CFR Part 50 release of radioactive materials in gaseous and should also be consulted. .-

liquid effluents. General Design Criterion 64,

" Monitoring radioactivity releases," of Appen- B. DISCUSSION 1 dix A to 10 CFR Part 50 requires that nuclear power plant designs provide means for monitor- As used in the context of this guide, quality ing effluent discharge paths and the plant assurasee comprises all those planned and sys-environs for radioactivity that may be released tematic actions that are necessary to previde from normal operations, including anticipated adequate confidence in the results of a mon-operational occurrences, and from postulated itoring program, and quality control comprises accidents . those quality assurance actions that provide a means to control and measure the character-O General Design Criterion 1, " Quality stand- istics of measurement equipment and processes to established requirements; therefore, quality

(#/

ards and records " of Appendix A to 10 CFR Part 50 requires that a quality assurance pro- assurance includes qthality control.

gram be established for those structures, sys-tems, and components of a nuclear power plant- To assure that radiological monitoring meas-that are important to safety in order to provide urements are reasonably valid, organizations adequate assurance that they will satisfactorily performing these measurements have found it perform their safety functions, necessary to establish quality assurance pro-grams. These programs are needed for the Appendix B, " Quality Assurance Critena for following reasons: (1) to identify deficiencies Nuclear Power Plants and Fuel Reprocessing in the sampling and measurement processes to Plants," to 10 CFR Part 50 establishes quality those responsible for these operations so that assurance- requirements for the design, con- corrective actiori can. be taken , and (2) to l struction , and operation of those structures, obtain some r 2asure of confidence in the systems, and components of these facilities that results of the monitoring programs in order to prevent or mitigate the consequences of postu- assure the regulatory agencies and the public lated accidents that could cause undue risk to that the results are valid, the health and safety of the public.

l The need of quality assurance is implicit inityExisting published guidance on specific qual-assurance actions that are spphcable to ra-all requirements for effluent and environmental diological monitoring is limited and, in ger.eral, moniterin g, and this need has been widely is restricted to quality control practices for recognized. Regulatory Guide 1.21. "Measur- radioanalytical laboratories (Refs.1-5). How-i ing, Evaluating, and Reporting Radioactivity in ever, quality assurance should be applied to all Solid Wastes and Releases of Radioactive Mate- steps of the monitoring process, which may rials in Liquid and Gaseous Effluents from include sampling, shipment of samples, receipt Light-Water-Cooled Nuclear Power Plants ;" of samples in the laboratory, preparat. ion of Regulatory Guide 4.1, " Programs for Monitor- samples , measurement of radioactivity , data ing Radioactivity in the Environs of Nuclear reduction , date evaluation , and reporting of Power Plants;" Regulatory Guide 4.8, "Envi* the monitoring results.

ronmental Technical Specifications for Nuclear Power Plants ;" and Regulatory Guide 4.14.

[-%} " Measuring, Evaluating, and Reporting Radio-

\ / activity in Releases of Radioactive Material in *Denniuons or specal tes ned in this swas are given in

~

Liquid and Airborne Effluents From Uranium the siossary on page o 15-W 4.15-2

i i

3 i l C  :

I f The scope of this guide is limited to the ele- 2. Speafication of Quahtications of Personnel  ;

ments of a quahty assurance program, which is j a planned, systematic, and documented pro- The qualifications of individuals performing ,

1 gram that includes quauty control. Guidance on radiological monitoring to carry out their j . principles and good practice in the monitoring assigned functions should be specified and <

l process itself and guidance on activities that documented (e.g. , as in a job description). .I  !

j can affect the quauty of the monitoring results t (e.g. , design of facilities and equipment) are An indoctrination and orientation program,

outside the scope of this guide. However, some appropriate to the size and complexity of the references are provided to documents that do

. organisation and to the activities performed, I '

provide some guidance in these areas. The r.hould provide that (a) personnel performing i citation of these references does not constitute quality-related activities are trained and quali- l an endorsement of all of the guidance in these fled in the principles and techniques of the ac-  !

documents by the NRC staff. Rather, these tivities performed, (b) personnel are made references are provided as sources of inforssa- aware of the nature and goals of the quality  !

tion to aid the licensee and the licensee's con- assurance program, and (c) proficiency of per- l tractors in developing and maintaining a sonnel who perform activities affecting quauty '

monitoring program. is maintained by retraining, reezamining, and I recertifying or by periodic performance re-  !

Every organhation actually performing efflu- views, as appropriate.  !

ent and environmental monitoring, whether an ,

NRC licensee or the licensee's contractor, 3. Opssating Precedures and lastructions '

should' include the quality assurance program elements presented in this guide. Written procedures should be prepared, re-  ;

viewed,'and approved for activities involved in J carrytag out the monitoring program, including  :

C. REGut.ATORY POSITION sample collection; packaging, shipment, and '

receipt of samples for offsite analysist prepa-The . quality assurance program of och ration and analysis of samples; maintenance, organization performing effluent or environ- storage, and use of radioactivity reference  ;

u. ental monitoring of nuclear facilities for nor- standards; calibration and checks of radiation mal operations should be documented by and radioactivity measurement systems; and written policies and procedures and records. reduction, evaluation, and reporting of data.

These documents should include the elements Individuals who review and approve these pro- '

given in this section. cedures should be knowledgeable in the sub-

. jects of the procedures. ,

In addition to its own program, a licensee ,

should require any contractor or subcontractor Guidance on principles and good practice in performing monitoring. activities for the 11 - many of these activities is presented in NRC

censee to provide a quality assurance program regulatory guides (Refs. 6-9) and other publi-and to routinely provide program data cations (Refs. 2-5,10-35). In addition to these summaries (sufficiently detailed to permit on- publications, Scientific Committee 18A of the going quality assurance program evaluation by NCRP has prepared NCRP Report 58, "A Hand-the licensee) consistent with the provisions of book of Radioactivity Measurements Proce-this guide, as follows: dures," (Ref. 36) that is a revision of NCRP Report 28 NBS Handbook 80, "A Manual of J

1. Organizational Structwo and Responsibilities Radioactivity Procedures,"  ;

of Managerial and Operational Personnel J

4. Records The structure of the organization as it re-lates to the management and operation of the The records necessary to document the ac-monitoring program (s), including quality as- tivities performed in the monitoring program surance policy and functions, should be pre- should be specified in the quality assurance seated. The authorities, duties, and responal- program. 4 bilities of the positions within this organisation '

down to the first-line supervisory level should One key aspect of quality controlls maintain-be described. . This should include responsibil- ing the ability to track and control a sample in itles for review and approval of written proce- its progress through the sequence of monitor-dures and for ae preparation, review, and ing processes. Records to accomplish this evaluation of monitoring data and reports. should cover the following processes: field and inplant collection of samples for subsequent analysis, including sample description; sample Persons and organizations performing quality receipt and laboratory identification coding; assurance functions should have - sufficient sample preparation and radiochemical process- l authority and crgannational freedom to identify ing (e.g. , !sboratory notebooksh radioactivity quality problems; to initiate, rer-- H . or measurements of ' samples, instrument back-provide solutions; and to verify implementation grounds , and analytical blanks; and data of solutions, reduction and verification.

4.15-3

. , , - - r -. m -

Quality crntrol records fer labor: tory ssvsre problems his bien encountsrsd with c:unting syst ms should includa ths risults cf equeous samplis cf rtdioretiva w:stss from musurements of radioactive check sources, operating nuclear reactors (Ref. 23).

i calibration sources, backgrounds, and blanks.

O R: cords relating to overall. laboratory per- Guidance on the principles and practice of Qarmance should include the results of analysis sampling in envut.N'lental monitoring is pro-of quality control samples such as analytical vided in seven* ,,uolications (Ref* 2, 4, 5, 21, 23-31, 33, '<5, 37). In addition, workers at blanks, duplicates, interlaboratory cross-check 1 samples and other quality control analyses; use the National E ureau of Standards (NBS) have ,

of standards (radioactivity) to prepare working published the a %ults of a survey of infortna- I str.ndards; preparation and standardization of tion on sampling, sample handling, and long-carrier solutions; and calibration of analytical term storage for environmental materials (Ref.15). Scme guidance on the principles and balances .

practice of air sampling is provided in Refer- i ences 17, 19, 24, 28-31, 33. Guidance on the Additional records that are needed should in- principles and practice of water sampling is clude the calibration of inline radiation detec- provided in numerous publications (Refs.13, i

tion equipment, air samplers , and thermo- 14, 25-27, 35, 37). l luminescence dosimetry systems; verification and documentation of computer programs; 6. Quality Cor.trolin the Radionnalytical Laboratory qualifications of personnel; and results of tudits.

6.1 Radionuclide Reference Standards-Use for The minimum period of retention of the rec- " " * " ***'*"*"I'""

crds should be specified. For nuclear power Reference standards are used to deterinine plants, requirements for record retention are counting efficiencies for specific radionuclidas givcn in the plant technical specifications. In or, in the case of gamma-ray-spectrometry giniral, for other types of facilities, only the systems, to determine countin final results of the monitoring prograa.s need function of gamma-ray energy.gA efficiency.ascounting effi-a b3 retained for the life of the facility. ciency value is used to convert a sample count-ing rate to the decay rate of a radionuclide or

5. Quality Control in Sampling (including Pack-aging, Shipping and Storage of Samples) to a radionuclide concentration. Guidance on the calibration and usage of germanium de-tectors for measurement of gamma-ray emission Continuous sampling of liquids and gases in- rates of radionuclides has been prepared as an v:lves the measurement of sample flow rates ANSI standard (Ref. 38). For converting and/or sample volumes. The accuracy of the gamma-ray emission rates to nuclear decay

vices used for this purpose should be deter- rates, two reports from the Oak Ridge National id on a regularly scheduled basis, and ad- Laboratory (Refs. 39 and 40) provide useful ustments should be made as needed to bring l compilations of gamma-ray intensities and other tha perforinance of the devices within specified nuclear decay data for radionuclides in routine limits. The results of these calibrations should releases from nuclear fuel cycle facilities. The )

ba recorded. The frequency of these calibra- data from Roterence 40 are included in NCRP tiins should be specified and should be based Report 58 (Ref. 36), i i

on the required accuracy, purpose, degree of usrge, stability characteristics, and other con- Radionuclide standards that have been ditions affecting the measurement. Procedures certified by NBS or standards that have been for continuous sampling should use methods obtained from suppliers who participate in thtt are designed to ensure that the sample is measurement assurance activities with NBS8 rspresentative of the material volumes sampled.

The collection efficiencies of the samplers used ,,,,,,

should be documented; usually such documen- ,,,, ,,,,,, ,, , ,, , ,, , ,,,,,

.oure. suppners and Nas involve two basa m--a: (t) tation is available from manufacturers of the The suppuer submits a canbreted radaoactivstr source (prof-sampling equipment. erably selected from a batch or prepared eetles of eeurces) to NBs for connrmauon that the suppuer's eaubrauon vahne agrees with NBs results withh certain specified lisaits or (2)

Procedures for grab samples should include was proetdes eaubrited radioacts, sty sources et undiscia.ed stips designed to ensure that the sample is **",$,0*,l,*"[,",3

, ,",,* '",,'Pj' ",b',0,b'* ' ** * ** D rspresentative of the material sampled. Rep- ut st. it certain sp.cified u=>u with the measurements of NBs. For the licete grab samples should be taken periodically rounne produccan of cn== rcas redaneeunty standards, too to determine the reproducibility of sampling. fzst meebanism is preternbie to the second but is not always g,,,,bte. Then two mechanaams an used both in Measurements Assurance Programs (MAPS) with key Laboratories and in other Procedures for sampling, packaging, ship. musurement anurance actinnes.

ping, and storage of samples should be .fwo key ubanary source suppuen parcespete in MAPS with da:igned to maintain the integrity of the sample NBs and use both of the two basic mechanisms: (1) The NaC from time of Collection to time of analysis. retenace labonton for me Connroaten Men unnents 6 Aqueous samples may present a particular gram (for effluent monitormg) of the NaC Offace of Inspeeden and Entorcement and (2) The EPA EnvuSnmental Monatoring problem in this regard, and one of the most and support Laborsiorr in Las veras. wtuch prepares and 4.15-4

' R, h'

t should be heed when such standards are avail. investigative and corrective action should be able. In these measurement assurance activ- taken when the measurement value falls outside ities , the supplier's calibration value should the predetermined control value.

agree - with the NBS value within the overall j uncertainty stated by the supplier in its certi. A check source for determining changes in l fication of the same batch of sources (when counting rate or counting efficiency should be I these are sampled for measurement by NBS) or of sufficient radiochemical purity to allow l in its certification of sinuar sources, correction for decay but need not have an l accurately known disintegration rate, i.e., I An " International Directory of Certified Ra- need not be a standard source. I

dioactive Materials" has been published by the I International Atomic Energy Agency (Ref. 41). For systems in which samples are changed i i manually, check sources are usually measured' I daily. For systems with automatic sample i- Acceptable standards for certain natural re.

j dionuclides may be prepared fine commercially changers, it may be more convenient to include available high-purity chemicals. For example, the check source within each batch of samples and thus obtain a measurement of this source i potassium-40 standards for gross beta-particle j measurements or gamma-ray spectrometry may within each counting cycle. For proportional

. be prepared gravinetrically from dried counter systems, the plateau (s) or response (s) i

reagent-grade potassium chloride, to the check source (s) should be' checked after l 1 each gas change. Background measureusents

' Th'e details of the preparation of working should be made frequently, usually daily or stand ards from certified standard solutions before each use, to ensure that levels are

within the expected range. For systems with shotCd be recorded. The working standard should be prepared in the same form as the un- automatic sample changers, background meas-known samples, or close approximation thereto. urements should be included within each meas- 1 urement cycle. l Efficiency calibrations should be - checked periodically (typically monthly to yearly) with For alpha- and gamma-ray-spectrometry sys-tems , energy-calibration sources (i.e . , a g

5 standard sources. In addition, these checks should be made whenever the need is in- source containing a radionuclide, or mixture of .  ;

dicated, such as when a significant change in radionuclides, emitting two or more alpha or the measurement system is detected by routine - gamma rays of known energies) are counted ta .!

measurements with a check source. determine the relationship between channel number and alpha- or gamma-ray energy. The 6.2 Performance Checks of Radiation Messmement frequency of these energy calibration checks - 1 Systems depends on the stability of the system but usually is in the range of daily to weekly. The Deterunination of the background counting results of these measurements should - be rate and the response of each radiation detec- recorded and compared to predetermined limits tion system to appropriate check sources in order to determine whether or not system should be performed on a scheduled basis for gain and zero level need adjustment.

systems in routine use. The results of these Adjustments should be made as necessary. '

l measurements should be recorded in a log and  !

plotted on a control chart. Appropriate Additional checks needed - for spectrometry systems are the energy resolution of the distrmtes caubnted redeneeuesty standards primarcy te lab- system and the count rate (or counting enterw. ineoived in naisissient on. utal meauerime-Additionally. seven maler todaspharmaceutacal manufacturers efficiency) of a check source. These should be '

(sono of widen suppir rodienetivity standards esemerusaar) determined periodically (usually weekly to partackpate un a MAP ergamased by the Atemic tadustrial Forum monthly for energy resolution and daily to and mss. In thas MAP. NBs distributes standards as test one= wegk]y for Count rate) and after system

" " changes, such as power failures or repairs, to '

1. $fd u s$'piu fYa"'Nactu~Y~2e,* MQ $

r determine if there has been any significant (first mecha use).

change in the system. The results of these Meuurement amurance inuractions that use e first me- measurements should be recorded.

emanase an avaiinnie e.a spacint was cabbretsen utveces. Nas will. en request and for a fee. perform calabrotions of repre-sentenve samples of standards provided by the supplier for 6.3 Anslysis of Qualify Control Samples NBs confirmatana of the supplier's reported values. CaMbratise services are arealable for a large variety of redsenuchdes pro-need certaan requirementa tea se seenpie stamanity and sustatie The analysis of quality control samples pro-acuvity rense) are met. Meuvrement usurance internetaene vides a means to determine the precision and enat un m neced seemanism an weissone na the usuance et accuracy of the monitoring processes and in-test studards er Nas. For a nomanal charee (beyond the price et v.e sandard) . ss neasseunty lan.iard Reference cludes both intralsboratory and interlaboratory i Matenals ($RMs> na se purenaeed as test sources of undas- measuremerits.

elesed actenty that can be used to demonstrate agressment.

withan certaan specified Maats, between the seured supF6er's measurements and those of NSs. A Report of Test (for the first mechanum) er a Repr of Mueurement (acend mechamaer,. The analysis of replicate samples (containingt 6** * ***"* suppaier's and was vanues is imaued significant detectable activity) provide.s al

• • • • • ""s"8 by NB to document the source supplaer's perucapataan ta the -

meuurement usurance acanty. means to determine precision; the analysts of 4.15-5 ,

J

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ii 'known concentrations of should be included frequently in groups of un '

. ,.4amples conta n ngprovides a means to determine known environmental samples that are analyzed

,dionuclides radiochemicany. Spiked and blank samples l

,2 curacy. The analysis of laboratory blanks should be submitted for analysis as unknowns l provides a means to detect and measure to provide an intrataboratory basis for estimat- I

- radioactive contamination of analytical samples,

!- ing the accuracy of the analytical results.

l a ' common source ' of error - in radiochemical These blanks .and spikes may include blind

- analysis of low-level samples. The analysis of replicates.  ;

• antlytical blanks also provides information on l l

the adequacy of background subtraction , 6.3.2 Intertsboratory Analyses j- particularly for environmental samples.

l The fraction of the analytical effort needed Analysis of effluent and environmental sam-

}

far the analysis of quauty control samples ples split with one or more independent labors-t diptnds to a large extent on (1) the mixture of tories is an important part of the quality as-

[ surance program because it provides a means

! sample types in a particular laboratory 'in a to detect errors that might not be detected by

! p2rticular. time period and (2) the history of intralaboratory measurements alone. When

( parformance of that laboratory in the analysis possible, these independent iaboratories should  !

j. tef quality control samples. However, . for be those whose measurements are traceable to  !

l anvironmental laboratories, it is found that at NBS .8 l

le:st 5%, and typicauy 10%, of the analytical lord should consist of quauty cot 4 trol samples.

l- Analysis of split flekt samples, such' as sam-6.3.1 Intinisboratory Analyses pies of milk, water, soil or sediment, and i

! vegetation, is particularly insportant in envi-j ' Riplicate samples, usually duplicates,'should roamental monitoring , programs to provide an l ba analysed routinely. These replicates should independent test of .the ability to measure j radionuclides at the very low concentrations i ba prepared from samples that are as homo- present in most enetroamental samples.

i

,.~

ganeous as' possible, such as well-stirred or i cixad . liquids - (water or milk) and solids The NRC Office of Inspection and Enforce - 1

(dried, ground, or screened sou, sediment, or ment conducts a Confir matory Measurements  !

J' vrgetation; or' the ash of these materials). ' Program for laboratories of licensees that mens- I l

Thsse samples may be replicates of monitoring ure nuclear reactor effhaents. The analyses of program samples, replicates of reference test

-terials, or both. The size and other physical liquid waste holdup tank samples, gas sesspies, l

d chemical characteristies of ,the replicate charcoal cartridges, and stack particulate i i filters are included in this program. ~The

! samples should be similar to those of single - results of the licensee's measureunents of sam-2 samples analyzed routinely. ples split with the NRC are compared to those l' The analysis of the replicate samples as blind of an NRC reference laboratory whose measure -

j ments are traceable to the National Bureau of replicates is desirable but is not practicable for Standards. Thus the results of this coenparison l- au laboratories or for au types of samples..For provide to the NRC an objective measure of the example : in - small laboratories it may not be accuracy of the licensee's analyses, i practicable to prevent tha_ analysts from being i , awzre that particular samples are replicates of Laboratories of licensees or their contractors f ens another. environmental measurements i Obtaining true replicates of au types of that- perform should participate in the EPA's Environmental Radioactivity Laboratory Intercomparison Stud-p

samples also is not practicable. For example, les (Cross-check) Program, or an equivalant 1  : cbtaining. replicate' samples of airborne mate. Pforram. This participation should include all rials usuauy is not practicable on a routine of the determinations (sample medium /radionu-btsis because it requires either a separate clide combinations) that are both offered by x stapling system or splitting a single sample EPA and included in the licensee's environ-1 (e.g... cutting a filter in half). Use of replicate mental monitoring program. Participation in the I. samplers usually is not econentically feasible EPA program provides an objective measure of I and splitting of samples results in replicates the accuracy of the analyses because the EPA j

7 that do not represent the usual sample size or measureusents are traceable to the National ~

nessurement configuration (counting geometry) 3 for direct measurement. However, simulated Bureau of Standards. If the mean result of a

cross-check analysis exceeds the control limit j eamples of airborne materials may be prepared as defined by EPA (Ref. 42), an investigation 4- in replicate and submitted for analysis as - should be made to determine the reason for this unknowns. deviation and corrective action should be taken E

' Analysis of intralaboratory blank and spiked

• was and wac iarre r s is m. n 4 r., . ca r.c moned-l samples -is. an. important. part of each environ-I mental laboratory's quality control program. To . osa et ene eers irmummmer u a aeekee t redinues and

} check for contamination from reagents and gg,e gy,wy,pp ,,,,,

Iother sources, known analytical blank samples M,.rul.gr

,, iy 7,u t 4.15-6

1 as necessary. Similarly, an investigation and concentrations and/or release rates of radio-

, any necessary corrective action should take active material in the monitored release path.

. place if the " normalized range," as calculated These correlations should be based on the 4

. by EPA, exceeds the control limit, as defined results of analyses for specific radionuclides in by EPA. A series of results that is within the grab samples from the release path.

j control limits but that exhibits a trend toward

these limits may indicate a need for an investi. Any flow-rate measuring devices associated l gation to determine the reason for the trend. with the system should be calibrated to deter-mine actual flow rates at the conditions of

, 6.4 ComputationalChecks temperature and pressure under which the i

l system win be operated. These' flow rate

' Procedures for the computation of the con. devices should be recalibrated periodically.

centration of radioactive materials should in-

. clude the independent verification of a sub. Whenever practicable, a check source that is j~ stantial fraction of the results of the computa. actuated remotely should be installed for in-l tion by a person other than the one performing territy checks of the detector and the asso-

the original computation. For computer calcula. ciated electrical system.

4 tions, the input data should be verified by a i knowledgeable individual. All computer pro. 8. Review and Analysis of Data grama should be documented and verified before initial routine use and after each modifi. Procedures for review and analysis of data cation of the program. The verification process should;be developed. These procedures should should include verification, by a knowledgeable cover examination of data from actual samples individual, of the algorithm used and test runs and from quality-control activities- for reason-in which the output of the computer computa. ableness and consistency. These reviews tion for given input can be compared to "true a should be performed on a timely basis. General values that are known or determined independ. criteria for recognizing deficiencies in data ently of the computer calculation. Documenta- should be established, tion of the program should include a descrip-l tien of the algorithm and, if possible, a Provisions should be made for investigation current listing of the program. Guidelines for and correction of recognized deficiencies and the documentation of digital computer programs for documentation of these actions, are given in ANSI N413-1974 (Ref. 43).

9. Audits

7. Quality Control for Continuous Effluent Monitoring Systems Planned and periodic audits should be made to verify implementation of the quality assur-Guidance on specification and performance of ance program. The audits should be performed onsite instrumentation for continuously mon. by individuals qualified in radiochemistry and itoring radioactivity in effluents is given in monitoring teche.! ques who do not have direct ANSI N13.10-1974 (Ref.18). responsibilities in the areas being audited.

The specified frequency of calibration for a Audit results should be documented and re-particular system should be based on con. viewed by management having responsibility in siderations of the nature and stability of that the area audited. Followup action, including system. For nuclear power plants, specific re. reaudit of deficient areas, should be taken quirements for cahbrations and checks of par. where indicated.

ticular effluent monitoring systems usually are included in the technical specifications for the D. IMPLEMENTATION plant.

The purpose of this section is to provide in-Initial calibration of each measuring system formation to applicants and licensees regarding should be performed using one or more of the the NRC staff's plans for using this regulatory reference standards that are certified by the guide.

National Bureau of Standards or standards that have been obtained from suppliers that partici. Except in those cases in which the applicant pate in measurement assurance activities with or licensee proposes an acceptable alternative NBS (see footnote 2). These radionuclide method, the staff will use the methods de-standards should permit calibrating the system scribed herein in evaluating an appheant's or, over its intended range of energy and rate licensee's capability for and performance in complying with specified portions of the Com-l .

. capabilities. For nuclear power plants. sources that have been related to this initial calibration mission's regulations after March 30. 1979.

should be used to check this initial calibration I' at least once per 18 months (normally during If an applicant or licensee wishes to use the:

l refueling outages), method described in this regulatory guide on i or before March 30,1979, the pertinent portions i Periodic correlations should be made during of the application or the licensee's performance -

operation to rehte monitor readings to the will be evaluated on the basis of this guide.

4.15-7

=. .

REFERENCES

1. Section 6.2, " Validation of Analyses ," ronmental Protection Agency Report, Office of Chapter 6, " Validity of Results," Methods of Research and Development, Environmental Mon-Radiochemical Analysis, World Health Organiaa- itoring Support Laboratory, EPA-600/4-76-049, tion, Geneva,1966. September 1976.

2. "Analytica1 Quality Contro1 Methods," Envi- 15. E.J. Maienthal and D. A. Becker, "A Sur-ronmental Radioactivity Surveillance Guide, vey on Current Literature on Sampling, Sample .

U.S. Environmental Protection Agency Report, Handling, and Long-Term Storage for Environ-ORP/SID 72-2, June 1972. mental Materials ," Interface ~5 (#4), 49-62 (1976). Also available from the jiuperintendent

3. Environmental Radiation Measurements, Re- of Documents , U.S. Government Printing port of NCRP SC-35, NCRP Report No. 50, Office, Washington, D.C. 20402, es NBS Tech-1976. nical Note #929 October 1976, C 13.46:929 S/N 003-003-01694-2.

4. L.G . Kanipe, Handbook for Analytical uallt Control & RadioanalyticafTaboratories, 16. Tritium Measurement Techniques, Report

. . nvironmental Protection Agency Report of NCRPTC~36,,NCRP Report No. 47, 1976.

EPA 600/7-77-088, August 1977.

17. " Guide to Sampling Airborne Radioactive

5. J.M. Mullins, C. Blincoe, J.C. Daly, et Materials in Nuclear Facilities," ANSI N13.1-al. , " Radiochemistry," Chapter 17, pp. 1007- 1969.

1031, Quality Assurance Practices for Health Laboratories , Stanlay. L. Inhorn "Teditor), 18. " Specification and Performance of On-Site American Public Health Association,1978. ' Instrumentation for Continuously Monitoring Radioactivity in Effluents," ANSI N13.10-1974.

6. Regulatory Guide 1.21, " Measuring. . Eval-unting, and Reporting Radioactivity in Solid 19. Air Sampling Instruments for Evaluation of A Wastes and Releases of Radioactive Materials in Atmos 3eric Contaminants , Eurth Edit $of Liquid and Gaseous Effluents from Light-Water- American Conference of Industrial Hygienists, ,

Cooled Nuclear Power Plants." 1972.

7. Regulatory Guide 4.5, " Measurements of 20. Users' Guide for Radioactivity Standards.

Radionuclides in the Environment-Sampling and Subcommittee on Radiochemistry and Subcom-Analyses of Plutonium in Soil." mittee on the Use of Radioactivity Standards, Committee on Nuclear Science , National

8. Regulatory Guide 4.6, " Measurements of Academy of Sciences-National Research Council Radionuclides in the Environment-Strontium-89 Report, NAS-NS-3115, February 1974, and Strontium-50 Analyses."

21. Environmental Impact Monitoring for Nu-

3. Regulatory Guide 4.13 "Perfonaance, clear Power Plants, Source Book of MoiiEorEr Testing, and Procedural Specifications for ReEods, Vol.1 Atomic IndustriH Forum Re-Thermoluminescence Dosimetry: Environmental Port. AIF/NESP-004, February 1975.

AppUcations."

22. Instrumentation for EnvironmentalMonitor-

10. HASL Procedures Manual, U.S. Energy .i,n_g: Radiation, Lawreiice Berkeley Laboratory Research and Development Administration Report, LBL-1, Vol. 3. First Ed. , May 1972; Report, HASL-300,1972 (updated annually). First update, February 1973; Second update, October 1973.

11. A Guide for Environmental Radiological Surveillance at ERDA Insta11auons, Energy Re. 23. C.W. Sill " Problems in Sample Treatment search and Development Administration Report, in Trace Analysis," National Bureau of Stand-ERDA 77-24, March 1977. ards Special Publication 4-22, Accuracy M Trace Analysis : Sampling, Sample Handling,

12. Handbook of Radiochemical Analytical Meth- an,,d Analysis, pp. 463-490 August 1976.

ods, U.S. EnT1ronmental Protection Agency Eport, EPA-680/4-75-001, February 1975. 24. " General Principles for Sampling Airborne Radioactive Materials," International Standard,

13. Standard Methods for the Examination of ISO-2889.1975.

O Water and Wastewater,"~ Thirteenth Edition.

American7ublic Health Association, IM5. 25. Manual of Methods M Chemical Analysis of Water and Wastes. EPA-625/6-74-003, U.S.

E"nvironmeriH1 Protection Agency. Office of

14. Handbook fo,e Sampling and Sam le Pres-

_ Technology Transfer, Washington, D.C. 20460, ervation of Water and Wastewater. . Env~l- gg74, 4.15-8

~ '

}j * .

P p

j 26. . " Standard Practices for Sampling Water," 34. L.H. Ziegler and H.M. Hunt, ualit 4 Method D 3370-76, Annual Book of ASTM Stand- Control for Environmental Measurements __s_uig  !

y ards (Part 31), Water, AmericanWety for Gamma-Ray Spectrometry, U.S. Envuonmental

, Tisting and Materials, Philadelphia, PA,1977. _ Protection Agency Report EPA-660/7-77-14, December 1977.

27. Biological Field and Laboratory Methods

~

for Measuring e uiH~t of Surfac.e waters 35. L.L. Thatcher, V.J. Janzer, and K.W.

i ind Effluents, ETA- - 3%T- 5fEce ofTe7 Edwards, " Methods for the Determination of search and Development, U.S. Environmental Radioactive Substances in Water and Fluvial 1

' Protection Agency, Cincinnati, Ohio, Sediments ," Chapter AS, Book 5 (Laboratory July 1973. Analysis) of Techniques of Water-Resources l Investigations of the United States. Geolorncal '

28. G.G. Eadie and D.E. Bernhardt, " Sampling Survey, 1977. Thipier A5 is available from the 1 i and Data Reporting Considerations for Air- Superintendent of Documents, U.S. Govern- l

} bortie Particulate Activity," U.S. Environ- ment Printing Office, Washington, D.C. 20402,  ;

mental Protection Agency Technical Note Stock Number 024-001-02928-6.

i ORP/LV-76-9, December,1976. I

?

36. A Handbook of Radioactivity Measurements l

29. A.J. Breslin, " Guidance for Air Sampling Proce3ures, Report of NCRP. SC-15A, NCRP

i at Nuclear Facilities," U.S. Energy Research Report No. 58, 1978. l and Development Administration Report HASL- I

. 312, November 1976. 37. P.E. Shelley, Sampling of. Water .and l Wastewater, U.S. Environmental Pi6tectliiis - l l

30. " Reference Method for the Determination Agency, Report EPA-600/4-77-039, August l

of Suspended Particulates in the Atmosphere 1977. .

l (High Volume Method)," gJualit Assurance Handbook for Air Pollution Weasurement 38. " Calibration and Usage of Germanium De-i tems, volume II Labient Air Specific Meth_5y_s-

~

ods, tectors for Measurement of Gamma-Ray Emission l 5ection 2 U.S. EnvirTo . mental Protection Rates of Radionuclides," ANSI N42.14-1978.

! Agency Report EPA-600/4-77-027a, May 1977.

39. D .C. Kocher (editor), Nuclear Decay i 31. E.D. Harward (editor), Program R rt: Data for Radionuclides Occurring in Routine Workshop on Methods for Measurint Ra taon i' -~ -

In and Around Uranium' Rills, report based on Feliisses from Nuclear Fuel LC cle 7aN Oak Ridge 'Nitional Laboratory Report onnL/

. presentations made at the workshop held in NUREG/TM-102, August 1977.

!- Albuquerque, New Mexico, ' May 23-26, 1977.

i Atomic Industrial Forum, Inc. , August 1977. 40. M.J. Martin (editor), Nuclear Decay Data for Selected Radionuclides, DaFl[idge NatEiiE i Eiboratory Report ORNL-5114, March 1976,

i. . 32. Measurement of Low-Level Radioactivity, i International Comadision on Radiation Units 41. International Directory of Certified Radio-

}, and Measurements (ICRU) Report 22, June 1,

~

active Materials, Internationil Atomic Energy j 1972. Agency Report, ST!/ PUB /398,1975.

42. Environmental Radioactivity Laboratory

33. American Public Health Association Inter- Intercomparison Studies Program, FY 1977, i l society Committee on Methods of Air Sampling EPA-600/4-77-001, January 1977.

i and Analysis, Morris Katz (editor), Methods of 4 Air Sampling -and Analysis, Second Edition-~ 43 " Guidelines for the Documentation of )

j Eerican PublicTealth Association,1977. Digital Computer Programs," ANSI N413-1974.

1 k k 4

1

)

b 4.15-9 j ..s i

GLOSSARY ,

1 Accuracy-a qualitative concept in the statis- Precision-a qualitative concept in the statis-l tical treatment of measurement data used to de- tical treannent of measurement data used to t scribe the agreement between the central describe the dispersion of a set of numbers I tendency of a set of numbers and their correct with respect to its central tendency.

j value (or the accepted reference value). It is

' also used to describe the agreement between an Quality Assurance {gA_}-the planned and l Individual value and the correct value (or the systematic actions that are necessary to pro- )

tecepted reference value). vide adequate confidence in the results of a ,

monitoring program. 1 Analytical Blank (Sample)-ideally, a sample having all of the constituents of the unknown Quality Control UC -those quality assurance sample except those to be determined. In re- actions that pmvide a means to control and dioanalytical practice, the term often refers to measure the characteristics of measurement the radiochemical processing of carrier (s) or equipment and processes to established re-tracers without the sample matrix material. quirements. Thus, quality assurance includes

" Blind" Replicate (Sample)-replicate samples that are not identified as replicates to ther per- Reference Test Material-e large batch of sons performing the analysis. homogeneous material from which aliquots may be taken for interlaboratory comparisons or for Calibration-the process of determining the internal use by the laboratory. The meterial numerical relationship between the observed must be uniform but need not be standardised.

output of a measurement system and the value, based on reference standards, of the charac-teristics being measured.

Spiked Sample-a sample to which a known Calibration Source-any radioactive source amount of radioactive material has been added.

that is used for calibration of a measurement Generally, spiked samples are submitted as system, unknowns to the analysts.

Check Source or instrument check source or_ ,

Split Sample-a homogeneous sample that is performance che source)-a radioactive source divided into parts, esca of which is analysed used to determine if the detector and all elec- independently by separate laboratory organi- '

tronic components of the system are operating zations.

correctly.

Standard (radioactive) Source-a radioactive Instrument Background-the response of the source having an accurately known radionu-instrument in the absence of a radioactive sam- clide content and radioactive decay rate or rate pie or other radioactive source, of particle or photon emission.

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,_ _ . _ . _ _ _.. _ . .. _ ..._.-. . . ~ .. _ _ _ . _ . _ _ _ _ . . - _ . _ . _ . _ . _ _ - . _ . .

l FNP-0-M-011 i

i TABLE OF CONTENTS E

i

{ DISTRIBUTION LIST . . . . . . ...................... 1 i

{ TABLE OF CONTENTS . .. . . . ...................... 11 j LIST OF TABLES- . . . . . . . ...................... v j LIST OF FIGURES . . . . . . . ...................... vil i

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii CHAPTER 1: INTRODUCTION . . . . .. . . . . . . . . . . . . . . . . . . 1-1 1

4 CHAPTER 2: LIQUID EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . 2-1 I

!, 2.1 LIMITS OF OPERATION 2-1 l 2.1.1 Liould Effluent Monitorina Instr"- ntation control 2-1 1 2.1.2 Lieuid Effluent concentration control 2-7 l 2.1.3 Liould Effluent Dose control 2-11 2.1.4 Liould Radwaste Treatment System Control 2-13

{ '2.1.5 Maior Channes to Liould Radioactive Waste Tre.gtment Systems 2-14 .

! 2.2 LIQUID RADWASTE TREATMENT SYSTEM 2-15 j_ 2.3' LIQUID EFFLUENT MONITOR SETPOINTS 2-19 p

2.3.1 General Provisions Renardina Setnoints 2-19 j 2.3.2 Setnoints for Radwasta System Discharas Monitors 2-21 l 2.3.3 Satooints for Monitors on Normally Low-Radioactivity St re== 2-29 1

l 2.4 LIQUID EFFLUENT DOSE CALCULATIONS 2-30 l, 2.4.1 calculation of Dose 2-30 l 2.4.2. Calculation of A.3 2-31 2.4.3 calculation of cF;y 2-32

2-42 2.5: . LIQUID EFFLUENT DOSE PROJECTIONS I 2.5.1 Thirty-One Day Dose Proiactions 2-42 2.5.2 Dose Proiections for Specific Releases 2 -4 2..

2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS 2-43 CHAPTER 3: GASEOUS EFFLUENTS . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 LIMITS OF OPERATION 3-1 3.1.1 Gaseous Effluent Monitoring Instrumentation Control 3-1 3.1.2 Gaseous Effluent Dose Rate Control 3-6 l r 3.1.3 Gaseous Effluent Air Dose Control 3-10 3.1.4 control on cameous Effluent Dose to a Member of the Public 3-12 11 Gen. Rev. 13

- _ _- _ _ . . _ _ . _ _ _ . _ . .. ._. . . _ m. _ . _ . . . _ _ _ _ _ _ _ _ _ _ . __-m.m.. _ _ _ _ . - . _ . _

t

!~

l FNP-0-M-011 TABLE OF CONTENTS'fcontinued)

V au 3.1.5 Gaseous Radwaste Treatment System control 3-14 3.1.6 MAJOR CHANGES to the GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM and the VENTILATION RYMAUST TREATMENT SYSTEM 3-16 3.2 GASEOUS RADWASTE TREATMENT SYSTEM 3 *.7 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3-19 3.3.1 General Provisions Recardina Noble Gas Monitor Setooints 3-19 3.3.2 Setooint for the Final Noble Gas Monitor on Each Release Pathway 3-21 1

3.3.3 Setooints for Noble Gas Monitors on Effluent Source Streams 3-25  !

3.3.4 Determination of Allocation Factors, AG 3-28 j 3.3.5 Setooints for Noble Gas Monitors with Soecial Raouirements 3-31

]

3.3.6 Setooints for Particulate and Iodine Monitors 3-31 l 1

3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3-32  ;

3.4.1 Dose Rates at and Beyond the Site Boundary 3-32 I i

3.4.2 Noble Gas Air Dose at or Bevond Site Boundarv 3-33 3.4.3 Dose to a Member of the Public at or Beyond Site Boundarv 3-37 3.4.4 Dome Calculations to Suonort Other Raouirements 3-40 l 3.5 GASEOUS ElFLUENT DOSE PROJECTIONS 3-46 3.5.1 Thirty-one Day Dose Proiections 3-46 3.5.2 Dose Proiections for Specific Releases 3-47 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS 3-48 CHAPTER 4: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . . . . . . . 4-1 4.1 LIMITS OF OPRRATION 4-1 4.1.1 Radioloaical Environmental Monitorina 4-1 4.1.2 Land Use Census 4-8 j 4.1 3 Interlaboratory Comoarison Proaram 4-10

4.2 RAD,tOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS 4-11 i

I CHAPTER 5: TOTAL DOSE DETERMINATIONS . . . . . . . . . . . . . . . . . 5-1 5.1 LIMIT OF OPERATION 5-1 l 5.1.1 Acclicability 5-1 5.1.2 Actions 5-1

$ 5.1.3 ' Surveillance Reauirements 5-2 j 5.1.4 RAA11 5-2 5.2 DEMONSTRATION OF COMPLIANCE 5-3 i

j CHAPTER 6: POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR

! ACTIVITIES INSIDE THE SITE BOUNDARY . . . . . .. . . . . . 6-1 1 6.1 REQUIREMENT FOR CALCULATION 6-1 L

' 111 Gen. Rev. 13

FNP-C-M-011 f TAsfE OF CGidEWiS (Continued) '

EAfsE 6.2 CALCULATIONAL METHOD 6-1  !

+

.. CHAPTER 7: REPORTS

.. . . . . . . . . . . . . . . . . . . . . . . . . 7-1  ;

7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 7-1 i

'7.1.1 Racuir-- r.t for Rer. ort 7-l' i 7.1.2 Report Contents 7-1

-7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 7-3 ,

7.2.1 Reauir-- iit for Reoort 7-3 7.2.2 Report Contents 7-3 7.3 MONTHLY OPERATING REPORT 7-7 7.4 SPECIAL REPORTS 7-7 t

CHAPTER 8 - METEOROLOGICAL MODELS . . . . . . . . . . . . . . . . . . . 8-1  ;

8.1 ATMOSPHERIC DISPERSION 8-1 8.1.1 Ground-Level Releases '

8-1 8.1.2 Elevated Releases 8-3 8.1.3 Mixed-Mode Releases 8-5 8.2 RELATIVE DEPOSITION 8-7

^

8.2'.1 Ground-Level Releases 8-7 8.2.2 Elevated Releases 8-7 .

8.2.3 Mixed-Mode Releases 8-8 8.3 ' ELEVATED PLUME-DOSE FACTORS 8-8 8.4 METEOROLOGICAL

SUMMARY

8-8 CHAPTER 9: METHODS AND PARAMETERS FOR CALCULATION OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, R,jj p . . . . . . . . . . . . . . . . . 9-1 9.1 INHALATION PATHWAY FACTOR 9-1 9.2 GROUND PLANE PATHWAY FACTOR 9-2 9.3 GARDEN VEGETATION PATHWAY FACTOR 9-3 9.4' GRASS-CON-MILK PATHWAY FACTOR 9-6 9.5' GRASS-GOAT-MILK PATHWAY FACTOR 9-10 9.6 GRASS-COW-MEAT PATHWAY FACTOR 9-14 CHAPTER 10: DEFINITIONS OF EFFLUENT CONTROL TERMS . . . . . . . ... . 10-1 10.1 TERMS SPECIFIC TO THE.ODCM 10-1  ;

10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS 10-5 1

iv Gen. Rev. 13

I FNP-0-M-011

LIST OF TABLES t

PAGE Table 2-1. Radioactive Liquid Effluent Monitoring Instrumentation 2-3 l Table 2-2. Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 2-5 Table 2-3. Radioactive Liquid Waste Sampling and Analysis Program 2-9 l Table 2-4. Applicability of Liquid Monitor Setpoint Methodologies 2-20 Table 2-5. Parameters for Calculation of Doses Due to Liquid Effluent Releases 2-35 Table 2-6. Element Transfer Factors 2-36 l Table 2-7. Adult Ingestion Dose Factors 2-37 Table 2-8. Site-Related Ingestion Dose Factors, Aj7 2-40 Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation 3-3 I Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3-5 Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program 3-8 Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies 3-20 l Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of )

Noble Gases 3-35 Table 3-6. Dose Factors for Exposure to Direct Radiation from Noble f Gases in an Elevated Finite Plume 3-36 i Table 3-7. Attributes of the Controlling Receptor 3-39 .

Table 3-8. Raipj for Ground Plane Pathway, All Age Groups 3-42 Table 3-9. Raipj for Inhalation Fathway, Child Age Group 3-43 Table 3-10. Rapj j for Cow Meat Pathway, Child Age Group 3-44 l Table 3-11. Raipj for Garden Vegetation Pathway, Child Age Group 3-45 Table 4-1. Radiological Environmental Monitoring Program 4-4 Table 4-2. Reporting Levels for Radioactivity Concentrations in Environmental Samples 4-6 i Table 4-3. Values for the Minimum Detectable Concentration 4-7 Table 4-4. Radiological Environmental Monitoring Locations 4-12 Table 6-1. Attributes of Member of the Public Receptor Locations Inside the SITE BOUNDARY 6-3 Table 8-1. Terrain Elevation Above Plant Site Grade 8-9 Table 8-2. Annual Average (17D) for Mixed Mode Releases 8-10 Table 8-3. Annual Average (i76) for Ground-Level Releases 8-11 Table 8-4. Annual Average (576) for Mixed Mode Releases 8-12 Table 8-5. Annual Average (676) for Ground-Level Releases 8-13 Table 9-1. Miscellaneous Parameters for the Garden Vegetation

/N Pathway 9-5

\ Table 9-2. Miscellaneous Parameters for the Grass-Cow-Milk Pathway 9-9 v Gen. Rev. 13

. . . . . . ~ -

4 FNP-O-M-011 LIST OF TABLES (Continued) 1 F.A91 Table 9-3. Miscellaneous Parameters for the Grass-Coat-Milk Pathway 9-13 Table 9-4. Miscellaneous Parameters for the Grass-Cow-Meat Pathway 9-17 Table 9-5. Individual Usage Factors 9-18 Table 9-6. . Stable Element Transfer Data 9-19 Table 9-7. Inhalation Dose Factors for the Infant Age Group 9-20 Table 9-8. Inhelation Dose Factors for the Child Age Group 9-23 Table 9-9. Inhalation Dose Factors for the Teenager Age Group 9-26 Table 9-10. Inhalation Dose Factors for the Adult Age Group 9-29 Table 9-11. Ingestion Dose Factors for the Infant Age Group 9-32 Table 9-12. Ingestion Dose Factors for the Child Age Group 9-35 Table 9-13. Ingestion Dose Factors for the Teenager Age Group 9-38 Table 9-14. Ingestion Dose Factors for the Adult Age Group 9-41 Table 9-15. External Dose Factors for Standing on Contaminated Ground 9-44 vi Gen. Rev. 13 i

_ . _ . ~ . __ __. . . . - . _ . _ . . _ _ _ _ . - . _ _ . _ _ _ . - . ~ . . . . . _ . _ . . _ . . _ . . _

.-__.._---..m I

] FNP-0-M-011 i t LIST OF FIGURES

\ .

P.h3

} Figure 2-1. Liquid Radwaste Treatment System (Typical of Both Units) 2-16  !

Figure 2-2. Steam Generator Blowdown System (Typical of Both Units) 2-17 l

} Figure 2-3.- Liquid Discharge Pathways 2-18 l Figure 3-1. Schematic Diagram of Routine Release Sources and Release r i Points (Typical of Both Units) 3-18 q Figure 4-1. Airborne Sampling Locations, 0-5000 feet 4-15 Figure 4-2. Indicator II (Community) Sampling Locations for Direct j- Radiation 4-16

{- Figure 4-3. Airborne Samp1 h; Locations, 0-20 miles 4-17 j Figure 4-4. Water Sampling Locations 4-18 l 3

4 Figure 8-1. Vertical Standard Deviation of Material in a Plume (og) 8-14 [6

! Figure 8-2. Terrain Recirculation Factor (K r) 8-15

{'

Figure 8-3. Plume Depletion Effect for Ground Level Releases 8-16 j Figure 8-4. Plume Depletion Effact for 30-Meter Releases 8-17

} Figure 8-5. Plume Depletion Effect for 60-Meter Releases 8-18  ;

i  !

l Figure 8-6. Plume Depletion Effect for 100-Meter Releases 8-19 1 Figure 8-7. Relative Deposition for Ground-Level Releases 8-20 '

k Figure 8-8. Relative Deposition for 30-Meter Releases 8-21 9 Figure 8-9. Relative Deposition for 60-Meter Releases 8-22 l' Figure 8-10. Relative Deposition for 100-Meter (or Greater) Releases 8-23 .

Figure 10-1. Site Map for Effluent Controls 10-8 i

l l

l vii Gen. Rev. 13 l l

.u. w

FNP-0-M-011 REFERENCES

/N

(') 1. J.S. Boegli, R.R. Bellamy, W.L. Britz, and R.L. Waterfield, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," NUREG-0133, October 1978.

l 2.

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. NRC Reculatorv Guide 1.10.2, March 1976.

3.

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. NRC Reculatory Guide 1.109. Revision 1, October 1977.

4. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Reculatory Guide 1.111, March 1976.

5. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Regulatory Guide 1.111. Revision 1, July 1977.

6. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implemer. ting Appendix I," U.S. NRC Reculatorv Guide 1.113, April 1977.

7. Joseoh M. Farlev Nuclear Plant Units 1 and 1_ Final Safety Analysis Report, Alabama Power Company.

8. Joseph M. Farlev Nuclear Plant Units 1 and 2 Environmental Report -

Operatina License Stace, Alabama Power Company.

A

/ 9. T.E. Young, T.S. Bohn, and W. Serrano, " Technical Evaluation Report for

\~ / ) the Evaluation of ODCM Revision 7 for Joseph M. Farley Nuclear Plant, 1

Units 1 and 2," EGG-PHY.8674, dated August 1989, transmitted by NRC .!l letter dated November 9, 1989. ,

I

10. W.M. Jackson, " Survey Report of Chattahoochee River Water Use Downstream l of Parley Nuclear Plant Liquid Effluent Discharge," dated July 19, 1990. I

11. J.E. Till and H.R. Meyer, eds., Radiolooical Assessment, U.S. NRC Report NUREG/CR-3332, 1983.

12. L.A. Currie, Lower Limit of Detection: Definition and Elaboration of a Proposed Position of Radiolooical Effluent and Environmental Measurements, U.S. NRC Report NUREG/CR-4007, 1984.

13. " Radiological Assessment Branch Technical Position", U.S. Nuclear Regulatory Commission, Revision 1, November 1979.

14. U.S. DOE Report PNL-5484.

15. D.C. Kocher, " Radioactive Decay Data Tables," U.S. DOE Report DOE / TIC-11025, 1981.

16. Internal Memorandum. J.E. Garlinoton to D.N. Morev, Alabama Power Company, June 4, 1990.

\m/'

viii Rev. 15

f' _ FNP-0-M-011

) CHAPTER 1 INTRODUCTION i

i The Offsite Dose Calculation Manual is a supporting document of the Technical

Specifications. As such, it describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous effluents, and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm setpoints. In addition, it contains the following

e The controls required by the Technical Specifications, governing the l

i radioactive affluent and radiological environmental monitoring programs.

j

, e schematics of liquid and gaseous radwaste effluent treatment systems,

which include designation of release points to UNRESTRICTED AREAS.

e A list and maps indicating the specific sample locations for the Radio-l j logical Environmental Monitoring Program.

• Specifications and descriptions of the information that must be included in the Annual Radiological Environmental Surveillance Report and the Annual Radioactive Effluent Release Report required by the Technical O specifications.

The ODCM will be maintained at the plant for use as a reference guide and training document of accepted methodologies and calculations. Changes in the calculational methods or parameters will be incorporated into the ODCM in order to ensure that it represents current methodology in all applicable areas. Any computer software used to perform the calculations described will be raintained current with the ODCM.

Equations and methods used in the ODCM are based on those presented in NUREG-0133 (Reference 1), in Regulatory Guide 1.109 (References 2 and 3), in Regulatory Guide 1.111 (References 4 and 5), and in Regulatory Guide 1.113 (Reference 6).

O 1-1 Gen. Rev. 13

4 i

i e

FNP-0-M-011

}

CHAPTER 2 LIOUID EFFLUENTS 2.1 LIMITS OF OPERATION The following Liquid Effluent controls implement requirements established by Technical specifications section 6.0. Terms printed in all capital letters are i defined in Chapter 10.

4 e

$ 2.1.1 Liould Effluent Monitorina Instrumentation control 1

In accordance with Technical specification 6.8.3.e(i), the radioactive-liquid j

cf fluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits specified in

} Section 2.1.2 cre not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with Section 2.3.

i-l 2.1.1.1 Applicability

! t This limit applies at all times.

i 2.1.1.2 Actions With a radioactive liquid effluent monitoring instrumentation channel alare/ trip cetpoint less conservative than required by the above control, immediately cuspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint to a conservative value.

With less than the mint .am number of radioactive liquid effluent monitoring i instrumentation channels OPERABLE, take the ACTION shown in Table 2-1.

This control does not affect shutdown requirements or MODE changes. 1 2.1.1.3 surveillance Requirements Ecch radioactive liquid effluent monitoring instrumentation channel shall be i demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Tchle 2-2. '

2-1 Gen. Rev. 13

. ~ . . . . . = . . . . .

. _ . . - . - . . . - . _ . _ . _ .. . -. . . . ~ . . _ . ~ . _ .. , . . - . _

h 0

FNP-0-M-011 2.1.1.4 Basis l

The radioactive liquid affluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid af fluents i

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 Section 2.3 to ensure that the alarm / trip will occur prior to exceeding the limits of Section 2.1.2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. '

I 4

) I 2-2 Gen. Rev. 13

i l

FNP-0-M-011 Table 2-1. Radioactive Liquid Effluent Monitoring Instrumentation e-ws ,

I

( )

'J I

OPERABILITY Requirementsa Instrument Minimum Channels l Operable ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RE-18) 1 28 I

b. Steam Generator Blowdown Effluent Line (RE-23B) 1 29

2. Flowrate Measurement Devices

a. Liquid Radwaste Effluent Line

1) Waste Monitor Tank No. 1 1 30

2) Waste Monitor Tank No. 2 1 30

b. Discharge Canal Dilution Line 4

(Service Water) 1 30

c. Steam Generator Blowdown Effluent Line 1 30 i (,,/

a. All requirements in this table apply to each unit.

1 1

i 1

I I l

1 rs 2-3 can. Rev. 13 1

4

- . - -_ . . - -=-_. - - _ .- - - _ - _ _ . - - . - .

l FNP-O-M-011 Table 2-1 (contd). Notation for Table 2 ACTION Statements I

. V ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 14 days provided that g ior to initiating a releases a.

At least two independent samples are analyzed in accordance with section 2.1.2.3, and

b. At least two technically qualified members of the Facility Staff

, independently verify the discharge line valving and (1) Verify the manual portion of the computer input for the 4

release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

t Otherwise, suspend release of radioactive effluents via this Pathway.

i 1

ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for grossradioactivity(begaorgamma)ataMINIMUMDETECTABLECONCENTRATION l no greater than 1 x 10- pCi/mL:

a. At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the specific activity of the i

O secondary coolant is greater than 0.01 pCi/ gram DOSE EQUIVALENT I-131.

0 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 /> when the specific activity of the secondary coolant is less than or equal to 0.01 C1/ gram DOSE EQUIVALENT I-131.

l 1

l ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue for up to 30 days provided the flowrate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be used to i estimate flow.

l l

l B

2-4 Gen. Rev. 13

l l

FNP-0-M-011 Table 2-2. Radioactive Liquid Ef fluent Monitoring Instrumentation Surveillance Requirements i

surveillance Requiremented a

Instrument CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK SOURCE CHECK CALIBRATION TEST

1. Cross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RE-18) D P Rb ga

b. Steam Generator Blowdown Effluent Line (RE-23B) D M Rb ga

2. Flowrate Measurement Devices l

a. Liquid Radwaste 4 Effluent Line l

1) Waste Monitor l Tank No. 1 DC NA R NA

2) Waste Monitor '

(

Tank No. 2 Dc NA R NA

b. Discharge Canal Dilution Line (Service Water) DC NA R Q

c. Steam Generator Blowdown Effluent Line DC NA R NA l

t i

p 2-5 Gen. Rev. 13

m.

FNP-O-M-011 Table 2-2 (contd). Notation for Table 2-2

(

(

a. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 10.2):

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

(a) Instrument indicates measured levels above the alarm / trip setpoint; (b) Loss of control power; or (c) Instrument controls loss of instrument power.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1 (a) Instrument indicates a downscale failure; or

(b) Instrument controls not set in operate mode.

b. The initial CHANNEL CALIBRATION shall be performed using on6 os more of

! the reference standards certified by the National Institute of Star.dards

, and Technology or using standards that have been obtained from suppliers

$ that participate in measurements assurance activities with NIST. For

' subsequent CHANNEL CALI5 RATION, scurces that have been related to the

!(\ initial calibration shall be used.

j c. CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on i days on which continuous, periodic, or batch releases are made.

} d. All requirements in this table apply to each unit.

i I

i t

i i

4 1

['\

2-6 Gen. Rev. 13 i

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l' I

)

FNP-O-M-011

'2.1.2 Liouid Effluent Concentration Control O In accordance with Technical Specifications 6.8.3.e(ii) and 6.8.3.e(iii), the ,

concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 10-1) shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 1 x 10-4 yCi/mL total activity.

! 2.1.2.1 Applicability This limit applies at all times.

2.1.2.2 Actions With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the limits stated in Section 2.1.2, immediately restore the concentration to within the stated limits.

This control does not affect shutdown requirements or MODE changes.

(

(. 2.1.2.3 Surveillance Requirements The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 2-3. The results of radioactive analyses shall be used with the calculational methods in Section 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2.

2.1.2.4 Basis This control is provided to ensure that the concentration of radioactive materials released in liquid waste ef fluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 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 I,10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 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 MPC

,(" in air (submersion) was converted to an equivalent concentration in watec using the methods described in International Commission on Radiological Protection 2-7 Gen. Rev. 13

1 r

k FNP-0-M-011 (ICRP) Publication 2 (1959). The resulting concentration of 2 x 10-4 wcs then multiplied by the ratio of the affluent concentration limit for Xe-135, stated in Appendix B, Table 2, Column 1 of 10 CFR 10 (paragraphs 20.1001 to 20.2401),

to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 (paragraphs 20.1 to 20.601), to obtain the limiting concentration of 1 x 10-4 j pci/mL.

1 l

1 1

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I l

O -

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1 I

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2-8 Gen. Rev. 13

l t

FNP-0-M-011 ,

Table 2-3. I Radioactive Liquid Waste Sampling and Analysis Program Q

1 I

Sampling and Analysis Requirements a,b l l

' MINIMUM Liquid DETECTABLE Minimum Release Sampling Analysis CONCENTRATION Type Type of Activity (MDC)

FREQUENCY FREQUENCY Analysis (uci/mL)

A. Waste Tanks Producing BATCH RELEASES i PRINCIPAL GAMMA 5 E-7 P P EMITTERS Each BATCH Each BATCH I-131 1 E-6 Dissolved and 1 E-5 One B TCH/M " * "* "'*"

All (Gamma Emitters)

P M H-3 1 E-5 Each BATCH COMPOSITE Gross Alpha 1 E-7 p g Sr-89, Sr-90 5 E-8

/~ Each BATCH COMPOSITE Fe-55 1 E-6 B. CONTINUOUS RELEASESC PRINCIPAL GAMMA 5 E-7 D W EMITTERS Grab Sample COMPOSITE I-131 1 E-6 Dissolved and 1 E-5 i Steam Grab ample M Entrained Gases Generator (Gamma Emitters) '

Blowdown D M H-3 1 E-5 Grab Sample COMPOSITE Gross Alpha 1 E-7 p D Q **~*5' O I E~O l Grab Sample COMPOSITE Fe-55 1 E-6

" "' PRINCIPAL GAMMA 5 E-7 P W EMITTERS Building Sump Grab Sample COMPOSITE H-3 1 E-5 O

2-9 Gen. Rev. 13

. ._ . ~ . _ _ . _ _ _ . . _ _ _ _ _ _ . . _ . _ . _ _ _ . _ . _ . . _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _- . _ . . . _ _

FNP-0-M-011 i

Table 2-3 (contd). Notation for Table 2-3

4. All requirements in this table apply to each unit. Deviation f rom the MDC j requirements of this table shall be reported in accordance with Section 1 7.2.

b. Terme printed in all capital letters are defined in Chapter 10.

c. Sampling will be performed only if the effluent will be discharged to the ,

environment.  !

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O 2-10 Gen. Rev. 13

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l FNP-0-M-011 2.1.3 Liould Effluent Dose Control I b

V In accordance with Technical Specifications 6.8.3.e(iv) and 6.8.3.e(v), the dose ,

or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid of fluents released, from each unit, to UNRESTRICTED AREAS (see Figure 10-1) shall J be limited: '

c. During any calendar quarter to less than or equal to 1.5 mrem to the total I body and to less than or equal to 5 mrem to any organ, and i b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

2.1.3.1 Applicability These limits apply at all times.

j 2.1.3.2 Actions With the calculated dose from the release of radioactive materials in liquid of fluents exceeding any of the limits of Section 2.1.3, preparc and submit to the

("~ s Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification

\

6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s); ,

l defines the corrective actions to be taken to reduce the releases; and defines I the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the limits of Section 2.1.3.

This control does not affect shutdown requirements or MODE changes.

1 2.1.3.3 Surveillance Requirements At least once per 31 days, cumulative dose contributions from liquid effluents

, for the current calendar quarter and the current calendar year shall be determined, for each unit, in accordance with section 2.4.

2.1.3.4 Basis 4

?

This control is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The limits stated in Section 2.1.3 implement the guides set forth in Section II. A of Appendix I. The ACTIONS stated in Section 2.1.3.2 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 will be kept "as low as 2-11 Gen. Rev. 13

M FNP-0-M-011 g is reasonably achievable." Also, for fresh water sites with drinking water

( supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the f acility will not result in radionuclide cencontrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculations in Section 2.4 implement the requirements in Section III.A of Appendix I, which state 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 KEMBER OF THE PUBLIC j through appropriate pathways is unlikely to be substantially underestimated. The cquations specified in Section 2.4 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 (Reference 3) and Regulatory Guide 1.113 (Reference 6).

This control applies to the release of liquid ef fluents from each unit at the cite. The liquid ef fluents f rom shared LIQUID RADWASTE TREATMENT SYSTEMS are to be proportioned between the units.

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2-12 Gen. Rev. 13

FNP-0-M-011 2.1.4 Liould Radwaste Treatment System Control

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s/

In accordance with Technical Specification 6.8.3.e(vi), the LIQUID RADWASTE TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the system shall be used to reduce radioactivity in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 10-1) would exceed 0.06 mrom to the total body or 0.2 mrem to cny organ of a MEMBER OF THE pUBLIC in 31 days.

2.1.4.1 Applicability This limit applies at all times.

2.1.4.2 Actions With radioactive liquid waste being discharged without treatment anu in excess of the above limits and appropriate portions of the LIQUID RALMA?tE TREATHENT SYSTEM not in operation, prepare and submit to the Nuclear Regulat:r'f Commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report which includes the following informations A. Explanation of why liquid radwaste was being discharged without treatment, (p) identification of any inoperable equipment or subsystems, and the reason

,l for the inoperability, I

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

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

This control does not affect shutdown requirements or MODE changes.

2.1.4.3 Surveillance Requirements Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with Section 2.5, during periods in which the LIQUID RADWASTE TREATMENT SYSTEMS are not being fully utilized.

The LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE:

a. by meeting the controls of Sections 2.1.2 and 2.1.3, or

?D V

2-13 Gen. Rev. 13

FNP-O-M-011

b. by operating the LIQUID RADWASTE TREATMENT SYSTEM equipment for at least

/~* 15 minutes at least once per 92 days unless the LIQUID RADWASTE TREATMENT SYSTEM equipment has been utilized to process radioactive liquid effluents during the previous 92 days.

2.1.4.4 Basis l

The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the UNRESTRICTED AREAS. 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 L achievable." -This control 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.

2.1.5 Maior chances to Liauid Radioactive Waste Treatment Systems a Licensee initiated MAJOR CHANGES TO LIQUID RADIOACTIVE WASTE TREATMENT SYSTEMS:

i

! a. Shall be reported to the Nuclear Regulatory Commission in the ' Annual Radioactive Effluents Release Report for the period in which the change l-l' was implemented, in accordance with Section 7.2.2.7.

b. Shall become effective upon review and approval in accords.nce with l- Technical Specification 6.5.3.1.

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! 2-14 Gen. Rev. 13 J -

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FNP-0-M-011 2.2 LIQUID RADWASTE TREATMENT SYSTEM L The Farley Nuclear Plant is located on the west bank of the Chattahoochee River cpproximately 35 river miles above the point where it empties into Lake Seminole.

There are two pressurized water reactors on the site. Each unit is served by a completely separate LIQUID RADWASTE TREATMENT SYSTEM that is illustrated cchematically in Figure 2-1. However, both units share a common demineralizer bed system for processing liquida prior to release from the site. As shown in Figure 2-2, the Steam Generator Blowdown System is a separate entity. Liquid discharge pathways are shown in Figure 2-3.

All liquid radwastes treated by the LIQUID RADWASTE TREATMENT SYSTEM are collected in 5,000-gallon Waste Monitor Tanks for sampling and analysis prior to release. Prior to sampling, each waste monitor tank is recirculated for a minimum of two tank content volumes, to ensure that a representative sample can be taken from the tank. Releases from the waste monitor tanks are routed to the Service Water discharge line (which provides dilution prior to release to the l UNRESTRICTED AREA), and thence to the Chattahoochee River. The Service Water discharge line also receives input from the Cooling Tower Blowdown and the Turbine Building Sump.

I Although no significant quantities of radioactivity are expected in the steam V(S generator blowdown processing system, this effluent pathway is monitored as a precautionary measure. The monitors serving this pathway provide for automatic j termination of release in the event that radioactivity is detected above predetermined levels. Like the LIQUID RADWASTE TREATMENT SYSTEMS, the Steam Generator Blowdown Systems discharge to the Service Water discharge line.

One potential release pathway, the Turbine Building Sump discharge, is not monitored during release, but is sampled regularly during discharges. Sampling l cnd analysis of releases via this pathway must be sufficient to ensure that the liquid effluent dose limits specified in Section 2.1.3 are not exceeded.

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a 2-15 Gen. Rev. 13 l

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'i rAunomr WASTE FLOOR CHEMICAL Ek HOLDUP DRAIN DRAIN y ramm e TANK TANK TANK genrr 1 osn.fi i

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g h MONITOR T=x 2 AMK 1 SERVICE WATER 3

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EVAPORATOR t 7 16 P COIIDEllSATE c TANK <-I D d f 1/2 018 n h 4s3 a

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l O Figure 2-2. Steam Generator Slowdown System (Typical of Both Units) 2-17 Gen. Rev. 13 i

. . . - . . - - - - . - _ - - . - . . - . = - . = - . . . . . . . ~ -

i FNP-0-M-011 O

Unit 1 Unit 2 Service Water Service Water Train A Train B Train A Train B I

1 4 Cooling Tower Blowdown Radwaste Discharge 1RE018 2RE018 Steam Cenarator Blowdown IRE 23B 2RE235 l i

l I Turbine Building Sump 7 l

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To River l

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O Figure 2-3. Liquid Discharge Pathways 2-18 Gen. Rev. 13 e

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1 FNP-O-M-011 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2.3.1 General Provisions Recardino Setooints j I

Liquid monitor setpoints calculated in accordance with the methodology presented in this section will be regarded as upper bounds for the actual high alarm )

setpoints. That is, a lower value for the high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be i

established at an appropriate level to give suf ficient warning prior to teaching l the high alarm setpoint. If no release is planned for a particular pathway, or '

[ if there is no detectable activity in the planned release, the monitor setpoint i should be established as close to background as practical to prevent spurious r

i alarms, and yet alarm should an inadvertent release occur.

Two basic setpoint methodologies are presented below. For radweste system +

discharge monitors, setpoints are determined to assure that the limits of Section 2.1.2 are not exceeded. For monitors on streams that are not expected to contain significant radioactivity, the purpose of the monitor setpoints is to cause an alarm on low levels of radioactivity, and to terminate the release where this is possible, section 2.1.1 establishes the requirements for liquid effluent monitoring instrumentation. Table 2-4 lists the monitors for which each of the O setpoint raethodologies is applicable.

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2-19 '

Gen. Rev. 13

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2 FNP-0-M-011

Table 2-4. Applicability of Liquid Monitor Setpoint Methodologies

. O Liquid Radwaste Discharge Monitors Setpoint Methods Section 2.3.2 Unit 1 or Unit 2 Waste Monitor Tanks Effluent

, Release Type: BATCM Monitor: 1RE-018 / 2RE-018 Unit 1 or Unit 2 Steam Generator Blowdown Effluent

Release Type: CONTINUOUS i Monitor: 1RE-023 8 / 2RE-023 B i

I I

Normally Low-Radioactivity Streams with Termination or Diversion upon Alarm 1

Farley Nuclear Plant has no liquid effluent streams in this category. l i

Normally Low-Radioactivity Streams with Alarm Only 4

3 Farley Nuclear Plant has no liquid effluent streams in this category.

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FNP-0-M-011 2.3.2 Setooints for Radwaite System Discharoe Monitors O 2.3.2.1 overview of Method LIQUID RADWASTE TREATMENT SYSTEM effluent line radioactivity monitors are intended to provide alarm and automatic termination of release prior to exceeding the limits specified in Section 2.1.2 at the point of release of the diluted ef fluent into the UNRESTRICTED AREA. Therefore, their alarm / trip setpoints are established to ensure compliance with the following equation (equation adapted l

from Addendum to Reference 1):

F+f

$ TF

• CECL (2.1) where CECL = the Ef fluent concentration Limit corresponding to the mix of radionuclides in the effluent being considered for discharge, in pCi/mL.

e= the setpoint, in pCi/mL, of the radioactivity monitor measuring the concentration of radioactivity in the affluent line prior to dilution and subsequent release. The setpoint represants a concen- '

tration which, if exceeded, could result in concentrations exceeding the limits of Section 2.1.2 in the UNRESTRICTED AREA.

f= the af fluent flowrate at the location of the radioactivity monitor, in gpm.

F= the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA, in gpm. A predetermined

) dilution flowrate must be assured for use in the calculation of the i radioactivity monitor setpoint.

s TF = the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could j

] accommodate ef fluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2; the

) tolerance factor must not exceed a value of 10. )

i

While equation (2.1) shows the relationships of the critical parameters that j O determine the setpoint, it cannot be applied practically to a mixture of radio--

l 1

2-21 Gen. Rev. 13 l 4

FNP-0-H-011 nuclides with dif ferent Ef fluent concentration Limits (E'JLs) . For a mixture of

/O radionuclides, equation (2.1) is satisfied in a practicaele manner based on the calculated ECL fraction of the radionuclide mixture and the dilution stream i

flowrate that can be assured for the duration of the release (Fd ), by calculating 1 the maximum permissible offluent flowrate (f ,) and the radioactivity monitor setpoint (c). l The setpoint method presented below is applicable to the release of only one tank  !

of liquid radwaste per reactor unit at a given time. Liquid releases must be

f controlled administratively to ensure that this condition is met; otherwise, the setpoint method may not ensure that the limits of Section 2.1.2 are not exceeded. '

0~

2.3.2.2 Setpoint Calculation Steps Steo 1: Determine the radionuclide concentrations in the liquid waste being i considered for release in accordance with the sampling and analysis l requirements of Section 2.1.2.

To ensure that sample analyses are based on samples that are representative of the waste being sampled, the liquid volume must be mixed thoroughly prior to campling. Mixing may be accomplished by any method that has been demonstrated O to achieve sufficient mixing for representative sampling. The Waste Monitor Tanks are recirculated for a minimum of two tank c.ntent volumes prior to .

campling. The Service Water discharge line is assumed to be well mixed, so that no additional mixing is required prior to sampling.

The total concentration of the liquid waste is determined by the results of all required analyses on the collected sample, as follows:

{Cf =

C,+[C,+Cf + C, + [ Cg (2.2) i s g where C, = the gross concentration of alpha emitters in the liquid waste, not l less than that measured in the most recent applicable composite sample.

C, = the concentration of strontium radioisotope s (Sr-89 or Sr-90) in tha.14a"'d wam W not less than that measured in the most recent applicable composite sample, d

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2-22 Gen. Rev. 13 l

w. ..

FNP-0-M-011 Cf= the concentration of Fe-55 in the liquid waste, not less than that

) measured in the most recent applicable composite sample.

Cg= the concentration of H-3 in the liquid waste, not less than that measured in the most recent applicable composite sample.

Cg= the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the sample for the release under consideration.

The C g term will be included in the analysis of each waste sample; terms for h gross concentrations of alpha emitters, Sr-89, Sr-90, Fe-55, and tritium will be included in accordance with the sampling and analysis program required for the waste stream (see Section 2.1.2). For each analysis, only radionuclides identified and detected above background for the given measurement should be included in the calculation. When using the alternate setpoint methodology of ctop 5.b, the historical maximum values of C., C,, Cg, and Cg shall be used.

I Stoo 2: Determine the required dilution factor for the mix of radionuclides  !

detected in the waste.  ;

1 b,O I Measured radionuclide concentrations ate used to calculate ECL fractions. The ,

ECL fractions are used along with a safety factor to calculate the required dilution factor; this is the minimum ratio of dilution flowrate to waste flowrate that must be maintained throughout the release to ensure that the Effluent Con-centration Limits of Section 2.1.2 are not exceeded at the point of discharge into the UNRESTRICTED AREA. The required dilution factor, GDF, is calculated as the sum of the dilution factors required for gamma emittets (RDFy ) and for non-l gamma-emitters (RDFny)8 C

RDF =

{i ggg;I +\(H) (W) )

(2.3)

= RDTy + RDTny C

E

{ ECL # (2.4)

RDr = #

T (SF) (TP)

O 2-23 Gen. Rev. 13

' l 0- l l

I FNP-0-M-011 Ca C, C Cg

+

. . f.

ECL a s ECL, ECL f ECL g, g2,5y RDi

• n7 (SF) (TF)

I

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where:  !

I l

C; = the measured concentration of radionuclide i as defined in step 1, l

1 in pCi/mt. The C., C,, cf, and Cg terms will be included in the calculation as appropriate. j h

I t ECLj = the Effluent Concentration Limit for radionuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 (except for noble gases as discussed Lelow). In the absence of information regarding the solubility classification of a given radionuclide in the waste

stream, the solubility class with the lowest ECL shall be assumed.

For dissolved or entrained noble gases, the concentration shall be limited to lx10-4 Ci/mL. For gross alpha, the ECL shall be 2x10-9 pCi/mL; if specific alpha-emitting radionuclides are measured, the i ECL for the specific radionuclide(s) should be used.

j SF = the safety factor selected to compensate for statistical l fluctuations and errors of measurement. The value for the safety s

4 factor must be between 0 and 1. A value of 0.5 is reasonable for

j. liquid _ releases; a. nore precise value may be developed if desired.

j

TF = the tolerance factor (as defined in Section 2.3.2.1).

I j sten 3: Determine the release-specific assured dilution stream flowrate. l l I i

l Determine the dilution stream flowrate that can be assured du sg the release period, designated T d i

If; simultaneous redioactive releases are planned from the same reactor unit, the  ;

unit's dilution stream must be allocated among all the simultaneous releases, l whether or not they are monitored during release. Normally, only the Waste Monitor Tank and steam Generator Blowdown effluents need be considered, unless I there is detectable radioactivity in one of the normally low-radioactivity l streams (see Table 2-4), or in the Turbine Building Sump. Allocation of the i dilution stream to multiple release paths is accomplished as follows:

iQ U

where

i

, 2-24 Gen. Rev. 13 l

rm FNP-0-M-011 Edp "

la i ^fp) (2.6)

Fdp = the dilution flowrate allocated to releaoe pathway p, in gpm.

AFp = the dilution allocation factor for releasa pathway p. Arp may be 4

assigned any value between 0 and 1 for each active release pathway, j

'l under the condition that the sum of the AFp for all active release j pathways for each unit does not exceed 1. (Notes Because the two l units have separate dilution streams, the two units do not affect

each other with respect to dilution allocation.]

l Fd= the assured minimum dilution flowrate for the unit, in gpm.

4 i

l If more precise allocation factor value., are desired, they may be determined

] based on the relative radiological impact of each active release pathway; this j may be approximated by multiplying the RDF of each effluent stream by its .

respective planned release flowrate, and comparing these values. If only one j release pathway for a given reactor unit contains detectable radioactivity, its j 3

AFp may be assigned the value of 1, making Fdp equal to Fd '

O I

For the case where RDF i 1, the planned release meets the limits of Section 2.1.2 i without dilution, and may be released with any desired effluent flowrate and dilution flowrate.

l I

sten 4

Determine the maximum allowable waste discharge flowrate.

t For the case where RDF > 1, the maximum permissible effluent disch'arge flowrate j k for this release pathway, fg (in gpm), is calculated as follows:

y mp . 's? a.,,

( ,,7 _ 3)

For the case RDF s 3, equation (2.7) is not valid. However, as discussed above, when RDF 51, the release may be made at full discharge pump capacity; the radio-activity monitor setpoint must still be calculated in accordance with step 5 below.

NOTE la Discharge flowrates are actually limited by the discharge pump capacity. When the calculated maximum permissible release flowrate exceeds the pump capacity, tne release may be made at full capacity. Discharge flowrates less than the pump capacity must be 2-25 Gen. Rev. 13

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l 4 FNP-O-M-011

! achieved by throttling if this is available; if throttling is not

-available, the release may not be made as planned.

NOTE 2: If, at the time of the planned release, there is detectable radic~

activity due to plant operations in the dilution stream, the diluting capacity of the dilution stream is diminished. (In addition, sampling and analysis of the other radioactive af fluents af fecting the dilution stream must be sufficient to ensure that the l liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.) Uncier these conditions, equation (2.7) {

must be modified to account for the radioactivity present in the l dilution stream prior to the introduction of the planned release:

F f C-& \

f mp . P (RDF - 1) 1.{ r J{

Tg 4 , ECLg, ,

(2.8) l where: i C;, = the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the

'% dilution stream.

fr = the affluent discharge flowrate of release pathway r.

l l

If the entire dilution stream contains detectable activity due to plant operations, whether or not its source is identified, fr=F' d and Cir is the concentration in the total dilution system. This j note does not a.pply: a) if the RDF of the planned release is s 1; or b) if the release contributing radioactivity to the dilution stream has been accounted for by the assignment of an allocation factor.

stes 5: Determine the mar (mum adia=ativity mani ene ==Pi at concentration. i Based on the values determined in previous steps, the radioactivity monitor cetpoint for the planned release is calculated to ensure that the limits of Section 2.1.2 will not be exceeded. Because the radioactivity monitor responds primarily to gamma radiation, the monitor setpoint cp for release pathway p (in pci/mI4.~ia hased on the concentration of gamma emitters in the waste stream, as follower

(

2-2C Gen. Rev. 13

_ _ . _ _ _ . ~ . . . _ . _ _ . . .. . _ _ . _ . . . . _ . . _ . . _ . . . _ . _ - _ . . _

i i

4

. FNP-O-M-Oli

$O #

P

" A Pb#8 (2.9)

Y E y

1

where

A p= an adjustment factor which will allow the setpoint to be i established in a practical manner to prevent spurious alarms while allowing .a margin between measured concentrations and the limits j of section 2.1.2.

i sten 5.a. If the concentration of gamma emitters in the effluent to l

be released is sufficient that the high alarm setpoint can be established at a level that will prevent spurious alarms, Ap should be calculated as follows:

1 A = x Mr P yy

, 1 , (Edp + fap) (2.10)

MF f ap O: where:

ADF = the assured dilution factor.

f,p = the anticipated actual discharge flowrate for the planned release (in gym), a value less than f 9

The release must then be controlled so that the actual effluent discharge flowrate does not exceed f,p at any time.

sten 5.b. Alternatively, Ap may be calculated as follows:

A = ADF - Mr" (2.11)

P Mr y i

sten 5.c. Evaluate the computed value of Ap as follows:

If Ap t 1, calculate the monitor setpoint, c.- p However, if cp is

'O V within about 10 percent of C g, it may be impractical to 2-27 Gen. Rev. 13

FNP-O-M-011 4 use this value of e. This situation indicates that measured concentrations are approaching values which would cause the limits of Section 2.1.2 tp be exceeded. ,

Therefore, steps snould be taken to reduce potential con-centrations at the point of discharge; these steps may include decreasing the planned effluent discharge flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the l ef fluent concentrations by further processing the liquid 3 planned for release. Alternatively, allocation factors )

, for the active liquid release pathways may be reassigned.

l When one or more of these actions has been taken, repeat

. Steps 1-5 to calculate a new radioactivity monitor i t setpoint.

l

, 1 If Ap < 1, the release may not be made as planned. Consider the

) alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the l actions taken.

- 2.3.2.3 Use of the Calculated setpoint I

The monitor actually *

The setpoint calculated above is in the units yC1/mL.

measures a count rate that includes background, so that the calculated setpoint must be converted accordingly i

e cp = c pEp+Sp (2.Sa)

> where c = the monitor setpoint as a count rate.

E e the monitor calibration factor, in count rate /(pci/mL). Monitor p

calibration data for conversion between count rate and concentration may include operational data obtained from

determining the monitor response to stream. cone =n*n* inns measured by liquid sample analysis.

B = the monitor background count rate. In all cases, monitor p

background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value.

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2-28 Gen. Rev. 13 4

__ _ _ . _ . _ _ _ . . . -_._..___._.__.______.__mm_._..

j i

FNP-0-M-011 The count rate units of c*, Ep , and Bpin equation (2.8a) must be the same (cpm f

or cys).

l i

2.3.3 setooints for Monitors on Normally Low-Radioactivity Streams l

i j Radioactivity in these streams (listed in Table 2-4 above) is expected to be at I

very low levels, generally below detection limits. Accordingly, the purpose of ,

these monitors is to alarm upon the occurrence of significant radioactivity in l

! these streams, and to terminate or divert the release where this is possible.

i l 2.3.3.1 Normal Conditions 1

When radioactivity in one of these streams is at its normal low level, its radio-l

{ cetivity monitor setpoint should be established as close to background as 4

l practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.

1*

2.3.3.2 conditions Requiring an Elevated Setpoint j Under the following conditions, radionuclide concentrations must be determined

' I end an elevated radioactivity monitor setpoint determined for these pathways: ,

.t e For streams that can be diverted or isolated, a new monitor setpoint must be established when it is desired to discharge the stream directly to the i dilution water even though the radioactivity in the stream exceeds the level which would normally be diverted or isolated.  ;

• For streams that cannot be diverted or isolated, a new monitor set; u.nt must be established whenever: the radioactivity in the stream becomes detectable above the background levels of the applicable laboratory analyses; or the associated radioactivity monitor detects activity in the stream at levels above the established alarm setpoint.

When an elevated monitor setpoint is required for any of these effluent streams, it should be determined in the same manner as described in section 2. 3.2.

However, special consideration must be given to Step 3. An allocation factor must be assigned to the normally low-radioactivity release pathway rander consideration, and allocation factors for other release pathways discht.rging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the normally low-radioactivity streams must be sufficient to ensura.that-the. liquid .

effluent dose limits specified in the controls of section 2.1.3 are not exceeded.

2-29 Gen. Rev. 13

- . . . _ . _ _ _ _, _ . . . _ . _ . _ . _ . - _ _ . _ _._.-._,m .. _ _ - _ __. _____.,

i i

FNP-0-M-011 4

2.4 LIQUID EFFLUENT DOSE CALCULATIONS t

4 The following sub-sections present the methods required for liquid of fluent dose l

j calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D , 7Air, and CF;y are summarized in

! Table 2-5.

5 2.4.1 calculation of Dose i

j The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a

]

l per-unit basis. Therefore, the doses calculated in recordance with this section j must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units.

For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid ef fluents released from each l l

unit to UNRESTRICTED AREAS will be calculated as follows (equation from Ref- I l

arence 1, page 15):

m <

07 = { Ajf { {Atf Cff E)f (2.12)  ;

i . l=1 i

=

4 where i

Dr = the cumulative dose commitment to the total body or to any organ 7, j in mram, due to radioactivity in liquid effluents released during )

l the total of the m time periods Atg. )

a l

Air = the site-related adult ingeution dose commitment factor, for the

. total body or for any organ 7, due to identified radionuclide i,

in (mromemL)/(h pCli. Methods for the calculation of Aj7 are i.

presented below in doction 2.4.2. The values of Ajy to bw used je

! dose calculations for releases from the plant site are listed in

! Table 2-?.

t

!l Atg = the length of time period 1, over which C;g and Fg are averaged for liquid releases, in h.

C;g = the average concentration of radionuclide i in undiluted liquid 4 effluent during time period 1, in pC1/mL. Only radionuclides i

2-30 Gen. Rev. 3 i

. - .-. - .. . -. -.- ..~ -. . - . . - . _ - - _ _ .

i l

)' FNP-O-M-011 identified and detected above background in their respective l samples should be included in the calculation.

j Fj = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:

a fg

=

Tf (2.13)  !

, rxZ g

j where: l ft= the average undil ded liquid waste flowrate actually

] observed during the period of radioactivity release, in 4

1 l gPm.

j j i l

Fg= the average dilution stream flowrate actually observed l during the period of radioactivity release, in gym.

! l Z= the applicable dilution factor for the receiving water

! body, in the near field of the discharge structure, i

during tha ponied o f. radioactivity release, from Table 2-5.

l  %

NOTE

In equation (2.13), the product (Fg x Z) is limited to 1000 cfs (= 448,000 gpm) or less. (Reference 1, Section 4 4.3.)

e

~

2.4.2 calculation of A37 l

{ The site-related adult ingestion dose commitment factor, A;7, is calculated as  ;

l follows (equation adapted from Reference 1, page 16, by addition of the irrigated garden vegetation pathway):

= 1.14 x 10 5 , -Al 'w + Uf Arg e I #/ + u, crg, Drf, (2.14)

Aj7 4

) where:

i 1.14 x 105 = a units conversion factor, determined by: I 100 pci/pci x 10 3 mL/L + 8760 h/y..

i

\(

• \

I i 2-31 Gen. Rev. 13 1

l

. . .~ . _ . -_

~ _ . ~ . - , . . . - . - . - . . . ~ - ~ - .. . . - . _ . -

FNP-O-M-011 U, = the adult drinking water consumption rate applicable to the plant site (L/y).

D, = the dilution f actor from the near _ field of the discharge structure lt for the plant site to the potable water intake location.

1; = the decay constant for radionuclide 1 (h-I) . Values of lj used in effluent calculations should be based on decay data from a  ;

recognized and current source, such as Reference 15.

', tw = the transit time from release to receptor for potable water

! consumption (h).

I

{ Ur = the adult rate of fish consumption applicable to the plant site I (kg/y).

} BF = the bioaccumulation factor for radionuclide i applicable to freshwater fish in the receiving water body for the plant site, in l (pci/kg)/(pci/L) = (L/kg). For specific values applicable to the l plant site,' see Table 2-6.

b10 l

tg = the transit time from release to receptor for fish consumption (h). ,

l

] Uy = the adult consumption rate for irrigated garden vegetation l applicable to the plant site (kg/y).

t l CFjy = the concentration factor for radionuclide i in irrigated garden 1

vegetation, as applicable to the vicinity of the plant site, in

]

! (pci/kg)/(pci/L). Methode for calculation of CFjy are presented l below in Section 2.4.3.

i l

l DFg7 = the dose conversion factor for radionuclide i for adults, in organ t (mrom/pci). For specific values, see Table 2-7.

1 2.4.3 calculation of CFy

-l

$- The concentration f actor for radionuclide i in irrigated garden vegetation, CFjy i

in (L/kg), is calculated as follows:

)

iO i

l 2-32 Gen. Rev. 13

FNP-0-M-011 e o For radionuclides other than tritium (equation adapted from *rference 3, equations A-8 and A-9):

%)

~

crfy = M*I # Il ~

• I + e~ l 'h (2.15)

Yv AEl P kg l

o For tritium (equation adapted from Reference 3, equations A-9 and A-10):

CF;y = M*L y {2,1s}

where:

M= the additional river dilution f actor from the near field of the ]

discharge structure for the plant site to the point of irrigation water usage.

I= the average irrigation rate during the growing season (L)/(m *h) 2 .

r= the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.

Yy = the areal density (agricultural productivity) of leafy garden vegetation (kg/m 2) fg = the fraction of the year that garden vegetation is irrigated.

Bjy = the crop to soil concentration factor applicable to radionuclide i, from Tabla 2-6 (pct /kg garden vegetation)/(pci/kg soil).

P= the effective surface density of soil (kg/m 23, 1; =- the-desey- constant for radionuclide 1 (h'I) . Values of 1 used in afflumat. ent~1*+4ans should be based on decay data from a recognized and current source, such as Reference 15.

1, = the rate constant for removal of activity from plant leaves by weatharing (h'L) .

O V

2-33 Gen. Rev. 13

FNP-0-M-011 AEi = the effective removal rate for activity deposited on crop leaves (h-I) calculated as: 1Ei " Ai+A w-i

! t, = the period of leafy garden vegetation exposure during the growing season (h).

tb= the period of long-term buildup of activity in soil-(h).

th= the time between harvest of garden vegetation and human consumption (h).

1

Ly = the water content of leafy garden vegetation edible parts (L/kg).

I i  !

J l

4 l

1

,' \

i 4

O

. U 3-i e

i i

i f )

0 k

t

! 2-34 Gen. Rev. 13 l

1

l

, l

, FNP-0-M-011 l Table 2-5. Parameters for Calculation of Doses Due to Liquid Ef fluent Releases l

\

, (V Dose calculation Receptor Locations

{

Tight Vicinity of plant discharge Drinkino Water: None (Ref. 10) l Irricated Garden Vecetation: Farms at River Mile 26 (Ref. 10)  !

Numerical Parameters ,

i l P.pyameter Value Reference j

Z 5 Ref. 2, Table A-1 Uw 0 L/y

• Ref. 10

. Dw 1.0

• Based on Ref. 1, Section 4.3.1 t, 12 h

• Ref. 3, Sec. A.2 j

Uf 21 kg/y Ref. 3, Table E-5  ;

1 tg 24 h Ref. 3, Sec. A.2 l

Uy 64 kg/y Ref. 3, Table E-5

! O M 0.04 Ref. 16

} 9 I 0.126 L/(m* h) Ref. 10, using pump capacity, ,

garden size, and irrigation 10% of the time during growing season.

r O.25 Ref. 3, Table E-15 Yy 2.0 kg/m2 Ref. 3, Table E-15 fg 0.1 Ref. 10 P 240 kg/m2 Ref. 3, Table E-15 1, 0.0021 h'I (i.e., half- Ref. 3, Table E-15 life = 14 d) t, 1440 h (= 60 d) Ref. 3, Table E-15 tb 1.31 x 105 h (= 15 y) Ref. 3, Table E-15 th 24 h Ref. 3, Table E-15 Ly 0.92 L/kg Based on Ref. 11, Table 5.16 (for lettuce, cabbage, etc.)

• Because there is no drinking water pathway downstream of the plant site, the consumption of drinking water is set to zero, and the default values of tw and D, are used.

s . .

2-35 Gen. Rev. 13 i

J

FNP-0-M-011 Table 2-6. Element Transfer Factors (O' ') Freshwater Fish Leafy Garden

, Vegetation Element BF i

+

Biv l H 9.0 E-01 i

4.8 E+00 C 4.6 E+03 5.5 E+00

! Na 1.0 E+02 5.2 E-02 1

P 3.0 E+03 1.1 E+00 Cr 2.0 E+02 2.5 E-04 Mn 2.0 E+01 2.9 E-02

, Fe 1.0 E+03 6.6 E-04 Co 1.0 E+02 9.4 E-03 j Ni 1.0 E+02 1.9 E-02 Cu 1.5 E+02 1.2 E-01 1 Zn 1.0 E+02 4.0 E-01 Br 4.2 E+02 7.6 E-01 l Rb 2.0 E+03 1.3 E-01 Sr 3.0 E+01 1.7 E-02 Y 2.5 E+01 2.6 E-03 a

Zr 2.0 E+02 1.7 E-04 1 Nb 1.0 E+02 9.4 E-03 Mo ,1.0 E+02 1.2 E-01 1 Tc 1.5 E+01' 2.5 E-01 l

~ IT Ru 1.0 E+01  !'.0 E-02 l (s,/ Rh 1.0 E+01 1.3 E+01 4

Ag 2.3 E+00 1.5 E-01 Sb 3.0 E+02 1.1 E-02 Te 2.0 E+03 1.3 E+00 I 2.0 E+01 2.0 E-02 Cs 2.0 E+02 1.C E-02 Ba 4.0 E+01 5.0 E-03 4 La 2.5 E+01 2.5 E-03

< Ca 2.0 E+02 2.5 E-03 Pr 2.5 E+01 2.5 E-03 Nd 2.5 E+01 2.4 E-03 W 1.2 E+03 1.8 E-02 Np 1.0 E+01 2.5 E-03

• Bioaccumulation Factors for freshwater fish, in (pCi/kg)/(pCi/L).

They are obtained from Reference 3 (Table A-1), except as follows:

Reference 9 for P; Reference 2 (Table A-8) for Ag; Reference 8 for Mn, Fe, Co, Cu, Z n , Mo , Sb, Te, I, Cs, Ba, and Ce; and Reference 14 for Zr and Nb.

+ Crop to soil concentration factors, in (pci/kg garden vegetation) per (pci/kg soil) . They are obtained from Reference 3 (Table E-1),

except as follows: Reference 2 (Table C-5) for Be and Sb.

O 4

2-36 Gen. Rev. 13 4

I l

l FNP-0-M-011 Table 2-7. Adult Ingestion Dose Factors Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI 3

H-3 No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 i

Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 1 Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 Mn-56 No Data 1.15E-07 2.04E-08 No Data 1.46E-07 No Data 3.67E-06 Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91E-06 No Data No Data 2.85E-06 3.40E-05 Co-58 No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51E-05 Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-63 1.30E-04 9.01E-06 4.36E-06 No Data No Data No Data 1.88E-06 Ni-65 5.28E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 ,

Cu-64 No Data 8.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 Zn-65 4.84E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-0E 1.97E-08 1.37E-09 No Data 1.28E-08 No Data 2.96E-09 .

Br-83 No Data No Data 4.02E-08 No Data No Data No Data 5.79E-08 Br-84 No Data No Data 5.21E-08 No Data No Data No Data 4.09E-13 Br-85 No Data No Data 2.14E-09 Ho Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.21E-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4.01E-08 2.82E-08 No Data No Data No Data 2.33E-21 Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data Po Data 4.94E-05 Sr-90 7.58E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 I

Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 All values are in (mrem /pci ingested) . They are obtained from Referenre 3 (Table E-11), except as follows: Reference 2 (Table A-3) far Rh-105, sb-124, and Sb-125.

(_

2-37 Gen. Rev. 13 i

l l

I

- . _ . =-

b FNP-0-M-011 Table 2-7 (contd). Adult Ingestion Dose Factors

! (Q

.J Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 i

Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-11 No Data No Data No Data 1.48E-05 Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-05 l

Ho-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 .

Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.1CE-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 O

2-38 Gen. Rev. 13

FNP-0-M-011 Table 2-7 (contd). Adult Ingestion Dose Factors Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 s

l I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 l 4

I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 Cs-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 2.59E-06 1 I

Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 1

1 Cs-137 7.97E-05 1.09E-04 7.14E-09 No Data 3.70E-05 1.23E-05 2.11E-06 4

Cs-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data 6.46E-11 3.92E-11 1.72E-07 4

('N Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 i l

La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 1 La-142 1.28E-10 5.82E-11 1.45E-11 No Data No Data No Data 4.25E-07 Ce-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05

, Co-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 4 W-187 1.03E-07 8.61E-08 3.01E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 i

O 2-39 Gen. Rev. 13

b FNP-0-M-011 f Table 2-8. Site-Related Ingestion Dose Factors, A;7 b

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.54E-01 2.54E-01 2.54E-01 2.54E-01 2.54E-01 2.54E-01 Na-24 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 Cr-51 0.00 0.00 1.25E+00 7.45E-01 2.74E-01 1.65E+00 3.13E+02 Mn-54 0.00 2.28E+02 4.34E+01 0.00 6.77E+01 0.00 6.97E+02 Mn-56 0.00 8.69E-03 1.54E-03 0.00 1.10E-02 0.00 2.77E-01 Fe-55 6.58E+03 4.55E+03 1.06E+03 0.00 0.00 2.54E+03 2.61E+03 Fe-59 1.02E+04 2.41E+04 9.22 +03 0.00 0.00 6.72E+03 8.02E+04 co-58 0.00 1.78E+02 3.99E+02 0.00 0.00 0.00 3.61E+03 Co-60 0.00 5.17E+02 1.14E+03 0.00 0.00 0.00 9.71E+03 Ni-63 3.14E+04 2.18E+03 1.05E+03 0.00 0.00 0.00 4.54E+02 Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 8.07E+00 3.79E+00 0.00 2.04E+01 0.00 6.88E+02 Zn -65 1.17E+03 3.71E+03 1.68E+03 0.00 2.48E+03 0.00 2.34E+03 Zn-69 3.94E-08 7.54E-08 5.24E-09 0.00 4.90E-08 0.00 1.13E-08 l

Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02 Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61E-18 l Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1 Rb-86 0.00 9.74E+04 4.54E+04 0.00 0.00 0.00 1.92E+04 1 Rb-88 0.00 0.00 0.00 0.00 0.00 0.00 0.00 .

Rb-89 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sr-89 2.23E+04 0.00 6.41E+02 0.00 0.00 0.00 3.58E+03 Sr-90 5.61E+05 0.00 1.38E+05 0.00 0.00 0.00 1.62E+04 Sr-91 7.07E+01 0.00 2.86E+00 0.00 0.00 0.00 3.37E+02 Sr-92 3.33E-01 0.00 1.44E-02 0.00 0.00 0.00 6.60E+00 Y-90 4.47E-01 0.00 1.20E-02 0.00 0.00 0.00 4.74E+03 Y-91m 1.04E-11 0.00 4.01E-13 0.00 0.00 0.00 3.04E-11 Y-91 8.58E+00 0.00 2.30E-01 0.00 0.00 0.00 4.72E+03 Y-92 4.60E-04 0.00 1.35E-05 0.00 0.00 0.00 8.07E+00 Y-93 3.09E-02 0.00 8.54E-04 0.00 0.00 0.00 9.81E+02 Zr-95 1.45E+01 4.64E+00 3.14E+00 0.00 7.27E+00 0.00 1.47E+04 Zr-97 3.01E-01 6.07E-02 2.77E-02 0.00 9.16E-02 0.00 1.88E+04 Nb-95 1.47E+00 8.17E-01 4.39E-01 0.00 8.08E-01 0.00 4.96E+03 l Mo-99 0.00 8.03E+02 1.53E+02 0.00 1.82E+03 0.00 1.86E+03 l Tc-99m 5.60E-04 1.58E-03l2.02E-02 0.00 2.40E-02 7.76E-04 9.37E-01 l

l l

l All values are in (mrom.mL)/(h yci). They are calculated using equation (2.14), and data from Table 2-5, Table 2-6, and Table 2-7.

When "No Data" is shown for a radionuclide-organ combination in p Table 2-7, A;7 factors in this table are presented as zero.

b 2-40 Cen. Rev. 13 l

l u J

l l

l l

PNP-0-M-011 Table 2-8 (contd). Site-Related Ittgestion Dose Factors, Air

, b

< v Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-101 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Ru-103 4.65E+00 0.00 2.00E+00 0.00 1.77E+01 0.00 5.42E+02 Ru-105 8.71E-03 0.00 3.44E-03 0.00 1.13E-01 0.00 5.33E+00 Ru-106 7.14E+01 0.00 9.03E+00 0.00 1.38E+02 0.00 4.62E+03 Rh-105 1.84E+00 1.34E+00 8.80E-01 0.00 5.68E+00 0.00 2.13E+02 Ag-110m 1.20E+00 1.11E+00 6.61E-01 0.00 2.19E+00 0.00 4.54E+02 sb-124 2.00E+03 3.77E+01 7.90E+02 4.83E+00 0.00 1.55E+03 5.66E+04 Sb-125 1.61E+03 1.73E+01 3.22E+02 1.43E+00 0.00 1.68E+05 1.42E+04 Te-125m 1.27E+04 4.60E+03 1.70E+03 3.81E+03 5.16E+04 0.00 5.06E+04 Te-127m 3.22E+04 1.15E+04 3.93E+03 8.23E+03 1.31E+05 0.00 1.08E+05 Te-127 8.89E+01 3.19E+01 1.92E+01 6.59E+01 3.62E+02 0.00 7.01E+03 Te-129m 5.40E+04 2.01E+04 8.54E+03 1.85E+04 2.25E+05 0.00 2.72E+05 j Te-129 8.89E-05 3.34E-05 2.17E-05 6.82E-05 3.74E-04 0.00 6.71E-05 Te-131m 4.76E+03 2.33E+03 1.94E+03 3.69E+03 2.36E+04 0.00 2.31E+05 Te-131 4.32E-16 1.80E-16 1.36E-16 3.55E-16 1.89E-15 0.00 6.12E-17 4

Te-132 9,75E+03 6.31E+03 5.92E+03 6.97E+03 6.0BE+04 0.00 2.98E+05 I-130 9.44E+00 2.78E+01 1.10E+01 2.36E+03 4.34E+01 0.00 2.40E+01 l I-131 1.86E+02 2.66E+02 1.52E+02 8.71E+04 4.56E+02 0.00 7.01E+01 I-132 7.02E-03 1.88E-02 6.57E-03 6.57E-01 2.99E-02 0.00 3.53E-03 l I-133 3.06E+01 5.33E+01 1.62E+01 7.83E+03 9.30E+01

/]

V I-134 2.91E-08 7.92E-08 2.83E-08 1.37E-06 1.26E-07 0.00 4.79E+01 0.00 6.90E-11 ,

I-135 1.71E+00 4.49E+00 1.66E+00 2.96E+02 7.20E+00 0.00 5.07E+00 co-134 2.99E+04 7.11E+04 5.81E+04 0.00 2.30E+04 7.64E+03 1.24E+03 Cs-136 2.96E+03 1.17E+04 8.42E+03 0.00 6.51E+03 8.92E+02 1.33E+03 Cs-137 3.83E+04 5.24E+04 3.43E+04 0.00 1.78E+04 5.92E+03 1.01E+03 Cs-138 9.12E-13 1.80E-12 8.92E-13 0.00 1.32E-12 1.31E-13 7.68E-18 Ba-139 5.64E-05 4.02E-08 1.65E-06 0.00 3.76E-08 2.28E-OS 1.00E-04 Ba-140 1.86E+03 2.34E+00 1.22E+02 0.00 7.95E-01 1.34E+00 3.83E+03 Ba-141 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Ba-142 0.00 0.00 0.00 0.00 0.00 0.00 0.00 La-140 9.93E-02 5.01E-02 1.32E-02 0.00 0.00 0.00 3.68E+03 La-142 2.19E-07 9.96E-08 2.48E-08 0.00 0.00 0.00 7.27E-04 Ce-141 4.40E+00 2.98E+00 3.38E-01 0.00 1.38Ee00 0.00 1.14E+04 Ce-143 4.77E-01 3.53E+02 3.91E-02 0.00 1.55E-01 0.00 1.32E+04 Co-144 2.34E+02 9.79E+01 1.26E+01 0.00 5.80E+01 0.00 7.91E+04 Pr-143 5.33E-01 2.14E-01 2.64E-02 0.00 1.23E-01 0.00 2.33E+03 Pr-144 0.00 0.00 0.00 0.00 0.00 0.00 0.00 l Nd-147 3.59E-01 4.15E-01 2.48E-02 0.00 2.43E-01 0.00 1.99E+03 W-187 1.47E+02 1.23E+02 4.30E+01 0.00 0.00 0.00 4.03E+04 Np-239 2.15E-02 2.11E-03 1.17E-03 0.00 6.60E-03 0.00 4.34E+02

/~'N 2-41 Gen. Rev. 13

FNP-0-M-011

' 2.5 LIQUID EFFLUENT DOSE PROJECTIONS i

O 2.5.1 Thirtv-One Day Dose Proiections l

In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT l SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 l 1

days; this applies during periods in which a discharge to UNRESTRICTED AREAS of j liquid effluents containing radioactive materials occurs or is expected. I l

. Projected 31-day doses to individuals due to liquid effluents may be determined i

! as follows:

f.

,p ,

D 7p = x 31 + D ra (2.17) i where:

4 i

D,p = the projected dose to the total body or organ r, for the next 31 days of liquid releases.

l Dye = the cumulative dose to the total body or organ r, for liquid

( releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

, t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).

I i D ra = the anticipated dose contribution to the total body or any organ 7, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in

{ addition to routine liquid effluents. If only routine liquid af fluents are anticipated, D 7, may be set to zero.

r 4

2.E.2 Dose Proiections for Soeci{.ie Releases

}

Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections- due to liquid releases, follow the

methodology,of Section 2.4, using sample analysis results for the source to be b released, and parameter values expected to exist during the release period.

2 2-42 Gen. Rev. 13

. - _ _ ~ . . . . _ - . . . ~ - . . . . . . - . - - - - . - . - - . . . - - _ . - . . . - . . _ . . - - . . - . - - . - .

i FNP-0-M-011 I 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS ih' v

, The following symbolic terms are used in the presentation of liquid effluent 4

calculations in the sub-sections above.

Section of Iggm Definition Initial Use 4

Ap = the adjustment factor used in calculating the J effluent monitor setpoint for liquid release pathway i- p: the ratio of the assured dilution to the required 1

J dilution [unitiess). 2.3.2.2 3

i ADF = the assured dilution facter Jor a planned release

[unitiess). 2.3.2.2 1

AFp = the dilution allocation factor for liquid release pathway p (unitiess). 2.3.2.2 l

l

! Aj7 = the site-related adult ingestion dose commitment f factor, for the total body or for any organ 7, due to a

4 identified radionuclide 1 ((mram mL)/(h*pCi)). The i values.of Ajf are listed in Table 2-8. 2.4.1 i:

I i Bjy = the crop to soil concentration factor applicable to i radionuclide i, ((pci/kg garden vegetation)/(pci/kg l soi1. Values are listed in Table 2-6. 2.4.3 5 l Brj = the bioaccumulation factor for radionuclide i for freshwater fish [(pci/kg)/(pci/L)). Values are listed in Table 2-6. 2.4.2 e= the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line, prior to dilution and subsequent release [pci/mL). 2.3.2.1 ep = the calculated effluent radioactivity monitor setpoint for liquid release pathway p [pci/mL). 2.3.2.2 c, = the gross concentration - of alpha emitters in the (~ liquid waste as measured in the applicable composite sample (pci/mL). 2.3.2.2 2-43 Gen. Rev. 13

FNP-0-M-011 Section of ( IRIE Definition Initial Use CECL = the Ef fluent Concentration Limit stated in 10 CFR 20, l Appendix B, Table 2, Column 2 (pci/mL). 2.3.2.1 1 Cr = the concentration of Fe-55 in the liquid waste as j measured in the applicable composite sample i (pci/mL). 2.3.2.2 4 Cg = the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed < on the applicable pre-release waste sample [pci/mL). 2.3.2.2 i l C; = the measured concentration of radionuclide i in a i sample of liquid effluent [pci/mL). 2.3.2.2 ) Cj= i the average concentration of radionuclide i in l I undiluted liquid effluent during time period 1 (O (pci/mL). 2.4.1

V

cir = the measured concentration of radionuclide i in j release pathway r that is contributing to radio-activity in the dilution stream (pci/mL). 2.3.2.2

C, = the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste as measured in the applicable composite sample (pci/mL). 2.3.2.2 Ct= the concentration of H-3 in the liquid waste as measured in the applicable composite sample

[yci/mL}. 2 . 3.1.2. CF y = the concentration factor for radionuclide i in irrigated garden vegetation ((pci/kg)/(pci/L)]. 2.4.2 D, = the dilution factor from the near field of the discharge structurer to the potable water intake

   %            location (unitiess).                                           2.4.2 2-44                         Gen. Rev. 13 I

1 l

                                                                             ~- . . _ .   .. ..

FNP-0-M-011 i Section of 1 tim Definition Initial Use D7 = the cumulative dose commitment to the total body or to any organ r, due to radioactivity in liquid effluents released during a given time period (mrom). 2.4.1 i l D, = the anticipated dose contribution to the total body or any organ r, due to any planned activities during the next 31-day period [mram). 2.5.1 l D rc = the cumulative dose to the total body or organ r, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mram). 2.5.1 D 7p

              =   the projected dose to the total body or organ r, for the next 31 days of liquid releases [mram).                       2.5.1 l

DFt r = the dose conversion factor for radionuclide i for p

 .( )             adults, in organ T [mrom/pci). Values are listed in Table 2-7.                                                        2.4.2

• ECLj = the liquid Ef fluent concentration Limit for radio-  !

nuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 [pci/mL). 2.3.2.2 f= the affluent flowrate at the location of the radio-activity monitor [gpm). 2.3.2.1 f,p = the anticipated actual discharge flowrate for a planned release from liquid release pathway p [gpm). 2.3.2.2 fg = the fraction of the year that garden vegetation is irrigated [unitiess). 2.4.3 f mp

                =            the maximum permissible effluent discharge flowrate- for- release 2.3.2.2 a
     ]                       pathway p [gpm).

2-45 Gen. Rev. 13

                   . . . - - - - . - . . ~ . .. . _ _ . .                      . - - . . - . . - . . . . . - - . . - . - - - . .              -. -

+ f f' FNP-0-M-011

i. ~

! Section of IEEE Definition Initlai Use t i i fr_ = the effluent discharge flowrate of release pathway r (gPm). 2.3.2.2 i fg= the average undiluted liquid waste flowrate actually l observed -during the period of a liquid release { (gpm). 2.4.1 j F= the dilution stream flowrate which can be assured f prior to the release point'to the UNRESTRICTED AREA [gpm). 2.3.2.1 Fd= the entire assured dilution flowrate for the plant site during the release period (gpm). 2.3.2.2 Fdp = the dilution flowrate allocated to release pathway p [gpm), 2.3.2.2 Fg = , the near-field average dilution factor in the

    .                            receiving                water   of    the   UNRESTRICTED                                  AREA

[unitless). 2.4.1 l Ft= the average dilution stream flowrate actually observed during the period of a liquid release (gym]. 2.4.1 l

          .I =                   the average irrigation rate during the growing season                                                               I 2

[L/(m *h)). 2.4.3 Ly = ths water content of leafy garden vegetation edible parts (L/kg). 2.4.3 M= the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage [unitiess). 2.4.3 P= the effective surface density of soil (kg/m 2). 2.4.3 O 2-46 Gen. Rev. 13

1 l l

                                                       .                FNP-0-M-0M Section of

, Term Definition Initial Use l r= the fraction of irrigation-deposited activity l retained on the edible portions of leafy garden vegetation. 2.4.3 RDF = the required dilution factors the minimum ratio by which liquid ef fluent must be diluted before reacning the UNRESTRICTED AREA, in order to ensure that the j limits of Section 2.1.2 are not exceeded [unitiess). 2.3.2.2 RDr y = the RDF for a liquid release due only to its concen-tration of gamma-emitting radionuclides (unitiess). 2.3.2.2 RDF ny = the RDF for a liquid release due only to its concen- j tration of non-gamma-emitting radionuclides (unitiess). 2.3.2.2 1 i SF = the safety factor selected to compensate for  ! O V statistical fluctuations and errors of measurement [unitiess). 2.3.2.2 t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration. 2.5.1 tb= the period of long-term buildup of activity in soil [h). 2.4.3 t, = the period of leafy garden vegetation exposure during the growing season [h). 2.4.3 tg = the transit time from release to receptor for fish consumption [h). 2.4.2 th= the time between harvest of garden vegetation and human consumption [h). 2.4.3 t, a the transit time f~ rom refense to receptor for potable V water consumption [h). 2.4.2 2-47 Gen. Rev. 13

   . _ _ . .   . _ _ _ _ _ _ . . . ~ . . .        _ - -.  . - _ .          .         . - . _ . . _ _ . . _ _ . _ _ _ _ .      .. . _ _ . _ _ .       . _ . ___ __

FNP-0-M-011 , i Section of l IgIm Definition Initial Use $ TF = the tolerance f actor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate effluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2 [unitiess); , the tolerance factor must not exceed a value of 10. 3.3.2.2 Ug = the adult rate of fish consumption [kg/y). 2.4.2 {

Uy = the adult consumption rate for irrigated garden vegetation [kg/y). 2.4.2 d

U, = the adult drinking water consumption rate applicable l to the plant site [L/y). 2.4.2 Y, = the areal density (agricultural productivity) of leafy garden vegetation [kg/m 2). 2.4.3 Z= the applicable dilution factor for the receiving , water body, in the near field of the discharge structure, during the period of radioactivity release [unitiess). 2.4.1 Atg = the length of time period 1, over which C g iand Fg are averaged for liquid releases (h). 2.4.1 1 AE

                    =                      the effective removal rate for activity deposited on crop leaves [h-I).                                                                                  2.4.3 Ag =                          the decay constant for radionuclide i [h-I).                                                        2.4.2 1, =                         the rate constant for removal of activity from plant leaves by weathering [h-I) .                                                                       2.4.3 i

2-48 Gen. Rev. 13 l

_ . . . - - , - - - - - _ - _ _ . -- . - - . . , - - _ . - . - . - . - - . . . . ~ - . . _ . - _ - _ - - 4 4 i FNP-O-M-011 l' CHAPTER 3 i f GASEOUS EFFLUENTS i'  :

                                                                                                                                              ^

i 3.1 LIMITS OF OPERATION ., 1 The following Limits of Operation implement requirements established by Technical l specifications section 6.0. Terms printed in all capital letters are defined in l . Chapter 10. 1 3.1.1 cameous Effluent Monitorina Instrumentation control , ! I i ! In accordance with Technical specification 6.8.3.e(i), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE 4 with their alare/ trip setpoints set to ensure.that the limits of section 3.1.2.a l i 1' are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with section 3.3. i 3.1.1.1 Applicability I 1 These limits apply as shown in Table 3-1. [ ] i I l 3.1.1.2 Actions ,] i with a radioactive gaseous effluent monitoring instrumentation channel alarm / trip ! setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous affluents monitored by the affected channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of section 3.1.2.a are met. with less than the minimum number of radioactive gaseous effluent monitoring !' instrumentation channels OPERABLE, take the ACTION shown in Table 3-1. i,

f. This control does not affact shutdown requirements or MODE changes.

3.1.1.3 surveillance Requirements

Each radioactive gaseous offluent monitoring instrumentation channel shall be i demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL j CALIBRATION- and CHANNEL FUNCTIONAL TEST operations at' the frequenciesr shown irr Table 3-2.

i O IO 2 a. 3-1 Gen. Rev. 13 l l I

l FNP-0-M-011 i 3.1.1.4 Basis I The radioactive gaseous effluent instrumentation is provided to monitor and i control, as applicable, the releases of radioactive materials in gaseous c,'fluents during actual or potential releases of gaseous effluents. The A24rm/ Trip Setpoints for these instruments shall be calculated and adjusted in , accordance with the methodology and parameters in Section 3.3 to ensure that the 1 l alarm / trip will occur prior to exceeding the limits of Section 3.1.2.a. The l OPERABILITY and use of this instrumentation is consistent with the requirements of General Design criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. l 1 I l l l

                                                                                                     )

l l

                                                                                                     )

l \ I l I l 1 \ 3-2 Gen. Rev. 13 i

FNP-O-M-011 T able 3-1. Radic2etive Gaseous Effluent Monitoring Instrumentation O OPERABILITY Requirementa b Minimum Channels Instrument OPERABLE Applicability ACTION i

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor I

(RE-14 or RE-22) 1 At all times 37 8

b. Iodine Sampler 1 At all times 39

c. Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36

3. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity

[V) Monitor (RE-14), with Alarm and Automatic ' Termination of Release 1 At all times 35 W ?

a. For continuous releases.

i

b. All requirements in this table apply to each unit.

d 3 1 3-3 Gen. Rev. 13

FNP-0-M-011 ("' Table 3-1 (contd). Notation for Table 3 ACTION Statements ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLfe: requirement, the contents of the tank (s) may be released to the environment for up to 14 days 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 discharge line valving, and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway. ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue for up to 30 days provided the flowrate is estimated at least once per 4 hours. Normal building ventilation may continue providcd the flowrate is estimated once per 4 hours. ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours. Normal building ventilation may continue provided grab samples of this pathway are taken once per 8 hours and analyzed for gross activity once per 24 hours. l l ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via the af f ected . pathway may continue for up to 30 days provided samples are continuously I collected with auxiliary sampling equipment as required in Table 3-3. O

                          ~

3-4 Gen. Rev. 13

d FNP-0-M-011 7- Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements

 . \

1 Surveillance Requiremente d Instrument CRANNEL CRANNEL CHANNEL SOURCE CALIBRA- FUNCTIONAL CHECK CHECK TION TEST MODESC

1. Steam Jet Air Ejector l

Noble Gas Activity Monitor (RE-15) D M Rb ga(2) 1,2,3,4

2. Plant Vent Stack

a. Noble Gas Activity Monitor RE-14 D M Rb ga(1,2)

RE-22 D M Rb ga(2) g

b. Iodine Sampler W NA NA NA All

c. Particulate Sampler W NA NA NA All

, d. Flowrate Monitor D NA R Q All

a. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section

, 10.2): f3 () (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room annunciation occur if , any of the following conditions exists: (a) Instrument indicates measured levuls above the alarm / trip setpoint; (b) Loss of control power; or (c) Loss of instrument power. (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room annunciation occurs if any of the following conditions exista: (a) Instrument indicates a downscale failure; or (b) Instrument controle not set in the OPERATE mode. l

b. The initial CHANNEL CALIBRATION shall be performed using one or more of the referenca a*madards certified by the National Institute of Standards an& Twehnology, or usinty standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. For any subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

c. MODES in which surveillance is required. "All" means "At all times."

d. All requirements in this table apply to each unit.

O 3-5 Gen. Rev. 13

      . _ . . . _ . - .      . _     m   .   .___.m.-__.-_.m          - _ - - _ _ _ . . _           _ _ > _ _. _.._m         . . _

l

                  -                                                                            FNP-0-M-011 A                   3.1.2   Caseous Effluent Dose Rate Control                                                                   '

I. J In accordance with Technical Specifications 5.8.3.e(iii) and 6.8.3.e(vii), the ' licensee shall conduct operations so that the dose rates due to radioactive , materials released in gaseous ef fluents from the site to areas at and beyond the  ; j- SITE BOUNDARY (see Figure 10-1) are limited as follows: i  :

 ;                 a. For noble gases:     Less than or equal to a dose rate of 500 mrom/y to the                                  I
;                         total body and   less than o. equi to a dose rate of 3000 mrom/y to the

, skin, and

b. ~For Iodine-131, Iodine-133, tritium, and for all radionuclides in j, particulate form with half-lives greater than 8 days: Less than or equal i to a dose rate of 1500 mrem /y to any organ.

3.1.2.1 Applicability 1 This limit applies at all times. 3.1.2.2 Actiona With a dose rate due to radioactive material releaseo in gaseous effluents

exceeding the limit stated in Section 3.1.2, immediately decrease the release 4

rate to within the stated limit. This control does not affect shutdown requirements or MODE changes. l 3.1.2.3 Surveillance Requirements

                 -The dose rates due to radioactive materials in areas at or beyond the SITE                                           i BOUNDARY due to releases of gaseous effluents shall be determined to be within f                   the above limits, in accordance with the methods and procedures in Section 3.4.1, by obtaining representative samples and performing analyses in accordance with the sampling and analysis progres specified in Table 3-3.

l- 3.1.2.4 Bacis This control is provided to ensure that gaseous effluent dose rates will be maintained within the limite that historically have provided reasonable assurance that radioactive material discharged in gaseous effluents will not result in a , dose to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS + 3-6 Gen. Rev. 13

FNP-0-M 011 OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. [ The dose rate limit for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days specifically applies to dose rates to a child via the inhalation pathway. This control applies to the release of gaseosa ef fluents from all reactors at the cite. l l l i I O . I O 3-7 Gen. Rev. 13

FNP-0-M-011 Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program Q ,) Sampling and Analysis Requirements a,b MINIMUM DETECTABLE Gaseous Minimum Type of CONCENTRA-Release Sampling Analysis Activity TION (MDC) Type FREQUENCY FREQUENCY Analysis (pCi/mL) P PRINCIPAL GAMMA 1 E-4 Waste Gas p Decay Tank Eac-- Iank Each Tank EMITTERS Gral ,, ample

                             ;4                         PRINCIPAL GAMMA       1 E-4 Containment                             PC      EMITTERS Each Purge Purge Grab Sample         Each Purge H-3                   1 E-6 Condenser                                      PRINCIPAL GAMhA       1 E-4 Steam Jet                                      EMITTERS Air           Mc d f                                                         i Mc Ejector,    Grab Sample                       H-3                   1 E-6       {

Plant Vent Stack ) i CONTINUOUS 8 We I-131 1 E-12 O Charcoal or Charcoal or Silver Silver Zeolite I-133 1 E-10 i'eolite Sample a We PRINCIPAL GAMMA 1 E-11 CONTINUOUSE Particulate EMITTERS Sample Plant Vent M Gross Alpha 1 E-11 E Contal ent 03NTINUOUSI p Purge Sample Q Sr-89, Sr-90 1 E-11 CONTINUOUSg COMPOSITE Particulate Sample N Gas Noble Gases 1 E-6 CONTINUOUSg (Gross Beta and Gamma) O 3-8 Gen. Rev. 13

FNP-0-M-011 Table *-3 (contd). Notation for Table 3-3

a. All requirements in this table apply to each unit. Deviation from the MDC requirements of this table shall be reported in accordance with Section 7.2. Deviation from the composite sampling requirements of this table shall be reported in accordance with Section 7.2.

b. Terms printed in all capital letters are defined in Chapter 10.

c. Analyses shall also be performed following shutdown from greater than or equal to 15% RATED THERMAL POWER, startup to greater than or equal to 15%

RATED THERMAL POWER, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one-hour period. j d. Tritium grab samples shall be taken from the plant vent stack at least once per 24 hours when the refueling canal is flooded.

e. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours af ter changing (or after removal from sampler) .

Sampling shall also be performed at least once per 24 hours for at least

2 days following each shutdown from greater than or equal to 15% RATED THERMAL POWER, startup to greater than or equal to 15% RATED THERMAL POWER, or THERHAL POWER change exceeding 15% of RATED THERMAL POWER in one hour, and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding MDC may be increased by a factor of 10.

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

g. The ratio of the senple flowrate to the sampled stream flowrate shall be known for the time period coverad by each dose or dose rate calculacion made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.

l l O 3-9 Gen. Rev. 13

FNP-0-M-011 3.1.3 cameous Effluent Air Dose Control In accordance with Technical Specifications 6.8.3.e(v) and 6.8.3.e(viii), the air dose due to noble gases relaased in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to { the followings i

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

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

3.1.3.1 Applicability This limit applies at all times. 3.1.3.2 Actions With the calculated air dose from radioactivt. a gases in gaseous effluents exceeding any of the above limits, prepare and suumit to the Nuclear Regulatory commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Repott which identifies the cause(s) for exceeding the limit (s); defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases.of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3. This control does not affect shutdown requirements or MODE changes. 3.1.3.3 Surveillance Requirements cumulative air dose contributions from noble gas radionuclides released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for l the current calendar quarter and current calendar year, shall be determined in accordance with section 3.4.2 at least once per 31 days. i 3.1.3.4 Basis This control is provided to implement the requirements of Sections II.8, III.A and IV.A of Appendix I, 10 CFR Part 50. Section 3.1.3 implements the guides set l forth in Section II.B of Appendix I. The ACTION statements in Section 3.1.3.2 provide the required operating thxibility and at the same time implement the 3-10 Gen. Rev. 13 j

FNP-0-M-011 O guides set forth in Section IV.A of Appendix I, assuring that the releases of radioactive material in gaseous af fluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III. A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the '.ctual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous ' effluents are consistent with the methodology provided in Regulatory Guide 1.109 I (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOUNDARY are based upon the historical annual average atmospheric conditions.  ; 1 l I l

                                                                                         )

l l O l G \ l O V 3-11 Gen. Rev. 13

     ._. ._- . _ .            .-              _ . -   _ .- -.        . . _ -   .    .. - - - - - . .. -              . ~

4 FNP-0-M-011 a 3.1.4 control on Gaseous Effluent Dose to a Member of the Publiq In accordance with Technical Specifications 6.8.3.e(v) and 6.8.3.e(ix), the dose to a KEMBER OF THE PUBLIC from I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see

Figure 10-1) shall be limited to the following

a. During any calendar quarters Less than or equal to 7.5 mrem to any organ, and 1

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

3.1.4.1 Applicability j This limit applies at all times. l 4 3.1.4.2 Actions With the calculated dose from the release of I-131, I-133, tritium, or radio-nuclides in particulate form with half-lives greater than 8 days, in gaseous

( effluents exceeding any of the above limits, prepare and submit to the Nuclear I

, Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Spacial Report which identifies the cause(s) for exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radiciodines and radionuclides in particulate form with half-lives greater than 8 days in gaseous affluents; and defines proposed corrective actions to assure that subsequent releases will be in compliance with the limits stated in Section 3.1.4. l This control does not affect shutdown requirements or MODE changes. l 3.1.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMBER OF THE PUBLIC from I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous ef fluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days. (-~

                                         ~

3-12 Gen. Rev. 13

FNP-O-M-011 () 3.1.4.4 Basis This control is provided to implement the requirements of Section II.C, III. A and IV.A of Appendix I, 10 CFR Part 50. The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section 3.1.4.2 provide tr.e required operat..ng 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 ef fluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The calculational methods specified in the surveillance Requirements of Section 3.1.4.3 implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be chown by calculational procedures based on models and data, such that the actual cxposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The calculational methods in Section 3.4.3 for calculating the doses due to the actual releases of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses based upon the historical annual average atmospheric conditions. The release specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the [ cxisting radionaclide pathways to man, in the areas at and beyond the SITE ) BOUNDARY. The pathways which were examined in the development of these I calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation wiyt. 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. 1 t 1 4 t v i 3-13 Gen. Rev. 13

_. . _ . , _ _.___m.____-. _ _ _ _ _ _. _ . . _ ..~. __ .._m _ m . _ _ . _ . _ . .

                                                                                                                                         )

i FNP-0-M-011 l

                                                                                                                                         )

3.1.5 Gaseous Radwaste Treatment Svetaru Control l l In accordance with Technical Specification 6.8.3.e(vi), the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous wastes prior to their discharge ,; hen the projected air doses due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.2 mrad for gamma radiation or 0.4 mrad for beta radiation in 31 days. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce e radioactive materials in gaseous wastes prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, to areas beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.3 mrom to any organ of a MEMBER OF THE PUBLIC in 31 days. 3.1.5.1 Applicability These limits apply at all times. 3.1.5.2 Actions with gaseous waste being discharged without treatment and in excess of the limits . in Section 3.1.5, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Techr.ical Specification 6.9.2, a Special Report which includes the following information:

a. Identification of the inoperable equipment or subsystem and the reason for inoperability,

b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and 4

! c. Summary description of action (s) taken to prevent a recurrence. i ! This control does not affect shutdown requirements or MODE changes. 7 , i i 4 3-14 Gen. Rev. 13 4

t J l FNP-0-M-011

3.1.5.3 Surveillance Requirements Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days, in accordance with Section 3.5.1, when the GASEOUS RADWASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE:

a. by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS RADWASTE TREATMENT SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREATMENT SYSTEM), or

b. by operating the GASEOUS RADWASTE TREATMENT SYSTEM equipment and the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

i 3.1.5.4 Basis The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use l whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive , l materials in gaseoue of fluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 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 theos systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. , I I O  ! 3-15 Gen. Rev. 13

                                                                                                                                  ]

FNP-0-M-011 ( Q) 3.1.6 MAJOR CHANGES to the CASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM anj the VENTILATION EXHAUST TREATMENT SYSTEM Licensee-initiated MAJOR CHANGES to the GASEOUS RADIOACTIVE WASTE TRE?.!:t?NT SYSTEM or the VENTILATION EXHAUST' TREATMENT SYSTEM:

a. Shall be reported to the Nu: lear Regulatory Commission in the Annual Radioactive Effluents Release Report for the period in which the change was implemented, in accordance with Section 7.2.2.7.

b. Shall become effective upon review and approval in accordance with Technical Specification 6.5.3.1.

A j i l 1 I I l O 3-16 Gen. Rev. 13

w-I P FNP-0-M-011

     ,      3.2   GASEOUS RADWASTE TREATMENT SYSTEM e,m At the Farley Nuclear Plant, there are six designated points where radioactivity may be released to the atmosphere in gaseous discharges: the Unit 1 and Unit 2 Plant Vent Stacks; the Unit 1 and Unit 2 Turbine Building Vents (steam jet air ejectors); and the Unit 1 and Unit 2 Integrated Leak Rate Test (ILRT) Vents. Of these six, only four are routine release pathways, since ILRT Vent releases are performed only infrequently.

Figure 3-1 gives schematic diagrams of the Waste Gas Treatment Systems and the Ventilation Systems (Reference 7). Dischkrges from the two reactor units are separated, with no shared systems. In each unit, Containment Purge and Waste Gas Decay Tank effluents are discharged through the respective Plant Vent, and are treated as contributions to the on-going Plant Vent CONTINUOUS release. Although Waste Gas Decay Tank effluents are released via the Plant Vent Stack, they are tracked separately and accounted for as BATCH releases. Table 3-4 summarizes the release height and release type characteristics of the various release pathways and source streams. Chapter 8 discusses the calculation of atmospheric dispersion parameters using the ground-level and mixed-mode (i.e., split-wake) models.

   ,f3,
     \

As established in Section 3.1.1, gaseous ef fluent monitor setpoints are required for the noble gas monitors on the two Plant Vents and the two Turbine Building Vents (steam jet air ejectors). Waste Gas Treatment System discharges are not monitored separately during release, but are sampled prior to release and are monitored by the downstream Plant Vent monitors during release. ILRT discharges tre not monitored during release, but are sampled prior to release; the ILRT Vent may be assigned an appropriate allocation factor during the release period, and dose calculations may be based on estimates of the activity concentration and the volume of air released. Sampling and analysis of both these release pathways must be suf ficient to ensure that the gaseous ef fluent dose limits specified in Section 3.1.3 and Section 3.1.4 are not exceeded. V

                                                              ^

3-17 Cen. Rev. 13

FNP-O-M-011 O l Wasta Gas Decay Tanks l 1

                 )[                                   II                                                  ][

i ll lE )l l[ lk )l )[ ][ l l 7 8 1 2 3 4 5 6 l 1 i ir 1r ir 1r i r i r ir ir l dk dk JL Mk d k d L JL JL l I sr 1r ir db dL JL

]l Waste Gas RE-13 MONITOR Compressor

                                   ;;                                                       - PLANT VENT STACK 2                                    0                                  -1 l                    ILRT                                    '- RE-22 MONITORS

• VENT CONTAINMENT AUX.

] PURGE gtog AN ~~ MEN T EXHAUST' Q PLENUM az-2 1 MONITOR i' AUXILIARY  !

BUILDING i i

4 i RE-15 MONITOR TURBINE i TURBINE BUILDING r BUILDING VENT j (STEAM JET AIR EJECTOR) .

O ~

Figure 3-1. Schematic Diagram of Routine Release Sources and Release Points ! (Typical of Both Units) 3-18 Gen. Resr. 13

FNP-0-M-011 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS t 3.3.1 General Provisions Recardina Noble Gas Monitor Setooints Noble gas radioactivity monitor setpoints calculated in accordance with the methodology presented in this section are intended to ensure that the limits of Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint. If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should a significant inadvertent release occur. Section 3.1.1 establishes the requirements for gaseous affluent monitoring instrumentation, and Section 3.2 describes the VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM. From those sections, it can be seen that certain monitors are located on final release pathways, that is, streams that are being monitored immediately before being discharged from the O plant; the setpoint methodology for these monitors is presented in Section 3.3.2. Other monitors are located on source streams, that is, streams that merge with l other streams prior to passing a final monitor and being discharged; the setpoint methodology for these monitors is presented in Section 3.3.3. Table 3-4 I identifies which of these setpoint methodologies applies to each monitor. Some additional monitors with special setpoint requirements are discussed in Section 3.3.5. l t l l O 3-19 Gen. Rev. 13

            ..           . . .            .           . .  . .    -  - _ _  - - . _ _ _ . . . .- . ~ _ ~ . .

e FNP-0-E 011 Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies Final Release Fathways with no Monitored source Streams Release Elevation: Ground-level Unit 1 or Unit 2 Turbine Buildino Vent Release Type: CONTINUOUS Monitor 1RE-15 / 2RE-15 Setpoint Method: Section 3.3.2 Maximum Flowrate: 1060 cfm (5.00 E+05 mL/s) Unit 1 or Unit 2 ILRT Vent Release Type: BATCH l Monitor: None Satpoint Method None Maximum Flowrates Release-dependent Final Release Fathways with one or More Monitored source Streams Release Elevation: Mixed-Mode Unit 1 or Unit 2 Plant Vent Stack Release Type: CONTINUOUS Monitor: 1RE-14 / 2RE-14, and 1RE-22 / 2RE-22 Satpoint Method: Section 3.3.2 w Maximum Flowrate: 150,000 cfm (7.08 i E+07 mL/s)

      )           Source Streams Unit       1 or Unit 2 Containment Purce Release Type:               CONTINUOUS Monitor:                    1RE-24 / 2RE-24 Setpoint Method:            Section 3.3.3 is optional. See Section 3.3.5.

Maximum Flowrates Release-dependent Source Stream: Unit 1 or 'Jnit 2 Waste Gas Decav Tanks Release Type BATCH Monitor: None Setpoint Method: None Maximum Flowrates Release-dependent (176)vb values for Use in setpoint Calculations Grnund-Lawal Enlaammer 4.87 x 10-5 ,j,3 (s Sector] Mixed-Mode Releases: 1.08 x 104 s/m 3 (SSE Sector)

                                   ~~~

3-20 Gen. Rev. 13

  . ~ . _ _ _      . . _ _ _ . . - .       ._ .. . _ _ . _       .__.._._...____.-____m-____...                 . _        _ ___ m _- _ . _ _

FNP-0-H-011 s .O 3.3.2 setooint for the Final Noble Gas Monitor on Each Release Pathway I 3.3.2.1 overview of Method , 4 Gaseous effluent radioactivity monitors are intended to alarm prior to exceeding the limits of Section 3.1.2.a. Therefore, their alarm setpoints are established to ensure compliance with the following equations i e = the lesser of (3.1) AG

• SF
• X
• Rg j where:

1 e= the setpoint, in ci/mL, of the radioactivity monitor measuring the  ! concentration of radioactivity in the ef' fluent line prior to release. The setpoint represents a concentration which, if l exceeded, could result in dose rates exceeding the limits of  ; Section 3.1.2.a at or beyond the SITE BOUNDARY. AG = an administrative allocatt..a factor applied to divide the release ,

         .                           limit among all the gaseous release pathways at the site.                                                         I l

SF = the safety factor selected to compensate for statistical ,  ! fluctuations and errors of measurement. I X= the noble gas concentration for the releane under consideration. Rt= the ratio of the dose rate 15c.it for the total body, 500 mrem /y, to the dose rate to the total body for the conditions of the release under consideration. Rk= the ratio of the dose rate limit for the skin, 3000 mrem /y, to the dose rate to the skin for the conditions of the release under 1 1 consideration. i l Equation (3.1) shows the relationships of the critical parameters that determine j l the setpoint. However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology. 3-21 Gen. Rev. 13 L

FNP-0-M-011 The basic setpoint method presented below is applicable to the radioactivity monitor nearest the point of release for the release pathway. For monitors measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section 3.3.3 must be applied. 3.3.2.2 Setpoint calculation Steps Stoo 1: Determine the concentration, Xjy, of each noble gat .dionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analyris requirements of Section 3.1.2. Then sum these concentrations  ?.o determine the total noble gas concentration, 2: X;y. I Stoo 2: Determine R t , the ratio of the dose rate limit for the total body, 500 mrem /y, to the total body dose rate due to noble gases detected in the release under consideration, as follows: 500 (R7D)vb b (Kl

• O iv) (3.2) j i

where: 500 = the dose rate limit for the total body, 500 mrem /y. (176)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v. Table 3-4 includes an indication of what release elevation is applicable to j each release pathway; releese elevation determines the appropriate l value of (176)vb-Kg = the total-body dose factor due to gamma emissions from noble gas radionuclide-1, in (mrom/y)/(pci/m 3), from Table 3-5. Qjy = the release rate of noble gas radionuclide i from the release pathway under consideration, in pCi/s, calculated as the product of X;y and f,y, where: 4 f'

   \                   X y=       the concentration of noble gas radionuclide i for the particular release, in pci/mL.

3-22 Gen. Rev. 13

j e FNP-0-M-011 f,y = the maximum anticipated flowrate for release pathway v b during the period of the release under consideration, in k mL/s. 4 steo 3: Determine kR ,.the ratio of the dose rate limit for the skin, 3000 mrem /y, to the skin dose rate due to noble games detected in the release under consideration, as follows: l

                                         ,.                3000 (X75)vb b, [IL I*1*18)1

• O v) i '(3 3)

I where: l 1 3000 = the dose rate limit for the skin, 3000 mram/y.

Lj = the skin dose factor due to beta emissions from noble gas radio-

nuclide 1, in (mrom/y)/(yci/m 3), from Table 3-5.

I {' Mj = the air dose factor due to gamma emissions from noble gas radio-nuclide 1, in (mrad /y)/(pci/m 3), from Table 3-5. 1.1 = the facter to convert air dose in mrad to skin dose in mrom. , All other terms were defined previously.  ; 4 steo 4: Determine the maximum noble gas radioactivity monitor setpoint con-centration. l l Based on the values determined in previous steps, the radioactivity monitor l cetpoint'for the planned release is calculated to ensure that the limits of  ; Section 3.1.2.a will not be exceeded. Because the radioactivity monitor responda ,

     .primarily to radiation from, noble gas radionuclides, the monitor setpoint c,y is i

i based-on the concentration of all noble gases in the waste stream, as follows: i i I l l i l 3-23 Gen. Rev. 13

                          . -    =-                    . . _ _     .      .      .-         . .. . .        _ . _ - -

FNP-0-M-011 AG y

• SF
• X;y
• R g e ny = the lesser of (3*4)

AG y

• ST = [ X;y = Rg i

where: env = the calculated setpoint, in pci/mL, for the noble gas monitor serving gaseous release pathway v. AG y = the administrative allocation factor for gaseous release pathway v, applied to divide the release limit among; all the gaseous release pathways at the site. The allocation factor may be assigned any value between 0 and 1, under the condition that the sum of the allocation factors for all simultaneously active final release pathways at the entire plant site does not exceed 1. Alternative methods for - determination of AG y are presented in Section 3.3.4. V SF = the safety factor selected to compensate for statistical l fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1. A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired. 1 X;y = the measured concentration of noble gas radionuclide i in gaseous stream v, as defined in Step 1, in pCi/mL. The values of Rt and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above. 1 l Steo 5: Determine whether the release is permissible, as follows: If c oy a p x;y, the release is permissible. However, if e nv is within about i 10 percent of p X;y, it may be impractical to use this value of c oy. I , This situation indicates that measured concentrations are ) approaching values which would cause the limits of Section 3.1.2.a m to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to 3-24 Gen. Rev. 13

     .-.. .. . . . . . . . . . - - . ~ . ..           . - ---_          ~ .. .. _ . .. - - -.                   ~ -      .-. ~ . _          _...~... - . . -

4 3 d 4 i s FNP-O-M-Oli f 1 adjust the allocation of the limits among the active release { points. The setpoint calculations (steps 1-4) must then be j repeated with parameters that reflect the modified conditions. l l I s If c,y < E aXjy, the release may not be made as planned. Consider the I l alternatives discussed in the paragraph above, and calculate a new , I setpoint based on the results of the actions taken. ) ) 3.3.2.3 Use of the Calculated Setpoint j f-l The setpoint calculated above is in the units yC1/mL. The monitor actually 1- measures a count rate that includes background, so that the calculated setpoint l must be converted accordingly: 1. c y

                                                                                =

(e gy

• Ey ) + B y (3.5) where:

eg = de monhor setpoint as a count rate. Ey = the monitor calibration factor, in count rate /(pci/mL). Monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to effluent stream concentrations

                                                  -measured by sample analysis.

B y =- the monitor ' background count rate. In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. Contributions to the monitor background may include any or all of the following. f, actors: ' ambi.ent, background radiation, plant-related radiation levels at- the monitor location (which may change between shutdown and power conditions), e.nd internal background due-to contamination of the monitor's sample chamber. The count rate units for c,y. , E y , and'Byin equation (3.5) must be the same (cpm or cps). s 3.3.3, seteoints for Noble Gas Monitors on Ef fluent Source Stre--- 3-25 Gen. Rev. 13

           --    -  .~. - _ .   . . _ _ _ . -      . ~ .-=- .. - -.- - - - --...                        -. ..

4 FNP-0-M-011 i Table 3-4 lists certain gaseous release pathways as being source streams. As may lA be seen in the figures of Section 3.2, these are streams that merge with other streams, prior to passing a final radioactivity monitor and being released. Unlike the final monitors, the source stream monitors measure radioactivity in i effluent streams for which flow can be terminated; therefore, the source stream 4 monitors have control logic to terminate the source stream release at the alarm setpoint. l 3.3.3.1 Setpoint of the Monitor on the Source Stream i ) Stoo 1: Determine the concentration X;,of each noble gas radionuclide i in l- source stream s (in pci/mL) according to the results of its required sample analyses (see Section 3.1.2). steo 2: Determine rg, the ratio of the dose rate limit for the total body, 500 mrem /y, to the total body dose rate due to noble gases detected q in the source stream under consideration. Use the Xis values and the maximum anticipated source stream flow rate f , in equation (3.2) to determine the total body dose rate for the source stream, substituting rt for R* t V The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the source stream is the same as the (X76)vb that applies to the respective merged stream. This is because the (176) value is determined by the meteorology of the plant site and the physical attributes of the release point, and is unaffected by whether or not a given source stream is operating. 1 steo 3: Determine rk, the ratio of the dose rate limit for the skin, 3000 mrom/y, to the skin dose rate due to noble gases detected in the source stream under consideration. Use the X;, values and the sawimum anticipated source stream flow rate f , in equation (3.3) to determine the skin dose rate for the source stream, substituting rk for R k* steo 4: Determine the maximust noble gas radioactivity monitor setpoint con - centration, as follows: 3-26 Gen. Rev. 13

u .-u&.a.4 ,..a:_ -. - 4 a w FNP-O-M-011 AG,* SF * [ Xg,* vg i cm = the lesser of (I

• O

, AG,

• SF * [ X;,
• vg i

where: c as = the calculated setpoint (in pCi/mL) for the noble gas monitor serving gaseous source stream s. AG, = the administrative allocation factor applied to gaseous source stream s. For a given final release point v, the sum of all the , AG, values for source streams contributing to the final release point must not exceed the release point *s allocation factor AG y. , 4 X;, = the measured concentration of noble gas radionuclide i in gaseous

source stream s, as definsd in Step 1, in pCi/mL.

The values of r and rk to be used in the calculatioe. are those which were g U determined in Steps 2 and 3 above. The safety factor, SF, was defined previously. steo 5: Determine whether the release is permissible, as follows: If c as 2 E X;,, the release s permissible. However, if e ns is within about i 10 percent of E X;,, it may be impractical to use this value of ca ,. I This situation indicates that measured concentrations are approaching values which would cause the limita of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce , contributing source terms of gaseous radioactive material, or to adjust the allocati.on of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions. If ens

• E Xis, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new l setpoint based on the results of the actions taken, i 3-27 Gen; Rev. 13

FNP-0-M-011 (" 3.3.3.2 Effect on the Setpoint of the Monitor on the Merged \ Stream Before beginning a release from a monitored source stream, a setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the previously-determined maximum allowable setpoint for the downstream final monitor on the merged stream must be redetermined. This is accomplished by repeating the steps of Section 3.3.2, with the following modifications. Modification 1: The new maximum anticipated flowrate of the merged stream is the sum of the old merged stream maximum flowrate ((f,y)old), and the maximum flowrate of the source stream  ; being considered for release (fu)* I (fav)new * (fav)old + fas (3 7) l l Modification 2: The new concentration of noble gas radionuclide i in the 1 l merged stream includes both the contribution of the j merged stream without the source stream, and the source stream being considered for release. (fav)old ' (Xiv)old

• f as *Xis (Xgy)new =

(3.8) (fav)new , l I 3.3.4 Determination of Allocation Factors, AG l When simultaneous gaseous releases are conducted, an administrative allocation ' factor must be applied to divide the release limit among the active gaseous release pathways. This is to assure that the dose rate limit for areas at and beyond the SITE BOUNDARY (see Section 3.1.2) will not be exceeded by simultaneous releases. The allocation factor for any pathway may be assigned any value between 0 and 1, under the following two conditions:

1. The sum of the allocation factors for all simultaneously-active final release paths at the plant site may not exceed 1.

2. The sum of the allocation factors for all simultaneously-active source screams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.

3-28 Gen. Rev. 13

  ~.          ._.                _       _    ._        _        __. .   .       _     _ _    ..._m _

PNP-0-M-011 Any of the following three methods may be used to assign the allocation f actors Q to the active gaseous release pathways:

1. For ease of implementiation, AG y may be equal for all release pathways:

1 AG y = p (3,93 where: 1

N= the number of simultaneously active gaseous release pathways,

.1

2. AGy for a given release pathway may be selected based on an earlmate of

[ the portion of the total SITE BOUNDARY dose rate (from all simultaneous { releases) that is contributed by the release pathway. During periods when l a given building or release pathway is not subject to gaseous radioactive

releases, it may be assigned an allocation factor of zero.

3. AG y for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows:

O (XIQ)vb E(Kj Qjy) . Aa y = I N (3.10) E (270)rb t E(KiOr) r=1 i where (X70)vb = the annual average SITE BousiOARY relative concentration . applicable to the gaseous release pathway v for which the  ! allocation factor is being determined, in s/m 3. l Kg = the total-body dose factor due to gamma emissions from noble gas-radionuclide .1, in (mrom/y)/( C1/m 3 ), from Table-3-5. l l I Qy= the release rate of noble gas radionuclide i from release pathway v, in pci/s, calculated as the product of Xjy and f,y, where: l I 3-29 Gen. Rev. 13

FNP-O-M-011 Xjy = the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the ellocation factor is being determined, in pci/mL. f,y . = the discharge flowrate applicable to gaseous release pathway v for which the allocation factor is being determined, in mL/s. (IGQ)g = the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r, in s/m3 . gir - the release rate of noble gas radionuclide i applicable to active release pathway r, in pci/s, calculated as the product of X;,and f ,, where , Xir = the concentration of noble gas radionuclide i applicable to active gaseous release pathway r, in pci/mL. f,= the discharge flowrate applicable to active gaseous release pathway r, in mL/s. N= the number of simultaneously active gaseous release pathways (including pathway v that is of interest). NOTE: Although equations (3.9) and (3.10) are written to illustrate the assignment of the allocation factors for final release pathways, they may also be used to assign allocation factors to the source streams that merge into a given final release pathway. 1 LO 3-30 Gen. Rev. 13

FNP-0-M-011 f) k ,) 3.3.5 Setooints for Noble Gas Monitors with Special Reauirements The Parley Nuclear Plant operating philosophy treats the Waste Gas Decay Tank supply monitors (1/2 RE-013) and the Containment Purge monitors (1/2 RE-024) as process monitors, not effluent monitors. However, as a matter of information, the following may be noted regarding their setpoints: 0 For 1/2 RE-013, _the alarm setpoint should be based on a concentration equivalent to no more than the Technical Specification limit for the maximum curie content of a Waste Gas Decay Tank. In converting the curie limit to an equivalent concentration at the location of RE-013, the maximum allowable Waste Gas Decay Tank pressure should be used. l 0 For 1/2 RE-024, the alarm setpoint concentration may be arrived at in 1 either of two ways. In the first method, the maximum setpoint concentration established by the Technical Specifications may be used. Alternatively, to provide early detection and termination of an abnormally high containment purge release, the (lower) setpoint concentration calculated according to Section 3.3.3 may be used, r~'s 3.3.6 Setooints for Particulate and Iodine Monitors () In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the ODCM establish setpoint l calculation methods for particulate and iodine monitors, rh Iv] 3-31 Gen. Rev. 13

FNP-0-M-011 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3.4.1 Dose Rates at and Beyond the Site Beyndary l Because tr.e dose rate limits for areas at and beyond the SITE specified in ] Section 3.1.2 are site limits applicable at any instant in time, the summations l extend over all simultaneously active gaseous final release pathways at the plant  ! site. Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (IGQ)vb value for each. 3.4.1.1 Dose Rat.es Due to Noblo Gases For the purpose of implementing the controle of $setion 3.1.2.a, the dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows: For total body dose rates: DR t " b y L ( M )vb b [K D i i i v)j (3.11) For skin dose rates: DRk l i (3.12) bv (X/0)vb [ bf ((Li+1*1M)Dv)J

• where:

DRt= the total body dose rate at the time of the release, in mrom/y. DRk= the skin dose rate at the time of the release, in mrem /y. Q;y = the release rate of noble gas radionuclide i, in yCi/s, equal to , the product of fg and Xjy, where: fw= the actual average flowrate for release pathway v during the period of the release, in mL/s. All other terms were defined previously. O v 3-32 Gen. Rev. 13

                     - -    ..-       . ~ - - .          . . .      ;--           _ .----~                _ .. ,

i FNP-0-M-011 3.4.1.2 Dose Rates Due to Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form with Half-Lives Greater than 8 Days For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate fore , with half-lives greater than 6 days, in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows: DR o = {v (YTC)vb b Plo O y (3*13) i < l where: t e e DR o = the dose rate to organ o at the time of the release, in mrem /y. j Pio = the site-specific dose factor for radionuclide i and organ o, in  ! f (mrem /y)/(pci/m 3). Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation ' pathway, the values of P;o may be obtained from Table 3-9, "R,;g for Inhalation Pathway, Child Age Group." I Qfy= the release rate of radionuclide i from gaseous release pathway v, I in pCi/s. For the purpose of implementing the controls of Section 3.1.2.b, only I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation. All other terms were defined previously. 3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas at or beyond the SITE BOUNDARY due to releases of noble gases from each unit shall be calculated as follows (adapted from Reference 1, page 28, by including only long-term releases): l 1 Dg = 3.17 x 10-I { (Y75)4 { Ng

• d pi, (3*14) v , i v

3-33 Gen. Rev. 13

, ~ . - - .

                                                                        - -~                    --

FNP-O-M-011 t v Dy - 3.17 x 10~8 ))v (y73) p3 }] g j.g' jy' (3.15)

                                                      ,          I where:

3.17 x 10-8 = a units conversion factor: 1 y/(3.15 x 107 m). Dg = the air dose due to beta emissions from noble gas radionuclides, in mrad. Dy = the air dose due to gamma emissions from noble gas radionuclides, in mrad. ' N; = the air dose factor due to beta emissions from noble gas radio-nuclide i, in (mrad /y)/(pci/m 3), from Table 3-5. Mj = the air dose factor due to gamma emissions from noble gas radio-nuclide i, in (mrad /y)/(pci/m 3), from Table 3-5. 0

  ,U 6;y  =   the cumulative release of noble gas radionuclide i from release pathway v, in pci, during the period of interest.                          !

All other terms were defined previously. Because the air dose limit is on a per-reactor-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned to the two units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit. The gaseous final release pathways at the plant site, and the (1/6)vb for each, are identified in Table 3-4. v 3-34 Gen. Rev. 13

FNP-0-M-011 [dh Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Cases 4 y - Body (K) $ - Skin (L) y - Air (M) $ - Air (N) Nuclide

(mrem /y) per (mrom/y) per (mrad /y) per (mrad /y) per (pCi/m 3) (yC1/m3) (pCi/m 3) (yci/m 3) i Kr-83m 7.56 E-02 0.00 E+00 1.93 E+01 2.88 E+02 )

Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 l Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 4 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1,63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 O Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 , Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 All values in this table were obtained from Reference 3 (Table B-1), with units converted. O 3-35 Gen. Rev. 13

1 FNP-0-H-011 Table 3-6. Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated 71 nite Plume O j I r The contents of this table are not applicable to the Farley Nuclear Plant. O . p b 1 O 3-36 Gen. Rev. 13

FNP-0-M-011 3.4.3 Dose to a Member of the Public at or Beyond Site Boundary l The dose received by an individual due to gaseous releases from each reactor unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The MEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the controlling receptor. The dosimetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7. Doses to a member of the public due to gaseous releases of I-131, I-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases): Dja

                             =  3.17 x 10-8   { [Ralpj E         Wip

• d v' '

(3*16) p i v where: O

 \j        Dj, =     the dose to organ j of an individual in age group a, due to gaseous releases of    I-131,  I-133,   tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in mrom.

3.17 x 10-8 = a units conversion factor: 1 y/(3.15 x 107 s). Raipj = the site-specific dose factor for age group a, radionuclide 1, exposurs pathway p, and organ j. For the purpose of implementing the controls of Section 3.1.4, the exposure pathways applicable to calculating the dose to the currently-defined controlling receptor are included in Table 3-7; values of Raipj for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are listed in Table 3-8 through Table 3-11. A detailed discussion of the methods and parameters used for calculating R,jp j for the plant site is presented in Chapter 9. That information may be used for recalculating the Raipj values if thm. *-Y == parmanarara change, or for calculating R,j j values p A for special radionuclides and age groups when performing the h assessments discussed in section 3.4.4 below. 3-37 Gen. Rev. 13

FNP-0-M-011 W y ;p = the annual average relative dispersion or deposition at the location of.the controlling receptor, for release pathway v, as I appropriate to exposure pathway p and radionuclide 1. For all tritium pathways, and for the inhalation of any radio-nuclide: Wy ;p _ is (17D)yp, the annual average relative dispersion  ; factor for release pathway v, at the location of the controlling i receptor (s/m 3). For the ground-plane exposure pathway, and for i all ingestion-related pathways for radionuclides other than { tritium: W yip is (676)yp, the annual average relative deposition l factor for release pathway v, at the location of the controlling i receptor (m-2). Values of (176)yp and (676)yp for use in calculating the dose to the currently-defined controlling receptor are included in Table 3-7.  ! 6 fy= the cumulative release of radionuclide i from release pathway v, during the period of interest (pci). For the purpose of implementing the controls of section 3.1.4, only I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation. In any . dose assessment using the methods of this sub-section, only radio- *j nuclides detactable above - background in their respective samples l should be included in the calculation. Because the member of the public dose limit is on a per-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials. originating in both reactor units, the activity discharged from the release point may be apportioned between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one unit or the other. I The gaseous final release pathways at the plant site, and the release elevation for.each ,. arm.identif M in Table 3-4. 1 3-38 Gen. Rev. 13

f FNP-0-M-011 Table 3-7. Attributes of the controlling Receptor i The locations of members of the public in the vicinity of the plant site, and the cxposure pathways associated with those locations, are determined in the Annual Land Use census. Dispersion and deposition values were calculated based on site meteorological data collected for the years 1971 through 1975. Based on the Land Use Census of June 7, 1991, the current controlling receptor for the plant site is described as follows. Sectors sw Distances 1.2 miles Ace Groues Child Exoosure Pathways Ground Plane Inhalation Garden Vegetation Grass / Cow / Meat Dispersion Factors fX76)yp: Ground-Level discharge points: 8.74 x .10 4 s/m 3 l l 3 Mixed-Mode discharge points: 8.03 x 10*7 s/m Deposition Factors (676) yp: Ground-Level discharge points: 2.64 x 10'8 nc2 Mixed-Mode discharge points: 1.05 x 10-8 m -2 l This location represents the residetce with the highest annual average X/Q and D/Q factors in the vicinity of the FNP. The referenced Land Use Census identified no locations where animala are maintained for milk within 5 miles of the plant site; thus, it is very unlikely that any real dairy location (which would be beyond 5 miles) would have a hiiher potential dose impact than the real residence location selected. A 3-39 Gen. Rev. 13

4 i FNP-0-M-011 3.4.4 Dose calculations to succort other Reauirements case 1: Under Technical Specification 6.6.1, a radiological impact assess-ment may be required to support evaluation of a reportable event. Dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters [(X/Q) and (D/Q)] for the period covered by the report, and using the appropriate l pathway dose factors (Raipj) for the receptor of interest. Methods for eniculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of R,jp j presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Rjj ap values f applicable to that receptor must first be calculated. Methods and parameters for calculating R,;pj for radionuclides and age groups other than those required in section 3.4.3 are presented in chapter 9. When calculating R,jpj for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the O- values in Chapter 9, if the specific values are known. ai case 2:- A dose calculation is required to evaluate the results of the Land Use census, under the provisions of section 4.1.2. In the event that the Land Use census reveals that exposure pathways have changed at previously-identified locations, or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be designated as the controlling j receptor. Such dose calculations may be performed using the equations in i Section 3.4.3, with the substitution of the annual average dispersion and i deposition values [(X76) and (D/D)) for the locations of interest, and + j using the appropriate pathway dose factors (Raipj) for the receptors of , interest. 1 i

Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of R,jp ; presented in Table 3-8 through Table 3-11 are applicable only . to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Raipj values applicable to that receptor must first be calculated.

O Methods and parameters for calculating . R,;pj for radionuclides and age i 3-40 Gen. Rev. 13

y _ _. _ _ _ _ _ . _ _ , _ _ _ _ _ . . . 3 FNP-O-M-011 groups other than those required in Section 3.4.3 are presented in Chapter 9. case 3: Under section 5.2, a dose calculation is required to support , l determination of total dose to a receptor of age group other than l that currently defined as the controlling receptor. Dose calculations shall be performed using the equations in Section 3.4.3, using the dispersion and deposition parameters defined in Table 3-7 for the controlling receptor, but substituting the appropriate pathway dose factors (R,jj) p for the receptor age group of interest. The values of R,jp ; presented in Table 3-8 through Table 3-11 are applicable 1 only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor age group, R,j p; values applicable to that receptor must first be calculated. Methods and parameters for calculating R ap j ; for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9. 1 1 C\ i a 1 4 4 a O V L-3-41 Gen. Rev. 13

FNP-0-M-011 Table 3-8. R,jpj for Ground Plane Pathway, All Age Groups Nuclide T. Body Skin H-3 0.00 0.00 i Cr-51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63E+09  ! Fe-55 0.00 0.00 Fe-59 2.73E+08 3.21E+08 I Co-58 3.79E+08 4.44E+08 I Co-60 2.15E+10 2.53E+10 ) Ni-63 0.00 0.00 ) Zn-65 7.47E+08 8.59E+08 l Rb-86 8.99E+06 1.03E+07 i Sr-89 2.16E+04 2.51E+04 Sr-90 0.00 0.00 l Y-91 1.07E+06 1.21E+06 ) Zr-95 2.45E+08 2.84E+08 Nb-95 1.37E+08 1.61E+08 ["' (_ ,/ Ru-103 Ru-106 1.08E+08 4.22E+08 1.26E+08 5.07E+08 Ag-110m 3.44E+09 4.01E+09 Sb-124 5.98E+08 6.90E+08 Sb-125 2.34E+09 2.64E+09 Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.08E+05 Te-129m 1.98E+07 2.31E+07 I-131 1.72E+07 2.09E+07 I-133 2.45E+06 2.98E+06 Cs-134 6.86E+09 8.00E+09 C -136 1.51E+08 1.71E+08 Cs-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Co-141 1.37E+07 1.54E+07 Ce-144 6.95E+07 8.04E+07 Pr-143 0.00 0.00

                                                                                      ]

Nd-147 8.39E+06 1.01E+07

1. Units are m 2-(mrom/yr)/(pci/s).

2. The values in the Total Body column also apply to the Bone, f-~ Liver, Thyroid, Kidney, Lung, and GI-LLI organs.

(j 3. This table also supports the calculations of Section 6.2. 3-42 Gen. Rev. 13

l I FNP-O-M-011 . Table 3-9. R,jp j for Inhalation Pathway, Child Age Group O V Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 i cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 1 Hn-54 0.00 4.29E+04 9.51E+03 0.00 1.00E+04 1.58E+06 2.29E+04 I Fe-55 4.74E+04 2.52E+04 7.77E+03 0.00 0.00 1.11E+05 2.87E+03 Fe-59 2.07E+04 3.34E+04 1.67E+04 0.00 0.00 1.27E+06 7.07E+04 I l Co-58 0.00 1.77E+03 3.16E+03 0.00 0.00 1.11E+06 3.44E+04 ) Co-60 0.00 1.31E+04 2.26E+04 0.00 0.00 7.07E+06 9.62E+04 ' Ni-63 8.21E+05 4.63E+04 2.80E+04 0.00 0.00 2.75E+05 6.33E+03 j Zn-65 4.26E+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05 1.63E+04  ! Rb-86 0.00 1.98E+05 1.14E+05 0.00 0.00 0.00 7.99E+03 Sr-89 5.99E+05 0.00 1.72E+04 0.00 0.00 2.16E+06 1.67E+05 Sr-90 1.01E+08 0.00 6.44E+06 0.00 0.00 1.48E+07 3.43E+05  ! Y-91 9.14E+05 0.00 2.44E+04 0.00 0.00 2.63E+06 1.84E+05 Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.96E+04 2.23E+06 6.11E+04 Nb-95 2.35E+04 9.18E+03 6.55E+03 0.00 8.62E+03 6.14E+05 3.70E+04 l Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 6.62E+05 4.48E+04 Ru-106 1.36E+05 0.00 1.69E+04 0.00 1.84E+05 1.43E+07 4.29E+05 l [ ) Ag-110m 1.69E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+06 1.00E+05 I

  '#     Sb-124        0.00          0.00      0.00      0.00      0.00      0.00       0.00   ,

Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 6.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 I-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00 2.84E+03 I-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.00 5.48E+03 Cs-134 6.51E+05 1.01E+06 2.25E+05 0.00 3.30E+05 1.21E+05 3.85E+03 Cs-136 6.51E+04 1.71E+05 1.16E+05 0.00 9.55E+04 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.20E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11E+01 1.74E+06 1.02E+05 Co-143 3.92E+04 1.95E+04 2.902+03 0.00 8.55E+03 5.44E+05 5.66Ee04 Co-144 6.77E+06 2.12E+06 3.61E+05 0.00 1.17E+06 1.20E+07 3.89E+05 Pr-143 1.85E+04 5.55E+03 9.14E+02 0.00 3.00E+03 4.33E+05 9.73E+04 0.00 4.81E+03 3.28E+05 8.21E+04 _Nd-147 1.08E+04 8.73E+03 6.81E+02 q Units are (mrem /yr)/(pci/m3 ) for all radionuclides.

 \J 3-43                           Gen. Rev. 13 1

l FNP-0-M-011 9

 /      Table 3-10. R aid for Cow Meat Pathway, Child Age Group
 ,        __Nuclide     Bone       Liver   T. Body   Thyroid   Kidney    Lung      GI-LLI H-3          0.00 2.34E+02   2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 Cr-51          0.00       0.00 8.79E+03 4.88E+03 1.33E+03 8.91E+03 4.66E+05 Mn-54          0.00 8.01E+06 2.13E+06         0.00 2.25E+06     0.00 6.72E+06 Fe-55     4.57E+08 2.42E+08 7.51E+07          0.00     0.00 1.37E+08 4.49E+07 Fe-59     3.76E+08 6.09E+08 3.03E+08          0.00     0.00 1.77E+08 6.34E+08 Co-58          0.00 1.64E+07 5.02E+07         0.00     0.00'    O.00 9.58E+07 Co-60          0.00 6.93E+07 2.04E+08         0.00     0.00     0.00 3.84E+08 Ni-63     2.91E+10 1.56E+09 9.91E+08          0.00     0.00     0.00 1.05E+08 Zn-65    3.75E+08 1.00E+09 6.22E+08          0.00 6.30E+08     0.00 1.76E+08 Rb-86          0.00 5.77E+08 3.55E+08         0.00     0.00     0.00 3.71E+07 Sr-89     4.82E+08        0.00 1.38E+07       0.00     0.00     0.00 1.87E+07 Sr-90     1.04E+10        0.00 2.64E+09       0.00     0.00     0.00 1.40E+08 Y-91     1.80E+06        0.00 4.82E+04       0.00     0.00     0.00 2.40E+08 Zr-95    2.66E+06 5.85E+05 5.21E+05          0.00 8.38E+05     0.00 6.11E+08 Nb-95     3.10E+06 1.21E+06 8.62E+05          0.00 1.13E+06     0.00 2.23E+09 Ru-103    1.55E+08        0.00 5.96E+07       0.00 3.90E+08     0.00 4.01E+09 Ru-106    4.44E+09        0.00 5.54E+08       0.00 5.99E+09     0.00 6.90E+10 Ag-110m    8.39E+06 5.67E+06 4.53E+06          0.00 1.06E+07     0.00 6.74E+08 Sb-124         0.00       0.00      0.00      0.00      0.00    0.00       0.00   ,l Sb-125         0.00       0.00      0.00      0.00      0.00    0.00       0.00      i Te-125m    5.69E+08 1.54E+08 7.59E+07 1.60E+08           0.00    0.00 5.49E+08 Te-127m    1.77E+09 4.78E+08 2.11E+08 4.24E+08 5.06E+09          0.00 1.44E+09 Te-129m    1.79E+09 5.00E+08 2.78E+08 5.77E+08 5.26E+09          0.00 2.18E+09 I-131    1.65E+07 1.66E+07 9.46E+06 5.50E+09 2.73E+07          0.00 1.48E+06 I-133    5.67E-01 7.02E-01 2.66E-01 1.30E+02 1.17E+00          0.00 2.83E-01 Cs-134    9.22E+08 1.51E+09 3.19E+08          0.00 4.69E+08 1.68E+08 8.16E+06 Cs-136    1.62E+07 4.46E+07 2.88E+07          0.00 2.37E+07 3.54E+06 1.57E+06 Cs-137    1.33E+09 1.28E+09 1.88E+08          0.00 4.16E+08 1.50E+08 7.99E+06 Ba-140    4.38E+07 3.84E404 2.56E+06          0.00 1.25E+04 2.29E+04 2.22E+07 Ce-141    2.22E+04 1.11E+04 1.64E+03          0.00 4.86E+03      0.00 1.38E+07 Ce-144    2.32E+06 7.26E+05 1.24E+05          0.00 4.02E+05      0.00 1.89E+08 Pr-143    3.34E+04 1.00E+04 1.66E+03          0.00 5.43E+03      0.00 3.60E+07 Nd-147    1.17E+04 9.47E+03 7.33E+02          0.00 5.19E+03      0.00 1.50E+07 Units are (mrom/yr)/(pC1/m 3) for tritium, and m 2-(mram/yr)/(pci/s) for all other radionuclides.

f~ 3-44 Gen. Rev. 13

 , e FNP-0-M-011

[ (m / Table 3-11. R,jpj for Garden Vegetation Pathway, Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 Cr-51 0.00 0.00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+06 Mn-54 0.00 6.65E+08 1.77E+08 0.00 1.86E+08 0.00 5.58E+08 Fe-55 8.01E+08 4.25E+08 1.32E+08 0.00 0.00 2.40E+08 7.87E+07 Fe-59 3.98E+08 6.43E+08 3.20E+08 0.00 0.00 1.86E+08 6.70E+08 Co-58 0.00 6.44E+07 1.97E+08 0.00 0.00 0.00 3.76E+08 Co-60 0.00 3.78E+08 1.12E+09 0.00 0.00 0.00 2.10E+09 N1-63 3.95E+10 2.11E+09 1.34E+09 0.00 0.00 0.00 1.42E+08 Zn-65 8.13E+08 2.16E+09 1.35E+09 0.00 1.36E+09 0.00 3.80E+08 Rb-86 0.00 4.52E+08 2.78E+08 0.00 0.00 0.00 2.91E+07 Sr-89 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-90 1.24E+12 0.00 3.15E+11 0.00 0.00 0.00 1.67E+10 l Y-91 1.86E+07 0.00 4.99E+05 0.00 0.00 0.00 2.48E+09 ) Zr-95 3.86E+06 8.48E+05 7.55E+05 0.00 1.21E+06 0.00 8.85E+08 , Nb-95 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.96E+08 l

   /'"}
     ~/

Ru-103 1.53E+07 0.00 5.90F+06 0.00 3.86E+07 0.00 3.97E+08 Ru-106 7.45E+08 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ag-110m 3.21E+07 2.17E+07 1.73E+07 0.00 4.04E+07 0.00 2.58E+09 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 To-125m 3.51E+08 9.50E+07 4.67E+07 9.84E+07 0.00 0.00 3.38E+08 Te-127m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00 1.07E+09 Ta-129m 8.41E+08 2.35E+08 1.31E+08 2.71E+08 2.47E+09 0.00 1.03E+09 I-131 1.43E+08 1.44E+08 8.17E+07 4.75E+10 2.36E+08 0.00 1.28E+07 I-133 3.53E+06 4.37E+06 1.65E+06 8.11E+08 7.28E+06 0.00 1.76E+06 Cs-134 1.60E+10 2.63E+10 5.55E+09 0.00 8.15E+09 2.93E+09 1.42E+08 l Cs-136 8.24E+07 2.27E+08 1.47E+08 0.00 1.21E+08 1.80E+07 7.96E+06 Co-137 2.39E+10 2.29E+10 3.38E+09 0.00 7.46E+09 2.68E+09 1.43E+08 Ba-140 2.77E+08 2.42E+05 1.61E+07 0.00 7.89E+04 1.45E+05 1.40E+08 d Ce-141 6.56E+05 3.27E+05 4.86E+04 0.00 1.43E+05 0.00 4.08E+08 , Co-144 1.27E+08 3.98E+07 6.78E+06 0.00 2.21E+07 0.00 1.04E+10 l Pr-143 1.46E+05 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.57E+08 Nd-147 7.15E+04 5.79E+04 4.48E+03 0.00 3.18E+04 0.00 9.17E+07 Units are (mrom/yr)/(pci/m 3) for tritium, and m 2.(mrem /yr)/(pci/s) for all other radionuclides. s_e 3-45 Gen. Rev. 13

N, q y ? 1 FNP-0-M-011 l Q . 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3.5.1 Thirty-one Day Dose Proiections In order to meet the requirements of the limit for operation of the gaseous radwaste treatment system (see Section 3.1.5), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to areak at or beyond the SITE BOUNDARY of gaseous ef fluents containing radioactive materials occurs or is expected. Projected 31-day air doses and dones to individuals due to gaseous affluents may be determined as follows: For air doses:

                                                          =  'Dfc' x 31 Dgp                 + DBa
                                                            '7(
                                                            'D  #

(3.17) D yp = x 31 + D, (o,) For individual doses: D,p = x 31 + D,, (3 18) wh.re, Dgp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases. Dge = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration. Dg, = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous ef fluents. If only routine gaseous affluents are anticipated, D$a may be set to zero. Q D, = the projected air dose due to gamma emissions from noble gases for the next 31 days of gaseous releases. 3-46 Gen. Rev. 13

l { i 1 FNP-O-M-011 O V D ye = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration. D,= 7 the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous af fluents. If only routine 1 gaseous ef fluents are anticipated, Dy , may be set to zero.  ! D op = the projected dose to the total body or organ o, due to releases of I-131, I-133, tritium, and particulates for the next 31 days of gaseous releases. Doc = the cumulative dose to the total body or organ o, due to releases of I-131, I-133, tritium, and particulates that have occurred in the elapsed portion of the current quarter, plus the release under consideration. l Da= the anticipated dose to the total body or organ o, due to releases of I-131, I-133, tritium, and particulates, contributed by any ' planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous affluents are anticipated, D ,omay be meid. to zero. t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter). 3.5.2 Dose Proinctions for Soecific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For air dose and individual dose projections due to gaseous effluent releases, follow the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to amist.during the release period. O V I i 3-47 Gen. Rev. 13

V 1 FNP-O-M-011 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS l Section of O' IRIg1 Definition Initial Use

      -AG =            the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit among all the release pathways (unitiess).                    3.3.2.1 AG, =           the administrative allocation factor for gaseous source stream    s,  applied to divide the gaseous release   limit   among  all   the   release   pathways (unitiess).                                                     3.3.3 AG y =          the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous release   limit   among  all  the   release    pathways (unitiess).                                                   3.3.2.2 j

l c= the setpoint of the radioactivity monitor measuring l the concentration of radioactivity in the effluent I line prior to release (pci/mL). 3.3.2.1 I O c as = the calculated noble gas effluent monitor setpoint for gaseous source stream o (pci/mL). 3.3.3 e ny = the calculated nobis gas effluent monitor setpoint i for release pathway v [pci/mL). 3.3.2.2 Dj, = the dose to organ j of an individual in age group a, due to gaseous releases of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days (mrom). 3.4.3 D,= o the anticipated done to organ o due to releases of non-noble-gas radionuclides, contributed by any planned activities during the next 31-day period (mrom). 3.5.1 Doc = the cumulative dose to organ o due to releases of non-noble-gas radionuclides that have occurred in the elapsed portion of the current quarter, plus the

 /                     release under consideration (mrom).                             3.5.1 3-48                           Gen. Rev. 13

[ i L FNP-O-M-011 Section of

         ~ 19I3                            Definition                      Initial Use D op =   the projected dose to organ o due to the next 31 days of gaseous releases 'of non-noble-gas radionuclides

[mrom). 3.5.1 Dg = the air dose-due to beta emissions from noble gas radionuclides [ mrad). 3.4:2 Dg, = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period [ mrad). 3.5.1 Dge = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [ mrad). 3.5.1 Dgp = the projected air dose due to beta emissions. f rom noble gases, for the next 31 days of gaseous releases [ mrad). 3.5.1 l Dy = the air dose due to gansna emissions from noble gas radionuclides [ mrad). 3.4.2 D,= y the anticipated air dose due to gamms emissions from noble gas teleases, contributed. by any planned activities during the next 31-day period (mrad). 3.5.1 l D ye = the cumelative air dose due to gamma emissions from l noble gas releases that have occurred.in the elapsed portion of the current quarter, plus the release

                   .under consideration [ mrad).                                 3.5.1 i D     =  the projected air dose due to gamma emissions from 9

noble gases, for the next 31 days of gaseous releases [ mrad). 3.5.1

   . s_/

3-49 Gen Rev. 13 ' m .

7 m FNP-0-M-011 M Definition Section of Initial Use (676)yp = the annual average relative deposition factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [m~2), 3.4.3 DRk= the skin dose rate at the time of the release [mrom/y). 3.4.1.1 DRg ' = the dose rate to organ o at the time of the release [mrom/y). 3.4.1.2 DRt= the total body dose rate at the time of the release [mrom/y). 3.4.1.1 f,y = the maximum anticipated actual discharge flowrate for release pathway v during the period of the planned release (mL/s). 3.3.2.2 O f ,= the maximum anticipated actual discharge flowrate for

• gaseous source stream a during the period of the planned release [mL/s). 3.3.3 K = the total body dose f actor due to gamma emissions from noble gas radionuclide 1, from Table 3-5

[(mrom/y)/(yC1/m3 )). 3.3.2.2 L = . the skin dose factor due to beta emissions from noble gas radionuclide 1, from Table 3-5 [(mrom/y)/(pci/m 3)). 3.3.2.2 Mj = the air dose f actor due to gamma emissions from noble gas radionuclide 1, from Table 3-5 [(mrad /y)/(pC1/m )}. 3 3.4.2 N= the number of simultaneously active gaseous release pathwaye [unitlees). 3.3.4 3-50 Gen. Rev. 13

FNP-0-M-011 Section of

     /N   ISIm                             Definition                          Initial Use Nj _ =  the air dose factor due to beta emissions from noble gas     radionuclide         i,     from      Table     3-5

[(mrad /y)/(pci/m 3)). 3.4.2 Pjo = the site-specific dose factor for radionuclide 1 (I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) and organ o. The values of Pjo are equal to the site-specific R,jpj values presented in Table 3-9 3 [(mram/y)/(pci/m )}. 3.4.1.2 4 Q;y = the release rate of noble gas radionuclide i from i release pathway v during the period of interest j ! [pci/s). 3.3.2.2 i Q fy= the release rate of radionuclide 1 (I-131, I-133, tritium, and radionuclides in particulate form with [ half-lives greater than 8 days) from gaseous release i pathway v during the period of interest (pci/s). 3.4.1.2 , 6;y = the cumulative relaase of noble gas radionuclide i

j. from release pathway v during the period of interest

[pci). 3.4.2 i 6 fy= the cumulative release of non-noble-gas radionuclide , i from release pathway v, during the period of interest [pci). 3.4.3 j R,;p; = the site-specific dose factor for age group a, radio-nuclide i, exposure pathway p, and organ j. Values l and units of Rjj ap for each exposure pathway, age group, and radionuclide that may arise in i calculations for implementing Section 3.1.4 are j listed in Table 3-8 through Table 3-11. 3.4.3 i Rk= the ratio of the skin dose rate limit for noble i gases, to the skin dose rate due to noble gases in

     \  -

the release under consideration [unitiess). 3.3.2.1 3-51 Gen. Rev. 13 l i

  , ..           -       . .      . .                  ~-        -    - ~ . . .       ,.     -   ~.       . .

4 , , FNP-0-M-011 Section of T.1IBl Definition Initial Use i 4 Rt= the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the release under consideration (unitiess). 3.3.2.1 l rk = the ratio of tha skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the source stream under consideration [unitiess). 3.3.3.1 rt = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration 4 (unitiess). 3.3.3.1 4 SF = the safety factor used in gaseous setpoint calculations to compensate for statistical } fluctuations and errors of measurement (unitiess). 3.3.2.2 1 r% < t= the number of whole or partial days elapsed in the current quarter, including the period of the release *I under consideration. 3.5.1 Wy ;p = the annual average relative dispersion [(X76)yp) or deposition ((576)yp) at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radio-nuclide 1. 3.4.3 X= the noble gas concentration for the release under consideration (yci/mL). 3.3.2.1 Xir = the concentration of radionuclide i applicable to active gaseous release pathway r (pci/mL). 3.3.4 Xj, = the measured concentration of radionuclide i in gaseous source stream s [pci/mL). 3.3.3 Xjy = the measured concentration of radionuclide. 1 in a gaseous stream v (pci/mL). 3.3.2.2 3-52 Gen. Rev. 13

l i FNP-O-M-011 Section of Iggg Definition Inittal Use l'\ V (X/Q) = the highest relative concentration at any point at or  ! beyond the SITE BOUNDARY (s/m3 ). 3.3.2.1  ; (176)4 = the annual average SITE BOUNDARY relative concen-tration applicable to active gaseous release pathway ) r (s/m3 ). 3.3.4 (i?6)vb = the highest annual average relative concentration at l the SITE BOUNDARY for the discharge point of release pathway v, from Table 3-4 [s/m3 ). 3.3.2.2 1 1 (176)yp = annual average relative dispersion f actor for release j pathway v, at the location of the controlling receptor, from Table 3-7 [s/m 3). 3.4.3 j l I O I i l 0 O 3-53 Gen. Rev. 13

1 FNP-O-M-011 CHAPTER 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM G) 4.1 LIMITS OF OPERATION l The following limits are the same for both units at the site. Thus, a single program including monitoring, land use st.rvey, and quality assurance serves both units. 4.1.1 Radioloaical Environmental Monitorina I In accordance with Technical Specification 6.8.3.f(i), the Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table 4-1, 4.1.1.1 Applicability , l l This control applies at all times. l 4.1.1.2 Actions l 4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1, submit V to the Nuclear Regulatory Commission (NRC),

in the Annual Radiological a

! Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations l from the required sampling schedule are permitted if specimens are unobtainable l due to hazardous conditions, unavailability, inclement weather, equipment malfunction, or other just reasons. If deviations are due to equipment malfunction, efforts shall be made to complete corrective action prior to the end of the next sampling period. 4.1.1.2.2 With the confirmedI measured level of radioactivity as a result of plant effluents in an environmental sampling medium specified in Table 4-1 l exceeding the reporting levels of Table 4-2 when averaged over any calendar quarter, submit within 30 days a Special Report to the NRC pursuant to Technical Specification 6.9.2. The Special Report shall identify the cause(s) for exceeding the 1 Lait (s) and define the corrective action (s) to be taken to reduce radioactive ef fluents so that the potential annual dose to a MEMBER OF THE PUBLIC 1 Defined as confirmed by reanalysis of the original sample, or analysis of a duplicate or new sample, as appropriate. The results of the confirm-atory analysis shall be completed at the earliest time consistent with the 7 ( analysis. 4-1 Gen. Rev. 13

gr-- , , . z q FNP-O-M-011 is less than the calendar year limits of Sections 2.1.3, 3.1.3, and 3.1.4. The [) methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in the special Report. l When more than one of the radionuclides in Table 4-2 are detected in the sampling 4 medium, this report shall be submitted ifs 9 9 concentration (1) , concentration (2) +...a 1. 0

  ;                            reporting level (1)     reporting level (2) 3 s

When radionuclides other than those in Table 4-2 are detected and are the result of plant effluents, this special Report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits stated in sections 2.1.3, 3.1.3, and 3.1.4. This Special Report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be described in the Annual Radiological Environmental Operatir.g keport. The levels of naturally-occurring radionuclides which are not included in the plant's effluent releases need not be reported. 4.1.1.2.3 If adequate samples of milk, or during the growing season, forage () or fresh leafy vegetation, can no longer be obtained from one or more of the sample locations required by Table 4-1, or if the availability is frequently or psrsis;ently wanting, efforts shall be medes to identify specific locations for  ; obtaining suitable replacement samples; and to add any 2. placement locations to

• the REMP given in the ODCM within 30 days. The specific locations from which , <

samples became unavailable may be de,',eted from the REMP. Pursuant to Technical Specification 6.14, documentation shall be submitted in the next Annual Radioactive Effluent Release Report for the change (s) in the ODCM, including revised figure (s) and table (s) reflecting the changes to the location (s), with supporting information identifying the cause of the unavailability of samples and justifying the selection of any new location (s). 4.1.1.2.4 This control does not af fect shutdoun requirements or MODE changes, i l O V 4-2 Gen. Rev. 13 f I i l

    -. - .. _ _ _..._ _ _. _ _ _-_._. . _-                                   . - _ - _ ~ _ . _ _ _ _. . . . . _ _ _ _ _

4 l FNP-O-M- M l 4.1.1.3 surveillance Requirements i The REMP samples shall be collected pursuant to Table 4-1 from the locations

described in section 4.2, and shall be analyzed pursuant to the requirements of I Table 4-1 and Table 4-3. Program changes may be initiated based on operational

! experience. i l Analys0s shall be performed in such a manner that the stated MINIMUM DETECTABLE l CONCENTRATIONS (MDCs) will be achieved under routine conditions. Occasionally l- background fluctuations, unavoidable small sample sizes, the presence of l interfering radionuclides, or other uncontrollable circumstances may render these f* ;MDCs unachievable. In such cases, the contributing factors will be identified ) and described in the Annual Radiological Environmental Operating Report. i. ! 4.1.1.4 Basis i The REMP required by this control provides representative measurements of j radiation and of radioactive materials in those exposure pathways, and for those ! radionuclides, which lead to the highest potential radiation exposures of MEMBERS l OF THE PUBLIC resulting from the plant operation. The REMP implements Section j ! IV.B.2, Appendix I,10 CFR 50, and thereby supplements the radiological ef fluent monitoring program by measuring concentrations of radioactive materials and j levels of radiation, which may then be compared with those expected on the basis

of the effluent measurements and modeling of the environmental exposure pathways.

i j The detection capabilities required by Table 4-3 are within state-of-the-art for routine environmental measurements in industrial laboratories. I e i i i j ' 1 4-3 Gen. Rev. 13

_ , _ . - , . - - _ _ . . . . ,u a n O, U G7 H

                     "   *#                                                                                                                                                                                                    E Exposure Pathway and/or SamPl es and     Sampling and Collection Frequency                                                                                                            Type and Frequency of Analysis                                a Sample       SamPl e                                                                                                                                                                                                   e Locations

• 7

1. AIRBORNE

                                                                                                                                                                                                                               'o Particulates                 continuous operation of                                                                                                       Earticulate sampler. Analyze for                                 $,

indicator 3 sampler with sample gross beta radioactivity 224 hours y Control 2 collection weekly. following filter change. Perform - gamma isotopic analysis on each j sample when gross beta activity is g

                                                                                                                                                              >10 times the yearly mean of control                             o samples. Perform gamma isotopic analysis on composite (by location)                              g2 sample quarterly.                                                <

r Radioiodine Radioiodine canister. Analyze o Indicator 3 weekly for I-131. @ Control 2 g et

• 2. DIRECT RADIATION
• s TLD Quarterly. Gamma dose quarterly. d Indicator I 16 3 Indicator II 16 $

(community) @ Control 3 g o

3. WATERBORNE .o Surface Composite ** sample collected Gamma isotopic analysis rnonthly. e Indicator 1 monthly. Tritium analysis of composite (by "

Control 1 location) sample quarterly. $ Ground Quarterly. Gamma isotopic and tritium analysis I Indicator 1 of each sample. Control 1 8 Sediment Semiannually. Gamma isotopic analysis yearly. 3 D Indicator 1 g 5 f u S

O O O e

                         ""      # UI Exposure             es                                                                                                                            UPe and Path       anq/or    ""P             Sampling and Collection Frequency 3 pe                                                                                                                                    gnal e s Locations

• o S

4. INGESTION o.

Milk semimonthly when animals are on Gamma isotopic and I-131 Indicator 3*** pasturer monthly at other times. analysis of each sample. Control I w Fish One sample in season, or semiannually Ganna isotopic analysis E. Indicator 1 if not seasonal. One sample of each of on edible portions. g control 1 the following species: -

1. Game Fish S

2. Bottom Feeding Fish p 9

Gamma isotopic analysis " Forage or Grab sample cut from green forage or Leafy 1 vegetation monthly. which includes I-131 M a vegetation 1 analysis of each sample. S j, Indicator y Control o ile E

• H Sample locations are shown in Table 4-4, and in Figure 4-1 through Figure 4-4.

x

      **      Compogite samples shall be collected by collecting an aliquot at intervals not exceeding 2                                                                                               8 hours.                                                                                                                                                                                   p o

Up to three sampling locations within 5 miles in different sectors with the highest dose r potential will be used as available. J N a

• 5 a

                                                                                                                                                                                                           ?

?  ? ? I t: 0

                                                         . - . . - - - - . _ - _ _ - - _ _ _ . - - - - - - , - _ _ _ - - - _ _ - _ - _ - - - - - _ _ _ - _            -__._-----___-__m_,-_---_- - - -       -_

                                                 ._ . . _ . .                                                                     ~

y U, . O e E X

                                                                                                                              ?

Reporting Level F Airborne Particulate or Forage or Leafy 88 *8 Water Gaseg Fish Milk vegetation ES Analysis (pCi/L) (pCi/m ) (pCi/kg, wet) (pCi/L) (pci/kg, wet) 7,0 e et H-3 2 E+4" ar J Mn-54 1 E+3 3 E+4 { Fe-59 4 E+2 1 E+4 e a co-58 1 E+3 3 E+4  % o Co-60 3 E+2 1 E+4 " Zn-65 3 E+2 2 E+4 5 o. I Zr-95 4 E+2 e o Nb-95 7 E+2 $ I-131 2 E+0b 9 E-1 3 E+0 1 E+2 N Cs-134 3 E+1 1 E+1 1 E+3 6 E+1 1 E+3 i

                                                                                                                     \     n Cs-137                  5 E+1                    2 E+1     2 E+3            7 E+1            2 E+3            $

n Ba-140 2 E+2 3 E+2 E D La-140 1 E+2 4 E+2 l o e

a. This is the 40 CFR 141 value for drinking water samples. If no drinking water pathway y exists, a value of 3 E+4 pC1/L may be used. ,

o n y b. If no drinking water pathway exists, a value of 20 pC1/L may be used. $ 3 o 7 N R  ? a e x a a l t O l l

            .        - . - . - . . .        - - . .   .     . - ~    .- ...      . _._                                          -
                                                                                                                                    ~. -                                                   - ..       . . . . .    . . - .                                  .-     ..

l O O O Minin;um Detectable Concentration (MDC)" $- e Forage or . Airborne Leafy e Particulate Fish Vegetation P Water orGasgs (pci/kg, Milk (pC1/kg, Sediment Analysis (pci/L) (Pci/m ) wet) (pCi/L) wet) (pCi/kg, dry) a gross beta 4 E+0 1 E-2 g , e H-3 2 E+3 b o es Mn-54 1.5 E+1 1.3 E+2 $ Fe-59 3 E+1 2.6 E+2 h e Co-58, Co-60 1.5 E+1 1.3 E+2 Zn-65 3 E+1 2.6 E+2 Zr-95 3 E+1 $

 ,        Nb-95                      1.5 E+1                                                                                                                                                                                                                    E t                                                                                                                                                                                                                                                             "
 "        I-131                       1 E+0c            7 E-2                                                                                                                       1 E+0       6 E+1                                                           $
                                                                                                                                                                                                                                                               =

Cs-134 1.5 E+1 5 E-2 1.3 E+2 1.5 E+1 6 E+1 1.5 E+2 g  ! e Cs-137 1.8 E+1 6 E-2 1.5 E+2 1.8 E+1 8 E+1 1.8 E+2 g o Ba-140 6 E+1 6 E+1 D g La-140 1.5 E+1 1.5 E+1 g

• es

a. See the definition of MINIMUM DETECTABLE CONCENTRATION in Section 10.1. Other peaks which o are measurable and identifiable as plant effluents, together with the radionuclides in this D table, shall be analyzed and reported in accordance with Section 7.1.

l

b. If no drinking water pathway exists, a value of 3 E+3 pCi/L may be used.

• o O

c. Jf no drinking water pathway exists, a value of 1.5 E+1 pCi/L may be used. $ ,

JD O k t { i

I I i i i FNP-O-M-011 4.1.2 Land Use census i . In accordance with Technical specification 6.8.3.f(ii), a land use census shall j be conducted and shall identify the location of the nearest milk animall and the nearest permanent residence, in each of the 16 meteorological sectors, within a distance of 5 miles. l 4.1.2.1 Applicability  ; f-l 1 l This control applies at all times. j i j 4.1.2.2 Actions I $ l 1 i 4.1.2.2.1 With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than values currently being calculated l in accordance with Section 3.4.3, identify the new location (s) in the next Annual l ! Radioactive Effluent Release Report. ] 4.1.2.2.2 With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent I 3 } greater than at a location from which samples are currently being obtained in i 4 accordance with Section 4.1.1, add the new location (s) to the REMP within 30 da'/s $ if samples are available. The sunpling location, excluding control station location (s), having the lowest calculated dose or dose commitment (via the s ue cxposure pathway) may be deleted from the REMP if new sampling locations are cdded. Pursuant to Technical Specification 6.14 submit in the next Annua) Radioactive Effluent Release Report any change (s) in the ODCM, including the revised figure (s) and table (s) reflecting any new location (s) and information cupporting the change (s). 4.1.2.2.3 This control does not affect shutdown requirements or MODE changes. 4.1.2.3 surveillance Requirements The_ land use census shall be conducted annually, using that information which  ! will provide good results, such as a door-to-door census, a visual census from

                                                                                                                             ]'

cutomobile or aircraft, consultation with local agriculture authorities, or some combination of these methods, as feasible. Results of the land use census shall be included in the Annual Radiological Environmental Operating Report. ] 1 Defined as a cow or goat that is producing milk for human consumption. 4-8 Gen. Rev. 13 j

a

  . .h 1

I- FNP-O-M-011

  '(     (                   .4.1.2.4   Basis This control is provided to ensure that changes in the use of UNRESTRICTED AREAS are identified and that modifications to the REMP are made if required by the                                        >

results of this census. This census satisfies the requirements of Section IV.B.3 , of Appendix I to 10 CFR Part 50. j

1. ,

3 4 f n i i i 1 1 i i j i ? a i I . i

j. i i

4 1 4 l 4 iO 4 1 e 4-9 Gen. Rev. 13 i 1

n .. ... I FNP-0-M-011 4.1.3 Interlaboratory Comparison Procram

     /~

N])

In accordance with Technical Specification 6.8.3.f(iii), analyses shall be i performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which satisfies the requirements of Regulatory Guide 4.15, Revision 1, February 1979. ' i 4.1.3.1 Applicability This control applies at all times. l l 4.1.3.2 Actions With analyses not being performed as required by Section 4.1.3, report the l corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Operating Report.

                                                                                                       )

This control does not affect shutdown requirements or MODE changes. 4.1.3.3 Surveillance Requirements OA summary of the results obtained as part of the required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental i Operating Report. l 4.1.3.4 Basis The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy .of the measurements of ' radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring, in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2, Appendix I, 10 CFR 50. l 1

           )

J b 4-10 Rev. 15

g;.. 1 O FNP-O-M-011 , 4.2 RADIOLOGICAL ENVIRONMENTAL MONI7'1 ING LOCATIONS Table 4-4, and Figure 4-1 through Figv: i 4-4 specify the locations at which the measurements and samples are taken for the REMP required by Section 4.1.1. l 1 l i l 1 O 4-11 Gen. Rev. 13

b FMP-0-M-011 Table 4-4. Radiological Environmental Monitoring Locations (U Exposure E"

             ,nd h aY                  SamPl ing Locations

• Ide$tigi.

r Sample cation

1. AIRBORNE Partic- Indicator Stations ulates River Intake Structure (ESE-0.8 miles)I PI-0501 South Perimeter (SSE-1.0 miles) PI-0701 Plant Entrance (WSW-0.9 miles) PI-1101 North Perimeter (N-0.8 miles) PI-1601 Control Stations:

Blakely, CA (NE-15 miles) PB-0215 Dothan, AL (W-18 miles) PB-1218 Neals Landing, FL (SSE-18 miles)I PB-0718 Community Stations: Georgia Pacific Paper Co. (SSE-3 miles) PC-0703 Ashford, AL (WSW-8 miles) PC-1108 Columbia, AL (N-5 miles) PC-1605 Radiciodine Indicator Stations: River Intake Structure (ESE-0.8 miles)I II-0501 South Perimeter (SSE-1.0 miles) II-0701 Plant Entrance (WSW-0.9 miles) II-1101 l North Perimeter (N-0.8 miles) II-1601 '

   /N               Control Stations:

Blakely, GA (NE-15 miles) IB-0215 Dothan, AL (W-18 miles) IB-1218

• Neals Landing, FL (SSE-18 miles)I IB-0718 Community Stations:

Georgia Pacific Paper Co. (SSE-3 miles)2 IC-0703

2. DIRECT RADIATION TLD Indicator I Stations:

Plant Perimeter (NNE-0.9 miles) RI-0101 l (NE- 1.0 miles) RI-0201 (ENE-0.9 miles) RI-0301 (E- 0.8 miles) RI-0401 (ESE-0.8 miles) RI-0501 (SE- 1.1 miles) RI-0601 (SSE-1.0 miles) RI-0701 (S- 1.0 miles) RI-0801 j (SSW-1.0 miles) RI-0901 (SW- 0.9 mtima) RI-1001 l (WSW-0.9 miles) RI-1101 (W- 0.8 miles) RI-1201 (WNW-0.8 miles) RI-1301 (NW- 1.1 miles) RI-1401 (NNW-0.9 miles) RI-1501 (N- 0.8 miles) RI-1601 O V 4-12 Gen. Rev. 13

i i

     /
   ,                                                                               FNP-0-M-011 i

(~'T Table 4-4 (contd). Radiological Environmental Monitoring Locations v' Exposure Pathway 3,,py, and/or Sampling Locations

• Identifi-Sample cation l'

TLD (contd) Control Stations: Blakely, GA (NE-15 miles) RB-0215 Neals Landing, FL (SSE-18 miles) RB-0718 Dothan, AL (W-15 miles) RB-1215 Dothan, AL (W-18 miles) RB-1218 4 Webb, AL (WNW-11 miles) RB-1311 j Haleburg, AL (N-12 miles) RB-1612 1 Indicator II (Communitv) Stations:

              '                 (NNE-4 miles)                                   RC-0104 (NE- 4 miles)                                   RC-0204 (ENE-4 miles)                                   RC-0304 l

(E- 5 miles) RC-0405 (ESE-5 miles) RC-0505 i (SE- 5 miles) RC-0605 (SSE-3 miles) RC-0703 (S- 5 miles) RC-0805 (SSW-4 miles) RC-0904 (SW- 1.2 miles) RC-1001 (SW- 5 miles) RC-1005 (WSW-4 miles) RC-1104 {g (WSW-8 miles) RC-1108

            )                  (W-   4 miles)

(WNW-4 miles) RC-1204 RC-1304 . (NW- 4 miles) RC-1404 i (NNW-4 miles) RC-1504 (N- 5 miles) RC-1605 ]

3. WATERBORNE Surface Indicator Station:

Georgia Pacific Paper Co. Intake Structure WRI (River Mile - 40) Control Station: Andrew Lock & Dam Upper Pier (River Mile - 47) WRB Ground Indicator Station: Georgia Pacific Paper Co. Well (SSE-4 miles) WGI-07 i Control Station: Whatley Well (SW-1.2 miles) WGB-10 Sediment Indicator Station: Smith's Bend (River Mile - 41) RSI Control Station: Andrews Lock & Dam Reservoir (River Mile - 47) RSB 4 O 4-13 Gen. Rev. 13

,. w. -- FNP-0-M-011 Table 4-4 (contd). Radiological Environmental Monitoring Locations i V Exposure Pathway Sample and/or Sampling Locations. Identifi-Sample cation

4. INGESTION Milk Indicator Station:

None (There are no milk animals within 5 miles per the current land use survey) Control' Station: Bruce Ivey Dairy Webb, AL (W-12 miles) MB-1212 Fish . Indicator Station: Smith Bend (River Mile - 41) Game Fish Bottom Feeding Fish FCI FBI Control Station: Andrews Lock & Dam Reservoir (River Mile - 47) Game Fish , Bottom Feeding Fish FGB 1 FBB Forage or Indicator Stelions: Leafy Vegetation South Southec~t Pertneter (ESE-1.0 miles) FI-0701 l North FI-1606 South Perimeter Perimeter (N-0.8 (S-1.0 miles)8 miles) FI-0801 Northeast Perimeter (NE-1.0 miles)8 FI-0201 Control Station:

     ~~g                       Dothan, AL (W-18 miles)
  '                                                                                   FB-1218 Distance Unit  1 andand  direction Unit  2 plantas ventmeasured stacks. from the centerpoint between

1. Not required by Section 4.1.1. Used as a spare station.

2. Not required by Section 4.1.1.

State of GA EPD. Use for comparison purposes with

3. Alternate forage plots. I O

I 4-14

    %s/                                                                                     Rev. 14

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i I I f 1 4 l 4 fetast SMIPumE A assassPues

                                        @ Ra, Palmtuuns a meet samrues O

Figure 4-1. Airborne Sampling Locations, 0-5000 feet

    ?

4-15 Gen. Rev. 13

, . . . . . ~ _ . . - . - - . - - . . - . . . . . . - . . . - - - _ . . . . . . - _ - - - . . - . ~ . - - . . _ . . . - . _ - ~ . - . 1 I 1 I FNP-0-M-011 i l I l i , ! ~ I i a Is euant l f !- is a f i

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t I f f i I i i ggg saa e snas O Figure 4-2. Indicator II (comununity) Sampling Locations for Direct Radiation 4-16 Gen. Rev. 13

            . _.      . _.            ~.                . . _ .                . . - . . . - . . . . .                        .. . ... .            - - - . .                  ~ .   . . _ . . . - - - .

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                                                                                                                                                                        & M sAIIPUEs g na pasmanmes ans menisaruus
                                                                                                                                                                        # mas naurums i

l Figure 4-3. Airborne Sampling Locations, 0-20 miles D . 4-17 Rev. 14

_ . _ _ . - _ . _ _ - - . __ _ . _ _ . . . . _ _ _ _ . _ . . _ . _ _ . . _ _ . . . . - .._m__ _ _ _ . ... _ ..._ _ ______ . _ _ _ _ . _ , _ _ . 9 FNP-O-M-011 h IO i l 1

l i

l t M 18 L i lacs a ass tosus jI n*gsas Marr

                                      ~~m                                                         ( saa m,m O                             annone        l
                                                               $p l
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ausame ISICATOR NATIOuS COIITIOL STATIOR$ ) E assessausensusimum essenunus, sus me sunsen 5 seu aus sumust esame muse O Figure 4-4. Water Sampling Locations 4-18 Gen. Rev. 13

4 4 i FNP-0-M-011  ! CHAPTER 5 TOTAL DOSE DETEPJiiNATIONS 5.1 LIMIT OF OPERATION d In accordance with Technical Specification 6.8.3.e(x), the dose or dose l commitment to any MEMBER Ol' THE PUBLIC over a calendar year, due to releases of radioactivity and to radiation from uranium fuel cycle sources, shall be limited to less than or equal to 25 mrom to the total body or any organ, except the I thyroid, which shall be limited to less than or equal to 75 mrom. 1 l l 5.1.1 Aeolicability This limit applies at all times. 5.1.2 Actions With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Section 2.1.3, 3.1.3, or 3.1.4, calculations shall be made according to Section 5.2 methods to determine whether p'L/ the limits of Section 5.1 have been exceeded. If these limits have been ,j cxceeded, prepare and submit a Special Report to the Nuclear Regulatory l Commission, pursuant to Technical Specification 6.9.2, within 30 days, which defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Section 5.1 and includes the cchedule for achieving conformance with the limits of Section 5.1. This Special Report, as defined in 10 CFR 20.2203, shall also include an analysis which catimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources (including all ef fluent pathways and direct radiation) for the calendar year that includes the release (s) covered by this report. This Special Report shall also describe the 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 limits of Section 5.1, and if the release condition resulting in violation of the provisions ot 4CLCEL19Q. has not already been corrected, the Special Report shall include a request for variance in accordance with the provisions of 40 CFR 190 and including the cpecified information of 40 CFR 190.11(b) . Submittal of the report is considered a timely request, and a variance is granted until staf f action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as O cddressed in other sections of this ODCM. 5-1 Gen. Rev. 13

FNP-O-M-011

    ; This control does not affect shutdown requirements or MODE changes.

5.1.3 surveillance Raouirements Cumulative dose contributions from liquid and gaseous effluents and from direct radiation shall be determined in accordance with Section 5.2. This requirement-is applicable only under the conditions set forth above in Section 5.1.2. l 5.1.4 Basis , l 1 This control is provided to meet the dose limitations of 40 CFR 190. The control requires the preparation and submittal of a Special Report whenever the calculated doses ftom plant radioactive effluents combined with doses due to direct radiation from the plant exceed the limits of 40 CFR 190. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a MEMBER OF THE PUBLIC for a calendar year to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment f to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible with the exception that dose contributions from other uranium fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle. .l i* , i 5-2 Gen. Rev. 13 - l-d

                    -                                   -                             r- - -

   ._ _ _ . _ . _ _ . - . - _ _ . . . _ _ _ . - . - . . . . _ . , . ~ . _ . - - - . . .                                  . _ _ _ _ _ _ . . . _ _ _ - -

4 m 4 L } i

FNP-0-M-011 ;

!! 5.2 DEMONSTRATION OF COMPLIANCE l t i

There are no other uranium fuel cycle f acilities within 5 miles of the plant [

j oite. Therefore, for the purpose of demonstrating compliance with the limits of j- Section 5.1, the total dose to a MEMBER OF THE PUBLIC in the vicinity of the ' l plant site due to uranium fuel cycle sources shall be determined as follows:  ; f i } Dn=DL+Da+Dp+Dy (5.1) l- - }. i l where: ' c > 4 Dn = the total dose or. dose commitment to the total body or organ k, in mrom. J' D,=t the dose to the same organ due to radioactivity discharged from the ' plant site in liquid effluents, calculated in accordance with Section 2.4.1, in mrom. l. jl Dg = the dose to the same organ due to non-noble-gas radionuclides  ; !. discharged from the plant site in gaseous effluents, calculated for the controlling receptor in accordance with section 3.4'.3, in mrem. I i j DD= the direct radiation dose to the whole body of an individual at the ' t ,

d. controlling receptor location, due to radioactive materials

retained within the plant site, in arem, Values of direct j radiation dose may be determined by measurement, calculation, or i l a combination of the two.

i h DN_= the external whole body dose to an individual at the controlling 'j. receptor location, due to gamma ray emissions from noble gas radio-nuclides discharged from the plant site in gaseous effluents, in I l 7 area. Dg is calculated as followe (equation adapted from Reference 1, page 22, by re-casting in cumulative dose form): DN = 3

• 17
• 10 E ( 7ElvP b #l* I" v i where:

5-3 Gen. Rev. 13

i FNP-O-M-011 3.17 x 10-8 = a units conversion factor: 1 y/(3.15 x 107 s). Qyi

                         =   the cumulative release of noble gas radionuclide i from j

release pathway v (pci), during the period of interest. Kg = the total-body dose factor due to gamma emissions from l noble gas radionuclide 1 (mrem /y)/(pci/m 3), from Table 3-5. (176)yp = annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, j 3 from Table 3-7 (s/m ]. I i As defined above, DL and Do are for different age groups, while DD and Dg are -l not age group specific. When a more precise determination of D3 is desired, values of DL and Dg may be calculated for all four age groups, and those values used in equation (5.1) to determine age group specific values of Dg; the largest value of Dn for any age group may then be compared to the limits of Section 5.1. I C') b 5-4 Gen. Rev. 13

i FNP-O-M-Oli i CHAPTER 6 i /

  \                                                  POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 4

i 6.1 REQUIREMENT FOR CALCULATION current FNP effluent controls as established by this ODCM do not require assessment of the radiation doses from radioactive liquid and gaseous affluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY j (Figure 10-1). However, when such an assessment is desired, it should be l performed in accordance with Section 6.2. j 6.2 CALCULATIONAL METHOD 2 For the purpose 'of performing the calculations required in Section 6.1, the dose ) to a member of the public inside the SITE BOUNDARY shall be determined at the locations, and for the receptor age groups, defined in Table 6-1. The dose to such a receptor at any one of the defined locations shall be determined as follows: 6 O ' b DIk " * (DA+DS+DP)*T o (' 1) 1 l where: I Dg = the total dose to the total body or organ k, in mrom. DA= the dose to the same organ due to inhalation of non-noble-gas radionuclides discharged from the plant site in gaseous affluents, calculated in accordance with section 3.4.3, in mrom. The (17 ) i value to be used is given for each receptor location in Table 6-1; depleted (176) values may be used in calculations for non-noble-gas radionuclides. i i D3= the dose to the same organ due to around clane deoosition of non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrom. The (D76) value to be used is given for each receptor location in Table 6-1. O 6-1 Gen. Rev. 13

i FNP-0-M-011 l ( op = the external whole body dose due to gamma ray emissions from noble gas radionuclides discharged from the plant site in gaseous effluents, calculated using equation (5.2), in mrom. The (X7Q) values that are to be used are given for each receptor location in Table 6-1. Fo = the occupancy factor for the given location, which is the fraction of the year that one individual MEMBER OF THE PUBLIC is assumed to be present at the receptor location [unitless). Values of Fo for j l each receptor location are included in Table 6-1. ' i I l

                                                                                    \

l l l i t i G 6-2 Gen. Rev. 13

FNP-0-M-011  ! Table 6-1. Attributes of Member of the Public Receptor Locations Inside the SITE BOUNDARY b Location: Visitor Center, WSW at 0.19 miles Ace Grouo: Child i Occuoancy Factors 1.37 E-03 (based on 12 hours per year) I piscersion and Deoosition Parameters i 1 l Parameter Ground-Level Mixed-Mode j i (176), s/m3 1.04 E-04 8.80 E-06 (676), ni2 4.80 E-07 6.20 E-08 ) Locations service Water Pond, SSW at 0.60 miles . Ace Grouo: Child Occucancy Factor 7.53 E-03 (based on 66 hours per year)

, 3 Discersion and Decosition Parameters:                                                           '

i Parameter Ground-Level Mixed-Mode (176), s/m3 4.74 E-05 9.75 E-07 (676), ac2 1.31 E-07 2.78 E-08 6-3 Gen. Rev. 13

FNP-0-M-011 l l Table 6-1 (contd). l Attributes of Member of the Public Receptor Locations 1 ( Inside the SITE BOUNDARY I Incation: River Water Discharge, SE at 1.02 miles Ace Grouo Child Occupancy Factor: 1.14 E-02 (based on 100 hours per year) Discersion and Deposition Parameters: Parameter Ground-Level Mixed-Mode (176), s/m3 1.63 E-05 7.05 E-07 (676), m-2 4.55 E-08 1.39 E-08

 /'s i

( i e

   /~                                      .                 .

6-4 Gen. Rev. 13

   . - - . -          _ .      . - . -          . - . . - . - - - _ _ - - - - . . -                                  - - - - - - ~ . - .

FNP-0-H-011 CHAPTER 7 j REPORTS 1 l $- 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT l 7.1.1 Recuirement for Reoort In accordance with Technical Specifications 6.9.1.6 and 6.9.1.7, the Annual Radiological Environmental Operating Report covering the REMP activities during the previous calendar year shall be submitted before May 1 of each year. (A cingle report fulfills the requirements for both units.) The material provided a chall be consistent with the objectives outlined in Section 4.1 and Section 7.1.2 ! cf the ODCM, and in Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. t j, 7.1.2 Rooort Contents } The materials specified in the following sub-sections shall be included in each Annual Radiological Environmental Operating Reports 7.1. 2 .1 - Data 1 { The report shall include summarizad and tabulated results of all REMP samples } t required by Table 4-1 taken during the report period,-in a format similar to that i contained in Table 3 of the Radiological Assessment Branch Technical Position (Reference 13); the results for any additional samples shall also be included. j In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing 4 j results; the missing data shall be submitted as soon as possible in a ! eupplementary report. The results for naturally-occurring radionuclides not f included in plant effluents need not be reported. i-7.1.2.2 Evaluations j Interpretations and analyses of trends of the results shall be included in the report, including the following (as appropriate) comparisons with pre-

l - operational studies, operational controls, and previous environmental operating reports; and an assessment of any observed impacts of the plant operation on the j cnvironment. If the measured level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4-2 is not the result of plant l O offluents, the condition shall be described as required by Section 4.1.1.2.2.

i 7-1 Gen. Rev. 13 c_- . . -

 . - .     - - , - .         .    .    . _ . _ ~   _ . . . .  . . - --        _    - . . - _ - . . .               .. -- -

t I I l FNP-0-M-011 I 7.1.2.3 Programmatic Information Also to be included in each report are the following: a summary description of the REMP; a map (s) of all sampling locations keyed to a table giving distances and directions from the center point between the Unit 1 and Unit 2 plant vent stacks; the results of land use censuses required by Section 4.1.2; and the results of licensee participation in the Interlaboratory Comparison Program j required by Section 4.1.3. J 7.1.2.4 Descriptions of Program Deviations H= Discussions of deviations from the established program must be included in each report, as follows: 7.1.2.4.1 If the REMP is not conducted as required in Table 4-1, a description of the reasons for not conducting the program as required, and the plans for preventing a recurrence, must be included in the' report. 7.1.2.4.2 If the MDCs required by Table 4-3 are not achieved, the contributing factors must be identified and described in the report. l l l l 7.1.2.4.3 If Interlaboratory Comparison Program analyses are not performed i as required by Section 4.1.3, the corrective actions taken to prevent a recurrence must be included in the report. I

                                                                                                                 \

I l iO 7-2 Rev. 15 l -

i 4 i FNP-0-M-011 7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT l l 7.2.1 Raouirement for Reoort i In accordance with Technical Specifications 6.9.1.8 and 6.9.1.9, the Annual Radioactive Effluent Release Report covering the operation of the units during 5 1 l the previous calendar year of operation shall be submitted before May 1 of each year. ( A single submittal may be made for Units 1 and 2. However, the submittal chall specify the. releases of radioactive material in liquid and gaseoue ) offluents from each unit and solid radioactive waste from the site. ) The report chall include a summary of the quantities of radioactive liquid and gaseous

- offluents and solid waste released from the units. The material provided shall j be consistent with the objectives outlined throughout this ODCM and the Process I Control Program (PCP) and in conformance with 10 CFR Part 50.36a and Section

. IV.B.1 of Appendix I to 10 CFR Part 50.

1 f 7.2.2 Reoort Contents l The materials specified in the following sub-sections shall be included in each $ Annual Radioactive Effluent Release Report: D 7.2.2.1 Quantities of Radioactive Materials Released .j ! The report shall include a summary of the quantities of radioactive liquid and gaseous ef fluents and solid waste released from the units as outlined in NRC Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with liquid and gaseous effluent data summarized on a quarterly basis and solid radioactive waste data summarized on a semiannual basis following the format of Appendix B thereof. Unplanned releases of radioactive materials in gaseous and liquid effluents from the site to UNRESTRICTED AREAS shall be included in the report, tabulated either by quarter or by event. For gamma emitters released in liquid and gaseous effluents, in addition to the principal gamma emitters for which MDCs are specifically established in Table 2-3 and Table 3-3, other peaks which are measurable and identifiable also shall be identified and reported. 7.2.2.2 Meteorological Data The report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form O. of an hour-by-hour listing of wind speed, wind direction, and atmospheric ctability, and precipitation (if measured) on magnetic tape; or in the form of 7-3 Gen. Rev. 13

5 i .] '[ FNP-0-M-011

. (/
;            joint frequency distributions of wind speed, wind direction, and atmosph6ric stability. In lieu of submission with the Annual Radioactive Effluent Melcate

Report, the licensee has the option of retaining this summary of required

? i meteorological data on site in a file that shall be provided to the NRC upon { request.  ! j! ! 7.2.2.3 Dose Assessments i ( The report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from each unit during the previous calendar ysar. Historical annual average meteorology or the meteorological conditions concurrent with the time of release of radioactive materials in gaseous eftluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway dose. This

assessment of radiation doses shalf be performed in accordance with Sections

( 2.1.3, 2.4, 3.1.3, 3.1.4, 3.4.2, 3.4.3, 5.1, and 5.2.

   ,        If a determination is required by Section 5.1.2, the report shall also include 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 pcthways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation; this dose assessment must be performed in accordance with Chapter 5.

7.2.2.4 Solid Radwaste Data For mach type of solid waste shipped offsite during the report period, the followicq information shall be included:

a. Containec volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. Solidification agent (e.g., cement, urea formaldehyde.)

m 7-4 Gen. Rev. 13

W<,

  ,A 4

FNP-0-M-011 7.2.2.5 Licensee Initiated Document Changes Licensee initiated changes shall be submitted to the Nuclear Regulatory Commission as a part of or concurrent with the Annual Radioactive Effluent l Release Report for the period in which any changes were made. Such changes to l l the ODCM shall be submitted pursuant to Technical Specification 6.14. This requirement includes: J 7.2.2.5.1 Any changes to the sampling locations in the radiological environmental monitoring program, including any changes made pursuant to Section l 4.1.1.2.3. Documentation of changes made pursuant to Section 4.1.1.2.3 shall include supporting information identifying the cause of the unavailability of camples. } 7.2.2.5.2 Any changes to dose calculation locations or pathways, including I 1 , cny changes made pursuant to Section 4.1.2.2.2. 1 7.2.2.6 Descriptions of Program Deviations Discussions of deviations from the established program shall be included in each report, as follows:

          , 2.2.6.1     The report shall include deviations from composite sampling res;irements included in Table 2-3 and Table 3-3.

I 7.2.2.6.2 The report shall include deviations from Minimum Detectable concentration (MDC) requirements included in Table 2-3 and Table 3-3. 7.2.2.7 Major Changes to Radioactive Waste Treatment Systems ' As required by Sections 2.1.5 and 3.1.6, licensee initiated MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (liquid and gaseous) shall be reported to the Nuclear Regulatory Commission in the Annual Radioactive Effluents Release Report covering the period in which the change was reviewed end accepted for implementation.I

      /
  'k w

1 In lieu of inclusion in the Annual Radioactive Effluents Release Report, I this same information may be submitted as part of the annual FSAR update. l 7-5 Gen. Rev. 13 i

                                                                         ._ ., -. _         _ - . . - ~

igi;n - ' pge p a(4 1 M// FNP-O-M-Oli I ( The discussion of each change shall contains 7.2.2.7.1 A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; 7.2.2.7.2 Suf ficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; i 7.2.2.7.3 A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; 7.2.2.7.4 An evaluation of the change which shows the predicted releases-of radioactive materials in liquid and gaseous effluents that differ from those previously predicted in the license application and amendments thereto; 7.2.2.7.5 An evaluation of the change which shows the expected maximum exposures to MEMBERS OF THE PUBLIC in the UNRESTRICTED AREAS and to the general population that differ from those previously estimated in the license application and amendment;s thereto; 7.2.2.7.6 A comparison of the predicted releases of radioactive materials, in O- liquid and gaseous effluents, to the actual releases for the period prior to when , the cht.nges are to be made; 7.2.2.7.7 An estimate of the exposure to plant operating personnel as a result of the change; and 7.2.2.7.8 Documentation of the fact that the change was reviewed and found i acceptable by the PORC. l Q l 7-6 Gen. Rev. 13

of . - . . , _ . 1

j1 . ';j:

l eg :pM.,

      $3 jr;
      /fp FNP-0-M-011
  .lNY
    /             7.3    MONTHLY OPERATING REPORT                                                       '

1.

  $1 1* V

• This ODCH establishes no requirements pertaining to the Monthly Operating Report. )

1 1 4 7.4 SPECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Commission in cecordance with Technical Specification 6.9.2, as required by Sections 2.1.3.2, 2.1.4.2, 3.1.3.2, 3.1.4.2, 3.1.5.2, 4.1.1.2.2, and 5.1.2. I E j' . i l i U O V 7-7 Gen. Rev. 13

                                                                                 ~
      ,77 q
     .s I,

FNP-O-M-011 CHAPTER 8 (/ METEOROLOGICAL MODELS The models presented in this chapter are those which were used to compute the specific values of meteorology-related parameters that are referenced throughout this ODCM. These models should also be used whenever it is necessary to calculate values of these parameters for new locations of interest. NOTE: Although Plant Farley has no pure elevated releases, the sections on elevated-mode calculations (8.1.2 and 8.2.2) are included for convenience in calculating mixed-mode values, and to preserve section number compatibility with the ODCMs of the other plants in the Southern Nuclear System. 8.1 ATMOSPHERIC DISPERSION Atmospheric dispersion may be calculated using the appropriate form of the sector-averaged Gaussian model. Gaseous release elevations may be considered to be either at ground-level, elevated, or mixed-mode. Facility release elevations O jr for each gaseous release point are as indicated in Table 3-4. I 8.1.1 Ground-Level keleases i Relative concentration calculations for ground-level releases, or for the ground-4 level portion of mixed-mode releases, shall be made as follows: 2.032 6 K r Djk (X/Q)(; = { (8.1) N r UJ Ed where i (X/Q)G = the ground-level sector-averaged relative concentration for a 1 given wind direction (sector) and distance (s/m 3 ), ! 2.032 = (2/fr)1/2 divided by the width in radians of a 22.5* sector, which l is 0.3927 radians. 4 6= the plume depletion factor for all radionuclides other than noble gases at a distance r shown in Figure 8-3. For noble gases, the i depletion f actor is unity. If an undepleted relative concentration 8-1 Gen. Rev. 13

FNP-0-M-011 I i is desired, the depletion factor is unity. Only depletion by 8% deposition is considered since depletion by radioactive decay would i be of little significance at the distances considered. 1 g= the terrain recirculation factor corresponding to a distance r,  ! taken from Figure 8-2. i njk = the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed. N= the tot 21 hours of valid meteorological data recorded throughout the period of interest for all sectors, wind speed classes, and stability categories. uj = the wind speed (mid-point of wind speed class j) at ground level (m/s). r= the distance from release point to locatien of interest (m). , I Ed= the vertical standard deviation of the plume concentration j distribution considering the initial dispersion within the building wake, calculated as follows:

                                                  '2    b2'l/2

[d 3, Eg = the lesser ofs ' og 0*O (Czk) od = the vertical standard deviation of the plume concentration distribution (m) for a given distance and stability category k as shown in Figure 8-1. The stability category is determined by the vertical temperature gradient AT/As ( *c/100 m or 'F/100 ft) . Plant Farley AT/Az values must be adjusted for As of 165 ft. tr = 3.1416 b.= the maximum height of adjacent plant structure, which is the containment building (40 m). O 8-2 Gen. Rev. 13

_ . ._ __ . ~_ _ . _.. _ _ _ _ -._.. . . - _ _ _ _ _ FNP-0-M-011 8.1.2 Elevated Releases t O. Relative dispersion calculations for elevated releases, or for the elevated portion of mixed-niode releases, shall be made as follows:

                                                                                -h 2' 2.032 K r                6g nji exP (X/0)g =                  [                             2
                                                                             ,2 ozk.                (s.3) jk
                                                                      ") C zk where:                                                                                                              '

(X/Q)E = the elevated release sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3) . 6k= the plume depletion factor for all radionuclides other than noble gases at a distance r for elevated releases, as shown in Figure 8-4, Figure 8-5, and Figure 8-6. For an elevated release, this factor is stability dependent. For noble gases, the depletion t factor is unity. If an undepleted relative concentration is desired, the depletion factor is unity. Only depletion by ' deposition is considered since depletion by radioactive decay would be of little significance at the distances considered. njk = the number of hours that wind of wind speed class j is directed 1 into the given sector during the time atmospheric stability category k existed. j uj = the wind speew (mid-point of wind speed class j) at the effective release height h (m/s). h= the effective height of the release (m), which is calculated as follows: h=h y + hp , - h, - c y (s.e} hy = the height of the release point (m). O 83 Gen. Rev. 13

   .. --     . - . . - ..        -- . - - . . - .                      .      _ . - = - . . - _ . - . . - .         . .    ~ . - - . .-

i FNP-0-M-011 [ hg= the maximum terrain height between the release point and the point of interest (m), from Figure 2.3-26 of Reference 7. h,= p the additional height due to plume rise (m) which is calculated as I j j follows and limited by hpr ( M 8 4 r 2 1 y# -3 hpr = 1. 44 d -. 3 (8.5) r  %, ' i j j i f 9 i 1 w# 3 - *d

                                                                                     ,   *) ,

hp,(max) = the lesser of: OR (***)

                                                                                          ,         ,1        1
                                                                                             ,m             -

J 1.5 33 6

                                                                           ,               , 9,
  'n Q                                                                                                                                     i d=

the inside diameter of the vent (m). 4 Wo= the exit velocity of the plume (m/s), i ! cy = the correction for low vent exit velocity (m), which is calculated 4 l as follows: f ' W# W 3 1. 5 -

• d for _# < 1. 5

                                                            "}s                             ")

, cy = og (S*T> e i W 0 for # 21.5

                                                                                            ")

l Fm= the momentum flux parameter 4(m2/s ), which is calculated as follows

,               (under the assumption that the affluent air and the ambient air have the same density):

j 8-4 Gen. Rav. 13

y , b b- ? I 1 FNP-O-M-011 [ Em " (Wo ) * (8 8) S= the stability parameter, which is calculated as follows: S , p + 9. 8 x 10-3 (8,9)

                                  =f9.8T       as 1

T= the ambient air temperature ('K). (AT/Az) = the rate of increase of the ambient air temperature with  ; increasing height above the ground (*K/m). I i All other symbols are as previously defined in Section 8.1.1. l 8.1.3 Mixed-Mode Releases I l Relative dispersion calculations for mixed-mode releases shall be made as ' follows; (X/0)y * (1 -E) * (X/0)E

• E'(X/0)G (s.10) wh.r..

(X/Q)M = the mixed-mode release sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3), E= the fraction of hours during which releases are considered as ground-level releases, calculated as follows: l l O 8-5 Gen. Rev. 13

g _ . ,2.__. _.. ...__ ___... _. _._.. _ - -

                                                                    . . _ . . _ _ . . ~ . . _ _ . _ . .         - _ _ _ _____..... _ _ _ _ _ .. ,

{ FNP-O-M-Oli 1 l i i j 1 I j W 0 1.0 for o 51.0 i ") i ! r 3 I 2.58 - 1.58 for 1. 0 < b 5 1. 5

'                                        E    = '

I' ") I (8.11)  ! l O.3 - 0.06 for 1.5 < b s 5.0

                                                                ~
                                                                    ")-                                 uj                                           l i                                                                                                W a

0 for o > 5.0 f "1 i i All other symbols are as previously defined, i i

                                                                                                                                                    )

1 9 W s 8-6 - Gen. Rev. 13

FNP-O-M-011 8.2 RELATIVE DEPOSITION

 ?%

. ('^,/ Plume depletion may be calculated using the appropriate form of the sector-cveraged Gaussian model. Gaseous release elevations may be considered to be oither at ground-level, elevated, or mixed-mode. Facility release elevations for cach gaseous release points are as indicated in Table 3-4. 8.2.1 Ground-Level Releases Relative deposition calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows: l P 2.55 D g K r (D/0)G " N r ) k (8.12) k where: (D/Q)g = the ground-level sector-averaged relative deposition for a given wind direction (sector) and distance (m-23 , 2.55 = the inverse of the number of radians in a 22.5* sector b) ( [= (2 w/16)-I) . Dg = the deposition rate at distance r, taken from Figure 8-7 for ground-level releases (m-I) . nk = the number of hours in which the wind is directed into the sector of interest, and during which stability category k exists. All other symbols are as defined previously in Section 8.1. 8.2.2 Elevated Releases Relative deposition calculations for elevated releases, or for the elevatrad portion of mixed-mode releases, shall be made as follows: 2.55 K (D/Q)E N r )) (#2k Dek) (8.13) g 1 where: fs

 \->

1 1 8-7 Gen. Rev. 13 I l J

~. - - _._. _ _ __ ,

                                                                                                      \

FNP-0-M-011 i (D/Q)E' = the elevated-plume sector-averaged relative deposition for a

- given wind. direction (sector) and distance (m-2) .

1 i Dd= the elevated plume deposition rate at distance r, taken from f Figure 8-8, Figure 8-9, or Figure 8-10, as appropriate to the plume { effective release height h defined in Section 8.1.2, for stability. I I class k' (m-I) . l All other symbols are as defined previously. 1' I f 8.2.3 Mixed-Mode Releases

      -Relative deposition calculations for mixed-mode releases shall be made as follows:

i, (D/0)y = (1 -E) * (D/Q)E + E * (D/Q)g (8.14) where: (D/Q)M = the mixed-mode release sector-averaged relative deposition for a given wind direction (sector) and distance (m-2), s E= the fraction of hours during which releases are considered as I ground-level releases, defined in Section 8.1.3. All other symbols are as previously defined, i 1 l l 8.3 ELEVATED PLUME' DOSE FACTORS These factors are not required in effluent dose calculations for FNP due to the  ! fact that all gaseous effluent releases are either ground-level or mixed-mode. 8.4 -METEOROLOGICAL

SUMMARY

A sununary of meteorological data for the years 1971 through 1975 is presented in Table 8-2 through Table 8-5. 8-8 Gen. Rev. 13

I l l I i FNP-0-M-011 ) j( t Table 8-1. Terrain Elevation Above Plant Site Crade i ) I I .l i 1 4 i l I l i I l l 1 J s j i 4 1 1 4 e I i J i l This table intentionally left blank.

• a 1 1

1 1 8-9 Gen. Rev. 13

                                                                     .-             . . _ ~ ..
}

FNP-0-M-011 Table 8-2. Annual Average (X76) for Mixed Mode Releases

  \'-)                                   Distance to Location, in miles Sector 0.25-0.5      0.5-0.99        1.0-1.49    1.5-1.99    2.0-2.49 N       2.16   E-06   9.21 E-07        5.92 E-07   3.83 E-07   2.42 E-07 NNE       2.35   E-06    1.02 E-06       6.18 E-07  3.82 E-07    2.34 E-07 NE        2.23   E-06   9.61 E-07        6.06 E-07   3.86 E-07   2.40 E-07 ENE         1.12  E-06    5.03 E-07       3.76 E-07  2.65 E-07    1.76 E-07 E        1.20  E-06    5.21 E-07       3.57 E-07  2.45 E-07    1.60 E-07 ESE         1.55  E-06   6.43 E-07        3.83 E-07  2.44 E-07    1.55 E-07 SE       2.47  E-06    9.69 E-07        5.52 E-07  3.47 E-07 2.19 E-07 SSE       2.77   E-06    1.08 E-06       6.57 E-07  4.34 E-07    2.81 E-07 s       2.50  E-06    9.37 E-07        5.90 E-07  4.09 E-07    2.74 E-07 SSW       2.02   E-06   8.29 E-07        6.30 E-07  4.16 E-07    2.66 E-07 SW       2.05  E-06    8.34 E-07        8.03 E-07   5.07 E-07   3.16 E-07 WSW         1.89  E-06   7.41 E-07        7.33 E-07  4.66 E-07    2.88 E-07 W         1.67  E-06   6.74 E-07        5.81 E-07  4.12 E-07    2.53 E-07 WNW         1.43  E-06    5.97 E-07       4.11 E-07  3.13 E-07    2.17 E-07 NW        1.32  E-06    5.65 E-07       3.88 E-07  2.68 E-07    1.77 E-07 NNW         1.66  E-06   7.21 E-07        4.85 E-07  3.23 E-073   2.07 E-07 Distance to Location, in miles Sector

( 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 (~ / N 1.65 E-07 1.24 E-07 1.01 E-07 9.11 E-08 8.27 E-08 NNE 1.55 E-07 1.15 E-07 9.23 E-08 8.28 E-08 7.48 E-08 NE 1.61 E-07 1.19 E-07 9.62 E-08 8.63 E-08 7.79 E-08 ENE 1.22 E-07 9.z8 E-08 7.61 E-08 6.88 E-08 6.24 E-08 E 1.12 E-07 8.54 E-08 7.09 E-08 6.43 E-08 5.86 E-08 ESE 1.07 E-07 8.13 E-08 6.75 E-08 6.12 E-08 5.58 E-08 SE 1.51 E-07 1.14 E-07 9.50 E-08 8.61 E-08 7.88 E-08 SSE _, 1.96 E-07 1.50 E-07 1.26 E-07 1.15 E-07 1.05 E-07 S 1.96 E-07 1.52 E-07 1.29 E-07 1.18 E-07 1.09 E-07 SSW 1.84 E-07 1.39 E-07 1.22 E-07 1.18 E-07 1.08 E-07 SW 2.13 E-07 1.60 E-07 1.30 E-07 1.27 E-07 1.15 E-07 WSW 1.92 E-07 1.57 E-07 1.26 E-07 1.13 E-07 1.02 E-07 W 1.68 E-07 1.69 E-07 1.34 E-07 1.19 E-07 1.08 E-07 WNW 1.74 E-07 1.72 E-07 1.35 E-07 1.21 E-07 1.09 E-07 NW 1.37 E-07 1.24 E-07 1.18 E-07 1.06 E-07 9.60 E-08 NNW 1.42 E-07 1.07 E-07 1.04 E-07 9.36 E-08 8.50 E-08 Values are in s/m 3, extracted from Reference 7. l ( 8-10 Gen. Rev. 13

FNP-0-M-011 () Table 8-3. Annual Average (i7D) for Ground-Level Releases Distance to Location, in miles 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 7.25 E-05 2.38 E-05 8.63 E-06 4.02 E-06 2.05 E-06 NNE 6.16 E-05 2.02 E-05 7.32 E-06 3.39 E-06 1.73 E-06 NE 5.86 E-05 1.94 E-05 7.04 E-06 3.24 E-06 1.65 E-06 ENE 5.27 E-05 1.74 E-05 6.32 E-06 2.92 E-06 1.49 E-06 E 6.28 E-05 2.02 E-05 7.27 E-06 3.40 E-06 1.75 E-06 ESE 6.18 E-05 1.97 E-05 7.09 E-06 3.33 E-06 1.72 E-06 SE 9.48 E-05 3.01 E-05 1.07 E-05 5.06 E-06 2.63 E-06 SSE 1.44 E-04 4.55 E-05 1.61 E-05 7.65 E-06 3.99 E-06 S 1.55 E-04 4.87 E-05 1.72 E-05 8.20 E-06 4.23 E-06 SSW 9.78 E-05 3.12 E-05 1.11 E-05 5.23 E-06 2.71 E-06 SW 7.40 E-05 2.40 E-05 8.74 E-06 4.05 E-06 2.07 E-06 WSW 6.01 E-05 1.97 E-05 7.18 E-06 3.31 E-06 1.68 E-06 W 5.76 E-05 1.88 E-05 6.79 E-06 3.14 E-06 1.60 E-06 WNW 5.55 E-05 1.82 E-05 6.55 E-06 3.03 E-06 1.55 E-06 NW 5.67 E-05 1.86 E-05 6.76 E-06 3.14 E-06 1.60 E-06 NNW 6.60 E-05 2.16 E-05 7.85 E-06 3.65 E-06 1.87 E-06 /\ Distance to Location, in miles

--                    2.5-2.99       3.0-3.49        3.5-3.99    4.0-4.49    4.5-4.99 N        1.19  E-06     8.24  E-07      6.09 E-07   5.35  E-07  4.71  E-07 NNE        1.00  E-06     6.94  E-07      5.13 E-07   4.50  E-07  3.96  E-07 NE        9.47  E-07     6.54  E-07      4.82 E-07   4.23  E-07  3.71  E-07 ENE        8.56  E-07     5.92  E-07      4.37 E-07   3.82  E-07  3.37  E-07 E        1.02  E-06     7.08  E-07      5.24  E-07  4.61  E-07  4.06  E-07 ESE        1.02  E-06     6.99  E-07      5.18  E-07  4.56  E-07  4.02  E-07 SE        1.54  E-06     1.07  E-06      7.99  E-07  7.04  E-07  6.20  E-07 SSE        2.34  E-06     1.64  E-06      1.22  E-06  1.08  E-06  9.49  E-07 5        2.51  E-06     1.76  E-06      1.31  E-06  1.16  E-06  1.02  E-06 SSW        1.58  E-06     1.10  E-06      8.17  E-07  7.19  E-07  6.33  E-07 I

SW 1.20 E-06 8.30 E-07 6.12 E-07 5.38 E-07 4.73 E-07 WSh' 9.65 E-07 6.67 E-07 4.91 E-07 4.31 E-07 3.79 E-07 W 9.20 E-07 6.37 E-07 4.71 E-07 4.13 E-07 3.63 E-07 WNW 8.92 E-07 6.18 E-07 4.56 E-07 4.01 E-07 3.52 E-07 { NW 9.25 E-07 6.41 E-07 4.73 E-07 4.16 E-07 3.65 E-07  ! NNW 1.10 E-06 7.50 E-07 5.54 E-07 4.87 E-07 4.28 E-07 l Values are in s/m 3 , extracted from Reference 7. v' 8-11 Gen. Rev. 13

1 FNP-0-M-011 1 f- s Table 8-4. Annual Average (D7Q) for Mixed Mode Releases l Distance to Location, in miles

                 *      #                                                                         I 0.25-0.5        0.5-0.99        1.0-1.49    1.5-1.99    2.0-2.49 l

N 3.82 E-08 1.78 E-08 7.53 E-09 3.39 E-09 1.62 E-09 NNE 4.57 E-08 2.08 E-08 8.69 E-09 3.88 E-09 1.85 E-09 NE 4.78 E-08 2.20 E-08 9.08 E-09 4.03 E-09 1.92 E-09 ENE 2.67 E-08 1.32 E-08 5.63 E-09 2.54 E-09 1.22 E-09 E 2.87 E-08 1.40 E-08 5.77 E-09 2.55 E-09 1.22 E-09 ESE 3.29 E-08 1.53 E-08 6.17 E-09 2.70 E-09 1.28 E-09 SE 5.30 E-08 2.37 E-08 9.31 E-09 4.01 E-09 1.90 E-09 SSE 5.07 E-08 2.35 E-08 9.53 E-09 4.19 E-09 1.99 E-09 S 4.86 E-08 2.29 E-08 9.16 E-09 4.00 E-09 1.90 E-09 l SSW 4.29 E-08 2.10 E-08 9.09 E-09 3.97 E-09 1.88 E-09 l SW 4.70 E-08 2.28 E 08 1.05 E-08 4.39 E-09 2.04 E-09 WSW 4.46 E-08 2.17 E-08 9.88 E-09 4.12 E-09 1.92 E-09 W 3.96 E-08 1.94 E-08 8.39 E-09 3.63 E-09 1.70 E-09 WNW 3.22 E-08 1.56 E-08 6.35 E-09 2.85 E-09 1.37 E-09 NW 2.83 E-08 1.35 E-08 5.55 E-09 2.46 E-09 1.18 E-09 NNW 3.24 E-08 1,55 E-08 6.59 E-09 2.97 E-09 1.42 E-09 Distance to Location, in miles ("% Sector (} N 2.5-2.99 8.71 E-10 3.0-3.49 5.64 E-10 3.5-3.99 3.10 E-10 4.0-4.49 3.37 E-10 4.5-4.99 2.91 E-10 NNE 9.91 E-10 6.43 E-10 4.44 E-10 3.82 E-10 3.30 E-10 NE 1.03 E-09 6.65 E-10 4.62 E-10 3.98 E-10 3.43 E-10 ENE 6.57 E-10 4.22 E 2.96 E-10 2.55 E-10 2.20 E-10 E 6.57 E-10 4.20 E-10 2.96 E-10 2.55 E-10 2.20 E-10 ESE 6.88 E-10 4.40 E-10 3.09 E-10 2.66 E-10 2.29 E-10 SE 1.01 E-09 6.48 E-10 4.55 E-10 3.90 E-10 3.36 E-10 SSE 1.07 E-09 6.85 E-10 4.79 E-10 4.12 E-10 3.55 E-10 S 1.02 E-09 6.49 E-10 4.59 E-10 3.94 E-10 3.40 E-10 SSW 1.00 E-09 6.41 E-10 4.50 E-10 3.86 E-10 3.32 E-10 SW 1.08 E-09 6.90 E-10 4.81 E-10 4.12 E-10 3.53 E-10 WSW 1.02 E-09 6.51 E-10 4.53 E-10 3.87 E-10 3.32 E-10 W 9.00 E-10 5.92 E-10 4.13 E-10 3.54 E-10 3.04 E-10 WNW 7.33 E-10 4.95 E-10 3.52 E-10 3.05 E-10 2.65 E-10 NW 6.37 E-10 4.11 E-10 2.91 E-10 2.50 E-10 2.14 E-10 NNW 7.66 E-10 4.95 E-10 3.45 E-10 2.97 E-10 2.56 E-10 Values are in m*2, extracted from Reference 7. O 8-12 Gen. Rev. 13

                  ~

. . l l I FNP-0-H-011

    -s  Table 8-5. Annual Average (D/Q) for Oround-Level Releases                            1 l
 \,_)

Distance to Location, in miles 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 2.50 E-07 7.84 E-08 2.53 E-08 9.61 E-09 4.28 E-09 NNE 2.48 E-07 7.77 E-08 2.51 E-08 9.53 E-09 4.24 E-09 NE 2.49 E-07 7.80 E-08 2.52 E-08 9.57 E-09 4.26 E-09 ENE 1.69 E-07 5.29 E-08 1.71 E-08 6.48 E-09 2.88 E-09 E 1.69 E-07 5.28 E-08 1.71 E-08 6.48 E-09 2.88 E-09 ESE 1.80 E-07 5.54 E-08 1.79 E-08 6.80 E-09 3.02 E-09 SE 2.75 E-07 8.63 E-08 2.79 E-08 i 1.06 E-08 4.71 E-09 l SSE 3.66 E-07 1.15 E-07 3.71 E-08 1.41 E-08 6.25 E-09 S 3.70 E-07 1.16 E-07 3.75 E-08 1.42 E-08 6.33 E-09 SSW 2.75 E-07 8.62 E-08 2.79 E-08 1.06 E-08 4.70 E-09 SW 2.60 E-07 8.15 E-08 2.64 E-08 1.00 E-08 4.45 E-09 WSW 2.31 E-07 7.24 E-08 2.34 E-08 8.88 E-09 3.95 E-09 I W 2.11 E-07 6.61 E-08 2.14 E-08 8.11 E-09 3.61 E-09 j WNW 1.83 E-07 5.73 E-08 1.85 E-08 7.02 E-09 3.12 E-09 NW 1.74 E-07 5.45 E-08 .1.76 E-08 6.68 E-09 2.97 E-09 NNW 2.13 E-07 6.67 E-08 2.16 E-08 8.19 E-09 3.64 E-09 Distance to Location, in miles 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 (' )\ N 2.22 E-09 1.45 E-09 9.79 E-10 8.27 E-10 6.99 E-10 ' NNE 2.20 E-09 1.43 E-09 9.71 E-10 8.20 E-10 6.93 E-10 NE 2.21 E-09 1.44 E-09 9.75 E-10 8.23 E-10 6.96 E-10 ENE 1.50 E-09 9.76 E-10 6.60 E-10 5.58 E-10 4.72 E-10 E 1.50 E-09 9.75 E-10 6.60 E-10 5.57 E-10 4.71 E-10 ESE 1.57 E-09 1.02 E-09 6.72 E-10 5.85 E 10 4.94 E-10 SE 2.44 E-09 1.59 E-09 1.08 E-10 9.11 E-10 7.70 E-10 SSE 3.25 E-09 2.12 E-09 1.43 E-10 1.21 E-10 1.02 E-10 S 3.29 E-09 2.14 E-09 1.45 E-10 1.22 E-09 1.04 E-10 SSW 2.44 E-09 1.59 E-09 1.08 E-10 9.10 E-10 7.69 E-10 SW 2.31 E-09 1.51 E-09 1.02 E-10 8.60 E-10 7.27 E-10 WSW 2.05 E-09 1.34 E-09 9.04 E-10 7.64 E-10 6.46 E-10 W 1.87 E-09 1.22 E-09 8.25 E-10 6.97 E-10 5.90 E-10 WNW 1.62 E-09 1.06 E-09 7.15 E-10 6.04 E-10 5.11 E-10 NW 1.54 E-09 1.01 E-09 6.80 E-10 5.75 E-10 4.86 E-10 NNW 1.89 E-09 1.23 E-09 8.34 E-10 7.04 E-10 5.95 E-10 values are in m-2, extracted from Reference 7. O 8-13 Gen. Rev. 13

  , . . . . . . . .   . - . . . . . - . . . . ~ . . . . - ~ . . . . . ~ - . - - . - . . - - - .                                                               . _ . . . - . . . . . . ~ . . . . . . . - . . - . . . . ~ . . - . . - . . .

i I FMP-0-M-011 l

teco . . .

I W A' ( I J J' P i / )~ 4 , / / / . l / / -- l l / ,f ,,* f ! / / / , , , -' ! ( / / /

                                                                                                                                                                                                              ~~

4 , / ' p / 100

                                                                                                         .'         l            /                '
                                                                                                                                                                      /                                            ~

i , a-4 ,' , , ,,,,,-- l 1 r x 1 e s- , t .r . / / / s l ) / 7 / / .., , - --- i I ,/ / / c// / / l fff /kf,/ ' , / l r

                                                                        / /
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                                                                  '* 1' / / / / / Of 1
                                                                                           ,             ,          y s s                     s           1

! / / /

                                                                                /              /

i f / f l i l

                                                                         /

l l ' o.i 1.o to too { PLuist TRAVEL DISTANCE (KILonstTEIMI q l  ! l Range of Vertical Range of Vertical ! category Temperature Gradient Temperature Oradient t ('C/100 m) ('F/100 ft) A AT/AE < -1.9 AT/At < -1.0 I i B -1.9 s AT/AE < -1.7 -1.0 s AT/As < -0.9 l c -1.7 s AT/At < -1.5 -0.9 s AT/&E e -0.8r 1 D -1. 5 s AT/A E < -0. 5 -0.8 s AT/AE < -0.3 ! E -0.5 s AT/AE < 1.5 -0.3 s AT/AE < 0.8 I F 1.5 s AT/As < 4.0 0.8 s AT/As < 2.2 i G 4.0 5 4T/45 2.2 s AT/AE i i l

}                                                           This graph is reproduced from Reference 5 (Figure 1).

{. Figure 8-1. Vertical Standard Deviation of Material La a Flume (og) i-J

 !                                                                                                                                   3-14                                                                                     Gen. Rev. 13 i

   . . . . _ . .-. _ . . . .                  ~ . _ . . . - - . - - . . - - . . . _ . . . . . . - . . . . - ~                         _ . - . . _ _ . . .     . - . - . . . _ _ . _ . . . - . .

1 4 1 1 1 4 f 4 FNP-0-M-011 l l 1 l  : I a 1 i ' ? l 1 b 4 e l i i I 4 , to 1 4 1 1 4 e i

                                          .                          i         ,

l 6 ' 6i j , I il ' l ll \ j w i -j E

                             >                                                                                \ ,

l M N I  %% 3 * **- j 1.0 i . < , i , , , , U 1( i i i 6 ' i t ie .l 1

                                           .                                i             l l

! I i l I ! l ll i l t i 4 < l l l l ( 0.1 .- l 0.1 1A to too r DesTAascs lufteneg7tas! I 1 I t i i i I s J l i ' i 's , 1 1 l 1 a 1 4 I\ This graph is reproduced from Reference 4. l J Figure 8-2. Terrain Recirculation Factor (h) 1 l l 8-15 Gen. Rev. 13  ;

i t

-                                                                                                                                        1 i(                                                                                                        FNP-O-M-Oli i

l t i 1 't I i k E

                     es                                                                                                                \

i s.s

sA i

l s.1 i i s.1 1A tea NBA 200A 4 PLuut TnAvst peerAsoca tu LoewTeam i i* 4 i i !O This graph is reproduced front Reference 5 (Figure 4). Figure 4-5. Plume Depletion Effect for 60-Meter Releases h=18 Gen. Rev. 13

_.._..__..._.___...___...._._.__.___..._._._.____..__.__-,m..._..__ FNP-0-M-011 l is

                                                                               --%         N
                                                                                                  .       =,     _meutnALm)
                                                                                                     '%   =,      %

N N_ l,, unsTAsta N e, um \.' s ' E STABLEIE,PA

                                                                                                                            \'     h es -
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                                                                                                                                       \

I d, tenActuinnsAAmnee im 'N N 88 , 1 l i

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eA L eJ

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u e.1 i

s.1 1A tea toes ases l PLussa YnAvstouTAmes ta:LoessTana j l l i i

                                                                                                                                                 . l I

This graph is ..y../. wood from Reference 5 (Figure 5). Figure S-6. Plane Depletion Effect for 100-Meter Releases 8-19 Gen. Rev. 13

1 I l 1 1 FNP-0-M-011 l 4 4 4 1 I 10-3 1 s . l l 4 4 l'

                                         \     %

4 e '

                                                     ~                                                                                                                                                l

! E N I N l \ 3' l r N i E  % 1 e f- E 104 A e c X s i T 1-X ! 't i E ! E X

                                                                                                                                                                               %6 w                                                                                                                                                      1 n

1 m 104 - I s i e i i d ! 10-7 1 i 0.1 1A 10.0 100.0 300A i j PLURSE TRAVEL DISTANCE (KILO 94ETEMS) i 4 4 4 i l This graph is reproduced from Reference 5 (Figure 6). Figure S-7. Relative Deposition for Ground-Level Releases { 4 1 S-20 Gen. Rev. 13

i i i, !b FNP-0-M-011 !v i e 1 I U-r Ru u, e " N E I

                                                                                -t j,                                                       [                         'l

, x

g # h E / h NEUTRAL l 5 / NSu <

ri N f j le .

                                                                          / NEUTRAL $M

( ,, w

                                                                                                                                  .T 1

a

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• i O 2 J D u l u g

y i e vm

.- I r x x

2 s

                                                                /                                             /                       N g 0                                                                                                                                                %                    i j                                               g                                                              STA.LE (E F.Gl z                                                ,

I I I J r r r i , h b l J I I ! f f 10-7 l 0.1 tr WK 100A 3DOA i PLUME TRAVELDISTANCE(K8LOteETERS) i ~ i t i 1~ This graph is reproduced from Reference 5 (Figure 7). FTgure S-8. Relative Deposition for 30-Meter Releases I j 8-21 Gen. Rev. 13

,.._..__.m. - . _ . . . _ _ _ _ . _ _ . . _ . _ _ . _ _ _ . _ _ _ _ _ _ . . _ . _ _ _ . . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ . . . _ _ . _ _ _ _ . . _ _ _ . _ _ . _ . . _ . . _ _ _ l l' l l t FNP-0-M-011 i 104

                                                                                    #%                        UNSTABLE (A,B C)

[ \

                                                                       /                                            N 10-6                 '

j / f NEUTRAL  %(D) k "

                                                                                                                                                                 ;m
                                                          ,                                     /                                                                                                                         l g                         i                                    a                                                                     s l                             w                       I                                    I                                                                             w
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k a b. f E 1 ,, I j' E w

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k A h 5 [ i E i 10-7 STABLE (E,F,0) j , i I I I I 104 J 0.1 1A 18A- 100.0 200A PLUME TRAVEL DISTANCE (KILOMETERS) O This graph is reproduced from Reference 5 (Figure 8). Figure 3-9. Reittive Deposition for 60-Meter Releases 8-22 Gen. Rev. 13

 ,- .    . - . - . . - .    . _ . . . .        .      . . .           . - - .                . ~ . . . . . - . . . - - - - . . . _ - . . ~ - _ . . . . . . . ~                     ..

I 4 i i l a FNP-0-M-011 , j [ 4 i j ., 1 3 4 i l ! 104 ' UNSTABLE (AA,Cl i l [ 'w N l i

• N I 104  ; -- h' i -

                                                                                                                           'W'x
                                      =

s ,1 ,1 w g i f NEUTRAL (D) xs 4 f a / I N 'N ' j z w [ / N- s-l @ / / N N 4 A l 10 z n

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! I s r l J I I 5 I i E r 1

                                        <                                       l                                                                                                     i
                                                                              /
                                        .J STABLE (E,F,0) g NO DEPLETION f

j 10-7 - 1 8 l I j < I I I i f i I I I j 104 i 0.1 1A 10A 100A 200A PLUME TRAVEL DISTANCE (KILOMETERS) 1 1 1 4 k

 ,I i

i i j This graph is reproduced from Reference 5 (Figure 9). j Figure 8-10. Relative Deposition for 100-Meter (or Greater) Releases i S-23 Gen. Rev. 13 s

i 1 , FNP-O-M-011 ] CHAPTER 9 f' METHODS AND PhRAMETERS FOR CALCULATION OF

    \

GASEOUS EFFLUENT PATHWAY DOSE FACTORS, Rgp ; t d' l 9.1 INHALATION PATHWAY FACTOR l 1, 3 For the inhalation pathway, Raipj in (mrem /y) per (pCi/m ) is calculated as i t follows (Reference 1, Section 5.3.1.1): l Raip) = Kg * (BR)a (DTA)at) (9 1) ' where Kg = the units conversion factor: 106 petjpet, l (BR), = the breathing rate of receptor age group a, in m3 /y, from l l Table 9-5. ' 1

                                        =   the inhalation dose factor for receptor age group a,
    /)                     (DFA) gjj radionuclide i, and organ j, in mram/pci, from Table 9-7                    ,

through Table 9-10. 9-1 Gen. Rev. 13

y .c FNP-0-M-011 i 9.2 GROUND PLANE PATHWAY FACTOR "l

O For the ground plane external exposure pathway, P.,j 2

p j in (m mrom/y) per (pci/s) is calculated as follows (Reference 1, Section 5.3.1.2):

                                                                                              -* -Af                (9.2)

Rapjj = Kg

• K2 * (88F) *

(DFG)lj_ . w ~! , where p . q Kg = the units conversion factor: 100 pCi/pci. the units conversion factors 8760 h/y. K2= 2 (SHF) = the shielding factor due to structure gdimensionless). The value used for (SHF) is 0.7, from (Reference 3, Table E-15). 3 j (DFG)jj= the ground plane dose facter for radionuclide i and organ j, in (mesm/h) per -(pci/m 2), from Table 9-15. Dose factors are the same for all age groups, and those for the total body also apply to all organs other than' skin. 1; = - the radioactive decay constant for radionuclide 1, in s -I. Values of 1; used in effluent. calculations should be 3 based on decay data from a recognized and current source, such as Referenc's 15. l t= the exposure time, in s. The value used for t is , 4.73 x 10 8 s (= 15 y), from (Reference '1, Section

5. 3.1'. 2 ) .

t i i t 9-2 Gen. Rev. 13

FNP-0-M-011

  . (~)

V 9.3 GARDEN VEGETATION PATHKAY FACTOR For radionuclides other than tritium in the garden vegetation consumption pathway, R,;pj in (m2 mrem /y) per (yCi/s) is calculated as follows (Reference 1, Section 5.3.1.5): Rapj ) = Kg * * (DFL)ag; Y v(A+A) l w (9.3) l -h] IL l

                                         \Ual fL*

• UaS f g* "h lh' )'

where Kg = the units conversion factor: 106 peijpeg, r= the fraction of deposited activity retained on the edible parts of garden vegetation (dimensionless). The value used for r is 1.0 for radioiodines and 0.2 for

    '(fA)                         particulates, from (Reference 3, Table E-1).

Yy = the areal density (agricultural productivity) of growing leafy garden vegetation, in kg/m 2, from Table 9-1. A = the radioactive decay constant for radionuclide i, in s -I. Values of A used in effluent calculations should be g based on decay data from a recognized and current source, j such as Reference 15. Aw= the rate constant for removal of activity on leaf and plant surfaces by weathering, in s-I, front Table 9-1. (DFL)aij = the ingestion dose factor for receptor age group a, radionuclide 1, and organ j, in mrem /pci, from Table 9-11 through Table 9-14. , Ug = the consumption rate of fresh leafy garden vegetation by a receptor in age group a, in kg/y, tuom Table 9-5. I i 9-3 Gsn. Rev. 13

1 1 1 l FNP-O-M-011

;t                                          U,3 =

the consumption rate of stored garden vegetation by a receptor in age group a, in kg/y, from Table 9-5. 1 l

)                                           ft=       the frar'. ion of the annual intake of fresh leafy garden vegetation that is grown locally (dimensionless), from Table 9-1.

l f g= the fraction of the annual intake of stored garden vegetation that is grown locally (dimensionless), from Table 9-1. i f tg = the average time between harvest of fresh leafy garden 1 vegetation and its consumption, in s, from Table 9-1. j thy = the average time between harvest of stored garden vegetation and its consumption, in s, from Table 9-1.

 ; m                             For tritium in the garden vegetation consumption pathway, R,jg in (mrom/y)

( per (yci/m 3) is calculated as follows, (Reference 1, Section 5.3.1.5), based on the concentration in air rather than deposition onto the grounds } ma id - "I * "3 - (8)av ' ( "a 't + "as 's ) o 75 - ( ) (' *) l !' where l K3= the units conversion factor: 103 g/kg. l i I H= the absolute humidity of atmospheric air, in g/m3 , from Table 9-1. 0.75 = the fraction of the mass of total garden vegetation that is water (dimensionless). " O.5 = the ratio of the specific activity of tritium in garden vegetation water to that in atmospheric water (dimensionless). and other parameters are as defined above. 9-4 Gen. Rev. 13 1

i

 <                                                                                                                                         l i

I FNP-0-M-011 Table 9-1. Miscellaneous Parameters for the Garden Vegetation Pathway  ; 1 The following parameter values are for use in calculating R,;p; for the i garden vegetation pathway only. The terms themselves are defined in l Section 9.3. Parameter Value Reference i g Ty 2.0 kg/m2 Ref. 3, Table E-15 5

5. 73 x 10-7 ,-1 1, Ref. 1, p.7e 33 (14-day half-life) fg 1.0 nef. 1, page 36 f

g 0.76 Ref. 1, page 33 tg 8.6 x 104 s Ref. 3, Table E-15 4 (1 day) 1 thy 5.18 x 106 s Ref. 3, Table E-15 1 (60 days) H 8 g/m 3 Ref. "l 1 1 1 i i J 9-5 Gen. Rev. 13

_ - - - - . =_ . - - - - __ .. 1 l I l FNP-0-M-011  ; 9.4 GRASS-COW-MILK PATHWAY FACTOR I For radionuclides other than tritium in the grass-cow-milk pathway, Raipj in (m2

• mrem /y) per (yci/s) is calculated as follows (Reference 1, Section 1

5.3.1.3): 1 Rapj; = Kg *

• Op
• Uap *Emi * (DFL)ajj

                                                                        .x, gg,                  0=U fp i     (1 - pf sI)*
                                          .            s+                          . , ~Xi tj         <
                                                                                                        \

Yp Yr i where: i Kg = the units conversion factor: 100 pCi/yci. I r= the fraction of deposited activity retained on the edible , parts of vegetation (dimensionless). The value used for r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Table E-1). 1 = the radioactive decay constant for radionuclide i, in s -I. Values of 1; used in ef fluent calculations should be based on decay data from a recognized and current source, such as Reference 15. A., = the rate constant for removal of activity on leaf and plant surfaces by weathering, in s-I, from Table 9-2. Op - ther cow +s consumption rate of feed, in kg/d, from Table 9-2. U,p = the consumption rate of cow milk by a receptor in age group a, in L/y, from Table 9-5. F an = the stable element transfer coefficient applicable to radionuclide i, for cow's milk, in d/L, from Table 9-6. 9-6 Gen. Rev. 13

l FNP-0-M-011 b Q (DFL)aij = the ingestion dose factor for receptor age group a, 1 radionuclide i, and organ j, in mrem /pci, from Table 9-11 l through Table 9-14. f = the fraction of the year that the cow is on pasture p (dimensionless), from Table 9-2. i f, = the fraction of the cow *s feed that is pasture grass while the cow is on pasture (dimensionless), from l Table 9-2. Yp = the areal density (agricultural productivity) of growing pasture feed grass, in kg/m 2, from Table 9-2. Y, = the areal density (agricultural productivity) of growing stored feed, in kg/m 2, from Table 9-2. than = the transport time from harvest of stored feed to its consumption by the cow, in s, from Table 9-2. l tg = the transport time from consumption of feed by the cow, to consumption of milk by the receptor, in s, from Table 9-2. For tritium in the grass-cow-milk pathway, Rap j j in (mrem /y) per (pci/m3) is calculated as follows (Reference 1, Section 5.3.1. 5 ) , based on the concentration in air rather than deposition onto the ground O*U R alp) ~ Kg

• K3*Of Uap
• Emi * (DFL)aj;
• 0.15 * (9.6) where:

K3= the units conversion factor: 103 g/kg. H= the absolute humidity of atmospheric air, in g/m 3, fror Table- 9-2. V 9-7 Gen. Rev. l'3

FNP-0-M-011 0.75 = the fraction of the mass of total vegetation that is water (dimensionless). 0.5 = the ratio of the specific activity of tritium in vegetation water to that in atmospheric water (dimensionless), and other parameters are as defined above. l (~ l 1 l l [

 \

9-8 Gen. Rev. 13

1 f% FNP-0-M-011 ( ) Table 9-2. Miscellaneous Parameters for the Grass-Cow-Hilk Pathway 1 l l The following parameter values are for use in calculating Raipj for the grass-cow-milk pathway only. The terms themselves are defined in Section 9.4. Parameter Value Reference A,w 5. 7 3 x 10-7 .I Ref. 1, page 33 i (14-day half-life) i Op 50 kg/d Ref. 3, Table E-3 f 1.0 p Ref. 1, page 33 i f, 1.0 Ref. 1, page 33 s Y 0.7 kg/mI Ref. 3, Table E-15 p i Ys 2.0 kg/m2 Ref. 3, Table E-15 (

    \' /

tkm 7.78 x 100 s Ref. 3, Table E-15

• I (90 days) a tg 1.73 x 105 s Ref. 3, Table E-15 (2 days) 1 H 8 g/m3 Ref. 3 4

i a b 1 O 9-9 Gen. Rev. 13

1 1 l i l FNP-O-M-011 's . 9.5 GRASS-GOAT-MILK PATHWAY FACTOR For radionuclides other than tritium in the grass-goat-milk pathway, Rjj ap in (m2 +mrom/y) per (yci/s) is calculated as follows (Reference 1, Section i 5.3.1.3): -l 1 + i l Raip)

• Kg=
• Qy
• Uap *I mi * (DFL)aij 1

                                                                 -Af t/im

( } i fp f, + (1-f pf,) e - X; If 4 . . i- Yp Ys i i

i l where:

i Kg = the units conversion factor: 106 petypet, i r= the fraction of deposited activity retained on the ed! * .e parts of vegetation (dimensionless) . The value used for

r is 1.0 for radiciodines and 0.2 for particulates, from 4

(Reference 3, Table E-1). i li= the radioactive decay constant for radionuclide i, in s -I. Values of A used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15. i 1, = the rate constant for removal of activity on leaf and plant surfaces by weathering, in s-I, from Table 9-3. Op = the goat's consumption rate of feed, in kg/d, from Table.9-3. U,p = the consumption rate of goat milk by a receptor in age group a, in L/y, from Table 9-5. F:ni = the stable element transfer coefficient. applicable to radionuclide i, for goat's milk, in d/L, from Table 9-6. 9-10 Gen. Rev. 13

                                                                                          ~ . . . - - _ . _ .

c _ FNP-0-M-011 ( (DFL)aij = the ingestion dose factor for receptor age group a, radionuclide i, and organ j, in mrom/pci, from Table 9-11 through Table 9-14.

                     =

f p the fraction of the year that the goat is on pasture (dimensionless), from Table 9-3. f, = the fraction of the goat's feed that.is pasture grass while the goat is on pasture (dimensionless), from Table 9-3. Y p= the areal density (agricultural productivity) of growing i pasture feed grass, in kg/m 2, from Table 9-3. Y, = the areal density (agricultural productivity) of growing stored feed, in kg/m 2, from Table 9-3. l than = the transport time from harvest of stored feed to its consumption by the goat, in s, from Table 9-3. O tg = the transport time from consumption of feed by the goat, . to consumption of milk by the receptor, in s, from Table 9-3. For tritium in the grase-goat-silk pathway, R,;p; in (mrem /y) per (pci/m3) is calculated as follows (Reference 1, Section 5. 3.1. 5 ) , based on the concentration in air rather than deposition onto the grounds Ralp)

• Kg
• K3
• Oy Vap. *Eml * ( DFL ) al)
• O
• 15 * *

(***) I where K3= the units conversion factor: 103 g/kg. H= the absolute humidity of atmospheric air, in g/m 3 , from Table 9-3. A 9-11 Gen. Rev. 13

  ...__.-.___._...___._.___.-_.______-_.__.__.-.______.._.._._.m   .

j- ) i 4

FNP-0-M-011 ,

1

0.75 = the fraction of the mass of total vegetation that is f

} water (dimensionless). ' I l 0.5 = the ratio of the specific activity of tritium in l vegetation water to that in atmospheric water , i (dimensionless). J i and other parameters are as defined above. i i i i i 9 i i ) i iO i i ! i' t

?                                                                                                                                  .

l l i 1 ( . 5 2 J 9-12 Gen. Rev. 13

FNP-0-H-011 [3 Tcble 9-3. Miscellaneous Parameters for the Grass-Goat-Milk Pathway

  \v)

The following parameter values are for use in calculating Rapj j for the ] grass-goat-milk pathway only. The terms themselves are defined in i Section 9.5. l Parameter Value Reference  ! 1, 5.73 x 10*7 s *I Ref. 1, page 33 (14-day half-life) l 4 Qp 6 kg/d Ref. 3, Table E-3 f p 1.0 Ref. 1, page 33 f, 1.0 Ref. 1, page 33 Yp 0.7 kg/m2 Ref. 3, Table E-15 Y, 2.0 kg/m2 Ref. 3, Table E-15 (m tg 7.78 x 106 s Ref. 3, Table E-15 ) (90 days) tg 1.73 x 105 s Ref. 3, Table E-15 (2 days) H 8 g/m3 Ref. 3 On V amu 9-13 Gen. Rev. 23

      ._,  _ _          , - _         _ _ _ _ _ .             __ .__ _ _ ~ _ _ _                                _ _      _ . _ , _ _

i j FNP-0-M-011 9.6 GRASS-COW-MEAT PATHWAY FACTOR For radionuclides other than tritium in the grass-cow-meat pathway, Rjj ap in (m2* mrem /y) per (pci/s) is calculated as follows (Reference 1, Section 5.3.1.4): R alp) " # Kg * ,

• Oy
• Uap *F 9 * (DFL)aij (Af + Ag)

                                                                                               ~

fp f, (1 - fp f,) e

                                                       ,                                               ,   -h; Ij-Yp                           Is where Kg =           the units conversion factor:                         106 petypet, r=            the fraction of deposited activity retained on the edible
   /]

V parts of vegetation (dimensionless). The value used for r is 1.0 for radioiodines and 0.2 for particulates, from (Reference 3, Table E-1). A = the radioactive decay constant for radionuclide 1, in s -I. Values of A used in of fluent calculations should be based on decay data from a recognized and current source, such as Reference 15. Aw= the rate constant for removal of activity on leaf and plant surfaces by weathering, in s-I, from Table 9-4. Op = the cow's consumption rate of feed, in kg/d, from Table 9-4. U,p = the consumption rate of meat by a receptor in age group a, in kg/y, from Table 9-5. Fg = the stable element transfer coefficient applicable to radionuclide i, for meat, in d/kg, from Table 9-6. 9-14 Gen. Rev. 13

m- _ FNP-0-M-011 (DFL)jj = the ingestion dose f actor for receptor age group a,

                             . radionuclide 1, and organ j, in mrem /pci, from Table 9-11 through Table 9-14.

f = p the fraction of the year that the cow is on pasture (dimensionless), from Table 9-4. f, = the fraction of-the cow's feed that is pasture grass while the cow is on pasture (dimensionless), from Table 9-4. Y p= the areal density (agricultural productivity) of growing l pasture feed grass, in kg/m 2, from Table 9-4. Y, = the areal density (agricultural productivity) of growing stored feed, in kg/m 2, from Table 9-4. thm = the transport time from harvest of stored feed to its consumption by the cow, in s, from Table 9-4. tg = the transport time from consumption of feed by the cow, to consumption of meat by the receptor, in s, from j Table 9-4. For tritium in the grase-cow-meat pathway, Raipj in (mrom/y) per (pC1/m3 ) is calculated as .follows (Reference 1, Section 5.3.1.4) , based on the concentration in air rathet than deposition onto the ground: O*U Rgp; = Kg

• K3
• Qy
• Uap
• Ep
• LDTL)g;
• O.15 = (9.10)

                                                                       >Ni where K3=         the units conversion factor:     103 g/kg.

H= the absolute humidity of atmospheric air, in g/m 3 , from Table 9-4. - f w 9-15 Gen. Rev. 13

      ._.-._._.m

[~_. i l i FNP-0-M-011 t { O.75 = the fraction of the mass of total vegetation that is  ; water (dimensionless). i 0.5 = the ratio of the specific activity of tritium in vegetation watwr to that in atmospheric water l (dimensionless). l i a and other parameters are as defined above. 1 i l V . 1 9-16 Gen. Rev. 13

                         - . _ . .. . - . . . .             _   ._   _ - _ . ._.       .    - . _ . . _ -       - . ~ . . .

t i FNP-0-M-011 O Table 9-4. (_/ Miscellaneous Parameters for the Grass-Cow-Meat Pathway The following parameter values are for use in calculating R,jj p for the grass-cow-meat pathway only. The terms themselves are defined in Section l 9.6.  ! i I l Parameter Value Reference 1, 5.73 x 10'7 s *I Ref. 1, page 33

(14-day half-life) s Qp 50 kg/d Ref. 3, Table E-3 f

p 1.0 Ref. 1, page 33 i f, 1.0 Ref. 1, page 33 I Yp O.7 kg/m2 Ref. 3, Table E-15 Y, 2.0 kg/m2 Ref. 3, Table E-15 O tg 7.78 x 106 s Ref. 3, Table E-15 . (90 days) tg 1.73 x 106 s Ref. 3, Table E-15 l (20 days) H 8 g/m3 Ref. 3 1 9-17 Gen. Rev. 13

l FNP-0-M-011 Table 9-5. Individual Usage Factors s Receptor Age Group Usage Factor

Infant Child Teenager i Adult Milk Consumption Rate, U,P 330 330 400 310 (L/y)

Meat Consumption Rate, U aP O 41 65 110 (kg/y) Freuh Leafy Garden Vegetation Consumption Rate, Uat 0 26 42 64 (kg/y) Stored Garden Vegetation

  )             Consumption Rate, Ug   g                  0                     520    630                    520 (kg/y) 1 j,            Breathing Rate, (BR),

1400 3700 8000 8000 (m3 /y) f-

 >(\~-
 )                                                                                                                      .
.i All values are from Reference 3, Table E-5.

N 9-18 Gen. Rev. 13

                                                                               ~     ~ ~ ~ ~ ~ ' - -'

rt FNP-0-M-011 Tabid 9-6. Stable Element Transfer Data O's 1 Cow Milk Goat Milk Meat Element

• Fm (d/L) Fm (d/L)+ Fg (d/kg)*

l 3 H 1.0 E-02 1.7 E-01 1.2 E-02  ! C 1.2 E-02 1.0 E-01 3.1 E-02 i Na 4.0 E-02 4.0 E-02 3.0 E-02 P 2.5 E-02 2.5 E-01 4.6 E-02 Cr 2.2 E-03 2.2 E-03 2.4 E-03 Mn 2.5 E-04 2.5 E-04 8.0 E-04 { Fe 1.2 E-03 1.3 E-04 4.0 E-02 Co 1.0 E-03 1.0 E-03 1.3 E-02 , Ni 6.7 E-03 6.7 E-03 5.3 E-02 Cu 1.4 E-02 1.3 E-02 8.0 E-03 Zn 3.9 E-02 3.9 E-02 3.0 E-02 Br 5.0 E-02 5.0 E-02 2.6 E-02 Rb 3.0 E-02 3.0 E-02 3.1 E-02 Sr 8.0 E-04 1.4 E-02 6.0 E-04 l Y 1.0 E-05 1.0 E-05 4.6 E-03 Zr 5.0 E-06 5.0 E-06 3.4 E-02 I Nb 2.5 E-03 2.5 E-03 2.8 E-01 Mo 7.5 E-03 7.5 E-03 8.0 E-03 . Tc 2.5 E-02 2.5 E-02 4.0 E-01 ' Ru 1.0 E-06 1.0 E-06 4.0 E-01 Rh 1.0 E-02 1.0 E-02 1.5 E-03 Ag 5.0 E-02 5.0 E-02 1.7 E-02 i sb 1.5 E-03 1.5 E-03 4.0 E-03

Te 1.0 E-03 1.0 E-03 7.7 E-02 J

I 6.0 E-03 6.0 E-02 2.9 E-03 Cs 1.2 E-02 3.0 E-01 4.0 E-03 Ba 4.0 E-04 4.0 E-04 3.2 E-03 La 5.0 E-06 5.0 E-06 2.0 E-04 4 Ce 1.0 E-04 1.0 E-04 1.3 E-03 Pr 5.0 E-06 5.0 E-06 4.7 E-03

Nd 5.0 E-06 5.0 E-06 3.3 E-03 W 5.0 E-04 5.0 E-04 1.3 E-03 Np S~. 0 E-06' 5.0 E-06 2.0 E-04

• Values from Reference 3 (Table E-1) except as follows:

Reference 2 (Table C-5) for Br and Sb.

                    +    Values from Reference 3, Table E-2 for H, C,       P,   Fe, Cu, Sr, L and, CA i.n. goat allk amL Table. E-1 f.oc a,LL.

other elements in cow milk, except as follows:

   'T                    Reference 2 (Table C-5) for Br and Sb in cow milk.

9-19 Gen. Rev. 13 1

FNP-0-M-011 ' O Table 9-7. Inhalation Dose Factors for tne Infant Age Group D Nuclide Bone Liver T.8ody Thyroid Kidney Lung GI-LLI H-3 No Data 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 C-14 1.89E-05 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 Na-24 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 P-32 1.45E-03 8.03E-05 5.53E-05 No Data No Data No Data 1.15E-05 Cr-51 No Data No Data 6.39E-08 4.11E-08 9.45E-09 9.17E-06 2.55E-07 Mn-54 No Data 1.81E-05 3.56E-06 No Data 3.56E-06 7.14E-04 5.04E-06 Mn-56 No Data 1.10E-09 1.58E-10 No Data 7.86E-10 8.95E-06 5.12E-05 Fe-55 1.41E-05 8.39E-06 2.38E-06 No Data No Data 6.21E-05 7.82E-07 Fe-59 9.69E-06 1.68E-05 6.77E-06 No Data No Data 7.25E-04 1.77E-05 Co-58 No Data 8.71E-07 1.30E-06 No Data 5.55E-04 No Data 7.95E-06 Co-60 No Data 5.73E-06 8.41E-06 No Data No Data 3.22E-03 2.28E-05 Ni-63 2.42E-04 1.46E-05 8.29E-06 No Data No Data 1.49E-04 1.73E-06 p i Ni-65 1.71E-09 2.03E-10 8.79E-11 No Data No Data 5.80E-06 3.532-C5 t V Cu-64 No Data 1.34E-09 5.53E-10 No Data 2.84E-09 6.64E-06 l 1.07E-05 ' Zn-65 1.38E-05 4.47E-05 2.22E-05 No Data 2.32E-05 4.62E-04 3.67E-05 4 Zn-69 3.85E-11 6.91E-11 5.13E-12 No Data 2.87E-11 1.05E-06 9.44E-06 Br-83 No Data No Data 2.72E-07 No Data No Data No Data No Data Br-84 No Data No Data 2.86E-07 No Data No Data No Data No Data I Br-85 No Data No Data 1.46E-08 No Data No Data No Data No Data Rb-86 No Data 1.36E-04 6.30E-05 No Data No Data b Dea 2.17E-06 Rb-88 No Data 3.98E-07 2.05E-07 No Data No Data No Data 2.42E-07 Rb-89 No Data 2.29E-07 1.47E-07 No Data No Data No Data 4.87E-08 Sr-89 2.84E-04 No Data 8.15E-06 No Data No Data 1.45E-03 4.57E-05 Sr-90 2.92E-02 No Data 1.85E-03 No Data No Data 8.032-03 9.36E-05 Sr-91 6.83E-08 No Data 2.47E-09 No Data No Data 3.76E-05 5.24E-05 All values ars- in- (mrem /pCL inhale 11L.- They are obtained from-Reference 3 (Table E-10). Neither Reference 2 nor Reference 3 contains data for Rh-105, sb-124, or Sb-125. 9-20 Gen. Rev. 13

l FNP-0-M-011 d Table 9-7 (contd). Inhalation Dose Factors for the Infant Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 7.50E-09 No Data 2.79E-10 No Data No Data 1.70E-05 1.00E-04 Y-90 2.35E-06 No Data 6.30E-08 No Data No Data 1.92E-04 7.43E-05 Y-91m 2.91E-10 No Data 9.90E-12 No Data No Data 1.99E-06 1.68E-06 Y-91 4.20E-04 No Data 1.12E-05 No Data No Data 1.75E-03 5.02E-05 Y-92 1.17E-08 No Data 3.29E-10 No Data No Data 1.75E-05 9.04E-05 Y-93 1.07E-07 No Data 2.91E-09 No Data No Data 5.46E-05 1.19E-04 Zr-95 8.24E-05 1.99E-05 1.45E-05 No Data 2.22E-05 1.25E-03 1.55E-05 Zr-97 1.07E-07 1.83E-08 8.36E-09 No Data 1.85E-08 7.88E-05 1.00E-04 4 Nb-95 1.12E-05 4.59E-06 2.70E-06 No Data 3.37E-06 3.42E-04 9.05E-06 Mo-99 No Data 1.18E-07 2.31E-08 No Data 1.89E-07 9.63E-05 3.48E-05 Tc-99m 9.98E-13 2.06E-12 2.66E-11 No Data 2.22E-11 5.79E-07 1.45E-06 i Tc-101 4.65E-14 5.88E-14 5.80E-13 No Data 6.99E-13 4.17E-07 6.03E-07 ! Ru-103 1.44E-06 No Data 4.85E-07 No Data 3.03E-06 3.94E-04 1.15E-05 Ru-105 8.74E-10 No Data 2.93E-10 No Data 6.42E-10 1.12E-05 3.46E-05 Ru-106 6.20E-05 No Data 7.77E-06 No Data 7.61E-05 8.26E-03 1.17E-04 1 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 7.13E-06 5.16E-06 3.57E-06 No Data 7.80E-06 2.62E-03 2.36E-05 . Sb-124 No Data No Data No Data No Data No Data No Data No Data 4 Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 3.40E-06 1.42E-06 4.70E-07 1.16E-06 No Data 3.19E-04 9.22E-06 Te-127m 1.19E-05 4.93E-06 1.48E-06 3.48E-06 2.68E-05 9.37E-04 1.95E-05 Te-127 1.59E-09 6,81E-10 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-05 Te-129m 1.01E-05 4.35E-06 1.59E-06 3.91E-06 2.27E-05 1.20E-03 4.93E-05 4 Te-129 5.63E-11 2.48E-11 1.34E-11 4.82E-11 1.25E-10 2.14E-06 1.88E-05 Te-131m 7.62E-08 3.93E-08 2.59E-08 6.38E-08 1.89E-07 1.42E-04 8.51E-05 Te-131 1.24E-11 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.47E-06 5.87E-06 9-21 Gen. Rev. 12 4

                        -. ~            . .   -.       ._   .         -    -  -

J J FNP-0-M-011 !/G Tcble 9-7 (contd). Inhalation Dose m ators for the Infant Age Group

  )

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI 4 Te-132 2.66E-07 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 I-130 4.54E-06 9.91E-06 3.98E-06 1.14E-03 1.09E-05 No Data 1.42E-06 I-131 2.71E-05 3.17E-05 1.40E-05 1.06E-02 3.70E-05 No Data 7.56E-07 I-132 1.21E-06 2.53E-06 8.99E-07 1.21E-04 2.82E-06 No Data 1.36E-06 I-133 9.46E-06 1.37E-05 4.00E-06 2.54E-03 1.60E-05 No Data 1.54E-06 I-134 6.58E-07 1.34E-06 4.75E-07 3.18E-05 1.49E-06 No Data 9.21E-07 4 I-135 2.76E-06 5.43E-06 1.98E-06 4.97E-04 6.05E-06 No Data 1.31E-06 1 i Cs-134 2.83E-04 5.02E-04 5.32E-05 No Data 1.36E-04 5.69E-05 9.53E-07 Cs-136 3.45E-05 9.61E-05 3.78E-05 No Data 4.03E-05 8.40E-06 1.022-06 Cs-137 3.92E-04 4.37E-04 3.25E-05 No Data 1.23E-04 5.09E-05 9.53E-07 Cs-138 3.61E-07 5.58E-07 2.84E-07 No Date 2 93E-07 4.67E-08 6.26E-07 Ba-139 1.06E-09 7.03E-13 3.07E-11 No Data 4.23E-13 4.25E-06 3.64E-05 Ba-140 4.00E-05 4.00E-08 2.07E"06 No Data 9.59E-09 1.14E-03 2.74E-05 *

Ba-141 1.12E-10 7.70E-14 3.55E-12 No Data 4.64E-14 2.12E-06 3.39E-06

] Ba-142 2.84E-11 2.36E-14 1.40E 12 No Data 1.36E-14 1.11E-06 4.95E-07 La-140 3.61E-07 1.43E-07 3.68E-08 No Data No Data 1.20E-04 6.06E-05 La-142 7.36E-10 2.69E-10 6.46E-11 No Data No Data 5.87E-06 4.2SE-05 Co-141 1.98E-05 1.19E-05 1.42E-06 No Data 3.75E-06 3.69E-04 1.54E-05 Ce-143 2.09E-07 1.38E-07 1.58E-08 No Data 4.03E-08 8.30E-05 3.55E-05 Co-144 2.28E-03 8.65E-04 1.26E-04 No Data 3.84E-04 7.03E-03 1.06E-04 Pr-143 1.00E-05 3.74E-06 4.99E-07 No Data 1.41E-06 3.09E-04 2.66E-05 Pr-144 3.42E-11 1.32E-11 1.72E-12 No Data 4.80E-12 1.15E-06 3.06E-06 l 1 Nd-147 5.67E-06 5.81E-06 3.57E-07 No Data 2.25E-06 2.30E-04 2.23E-05 W-187 9.26E-09 6.44E-09 2.23E-09 No Data No Data 2.83E-05 2.54E-05 Np-239 2.65E-07 2.37E-08 1.34E-08 No Data 4.73E-08 4.25E-05 1.78E-05 0 9-22 Gan. Rev. 13

t FNP-0-M-011

' r" j(          Tcble 9-8.      Inhalation Dose Factors for the Child Age Group Nuclide      Bone       Liver     T. Body  Thyroid   Kidney      Lung     GI-LLI H-3      No Data    3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07         3.04E-07 C-14    9.70E-06 1.82E-06      1.82E-06 1.82E-06 1.82E-06 1.82E-06       1.82E-06 Na-24    4.35E-06 4.35E-06 4.35E-06          4.35E-06 4.35E-06 4.35E-06 4.35E-06

P-32 7.04E-04 3.09E-05 2.67E-05 No Data No Data No Data 1.14E-05 Cr-51 No Data No Data 4.17E-08 2.31E-08 6.57E-09 4.59E-06 2.93E-07 Mn-54 No Data 1.16E-05 2.57E-06 No Data 2.71E-06 4.26E-04 6.19E-06 Mn-56 No Data 4.48E-10 8.43E-11 No Data 4.52E-10 3.55E-06 3.33E-05

Fe-55 1.28E-05 6.80E-06 2.10E-06 No Data No Data 3.00E-05 7.75E-07 Fe-59 5.59E-06 9.04E-06 4.51E-06 No Data No Data 3.43E-04 1.91E-05 Co-58 No Data 4.79E-07 8.55E-07 No Data No Data 2.99E-04 9.29E-06 co-60 No Data 3.55E-06 6.12E-06 No Data No Data 1.91E-03 2.60E-05 Ni-63 2.22E-04 1.25E-05 7.56E-06 No Data No Data 7.43E-05 1.71E-06 Ni-65 8.08E-10 7.99E-11 4.44E-11 No Data 2.27E-05 NJ No Data 2.21E-06 Cu-64 No Data 5.39E-10 2.90E-10 No Data 1.63E-09 2.59E-06 9.92E-06 Zn-65 1.15E-05 3.06E-05 1.90E-05 No Data 1.93E-05 2.69E-04 4.41E-06 Zn-69 1.01E-11 2.61E-11 2.41E-12 No Data 1.58E-11 3.84E-07 2.75E-06 Br-83 No Data No Data 1.28E-07 No Data No Data No Data No Data 8r-84 No Data No Data 1.48E-07 No Data No Data No Data No Data Br-85 No Data No Data 6.84E-09 No Data No Data No Data No Data Rb-86 No Data 5.36E-05 3.09E-05 No Data No Data No Data 2.16E-06 Rb-88 No Data 1.52E-07 9.90E-08 No Data No Data No Data 4.66E-09 Rb-89 No Data 9.33E-08 7.83E-08 No Data No Data No Data 5.11E-10 Sr-89 1.62E-04 No Data 4.66E-06 No Data No Data 5.83E-04 4.52E-05 Sr-90 2.73E-02 No Data 1.74E-03 No Data No Data 3.99E-03 9.28E-05 Sr-91 3.28E-08 No Data 1.24E-09 No Data No Data 1.44E-05 4,70E-05 All values are in (mram/pci inhaled) . They are obtained from

(*,3 Reference 3 (Table E-9). Neither Reference 2 nor Reference 3 () contains data for Rh-105, Sb-124, or Sb-125. 9-23 Gen. Rev. 13

   .a; .~                                                                                          ~

l l I l l l l FNP-0-M-011 Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group i l Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 3.54E-09 No Data 1.42E-10 No Data No Data 6.49E-06 6.55E-05 Y-90 1.11E-06 No Data 2.99E-08 No Data No Data 7.07E-05 7.24E-05 I Y-91m No Data 1.37E-10 4.98E-12 No Data No Data 7.60E-07 4.64E-07 i 4 Y-91 2.47E-04 No Data 6.59E-06 No Data No Data 7.10E-04 i 4.97E-05 j j Y-92 5.50E-09 No Data 1.57E-10 No Data No Data 6.46E-06 6.46E-05 Y-93 5.04E-08 No Data 1.38E-09 No Data No Data 2.01E-05 1.05E-04 ! Zr-95 5.13E-05 1.13E-05 1.00E-05 No Data 1.61E-05 6.03E-04 1.65E-05 Zr-97 5.07E-08 7.34E-09 4.32E-09 No Data 1.05E-08 3.06E-05 i 9.49E-05 Nb-95 6.35E-06 2.48E-06 1.77E-06 No Data 2.33E-06 1.66E-04 1.00E-05 Mo-99 No Data 4.66E-08 1.15E-08 No Data 1.06E-07 3.66E-05 3.42E-05 Tc-99m 4.81E-13 9.41E-13 1.56E-11 No Data 1.37E-11 2.57E-07 i 1.30E-06

    -      Tc-101  2.19E-14 2.30E-14 2.91E-13     No' Data 3.92E-13 1.58E-07  4.41E-09 Ru-103  7.55E-07  No Data   2.90E-07   No Data  1.90E-06 1.79E-04  1.21E-05 Ru-105  4.13E-10  No Data   1.50E-10   No Data  3.63E-10 4.30E-06  2.69E-05 Ru-106  3.68E-05  No Data   4.57E-06   No Data  4.97E-05 3.87E-03 J

1.16E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data , Ag-110m 4.56E-06 3.08E-06 2.47E-06 No Data 5.74E-06 1.48E-03' 7.71E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data i Sb-125 No Data No Data No Data No Data No Data No Data No Data i Te-125m 1.82E-06 6.29E-07 2.47E-07 5.20E-07 No Data 1.29E-04 9.13E-06 Te-127m 6.72E-06 2.31E-06 8.16E-Q7 1.64E-06 1.72E-05 4,00E-04 1.935 05 Te-127 7.49E-10 2.57E-10 1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 Te-129m 5.19E-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.91E-05 Te-129 2.64E-11 9.45E-12 6.44E-12 1.93E-11 6.94E-11 7.93E-07 6.89E-06 Te-131m 3.63E-08 1.60E-08 1.37E-08 2.64E-08 1.08E-07 5.56E-05 8.32E-05 Te-131 5.87E-12 2.28E-12 1.78E-12 4.59E-12 1.59E-11 5.55E-07 3.60E-07 O V 9-24 Gen. Rev. 13

FNP-0-M-011 Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lang GI-LLI Te-132 1.30E-07 7.36E-08 7.12E-08 8.58E-08 4.79E-07 1.02E-04 3.72E-05 I-130 2.21E-06 4.43E-06 2.28E-06 4.99E-04 6.61E-06 No Data 1.38E-06 I-131 1.30E-05 1.30E-05 7.37E-06 4.39E-03 2.13E-05 No Data 7.68E-07 I-132 5.72E-07 1.10E-06 5.07E-07 5.23E-05 1.69E-06 No Data 8.65E-07 I-133 4.48E-06 5.49E-06 2.08E-06 1.04E-03 9.13E-06 No Data 1.48E-06 I-134 3.17E-07 5.84E-07 2.69E-07 1.37E-05 8.92E-07 No Data 2.58E-07 I-135 1.33E-06 2.36E-06 1.12E-06 2.14E-04 3.62E-06 No Data 1.20E-06 Co-134 1.76E-04 2.74E-04 6.07E-05 No Data 8.93E-05 3.27E-05 1.04E-06 Cs-136 1.76E-05 4.62E-05 3.14E-05 No Data 2.58E-05 3.93E-06 1.13E-06 Cs-137 2.45E-04 2.23E-04 3.47E-05 No Data 7.63E-05 2.81E-05 9.78E-07 Cs-138 1.71E-07 2.27E-07 1.50E-07 No Data 1.68E-07 1.84E-08 7.29E-08 g Ba-139 4.98E-10 2.66E-13 1.45E-11 No Data 2.33E-13 1.56E-06 1.56E-05 Ba-140 2.00E-05 1.75E-08 1.17E-06 No Data 5.71E-09 4.71E-04 2.75E-05 Ba-141 5.29E-11 2.95E-14 1.72E-12 No Data 2.56E-14 7.89E-07 7.44E-08 Ba-142 1.35E-11 9.73E-15 7.54E-13 No Data 7.87E-15 4.44E-07 7.41E-10 La-140 1.74E-07 6.08E-08 2.04E-08 No Data No Data 4.94E-05 6.10E-05 La-142 3.50E-10 1.11E-10 3.49E-11 No Data No Data 2.35E-06 2.05E-05 Co-141 1.06E-05 5.28E-06 7.83E-07 No Data 2.31E-06 1.47E-04 1.53E-05 Ca-143 9.89E-08 5.37E-08 7.77E-09 No Data 2.26E-08 3.12E-05 3.44E-05 Co-144 1.83E-03 5.72E-04 9.77E-05 No Data 3.17E-04 3.23E-03 1.05E-04 Pr-143 4.99E-06 1.50E-06 2.47E-07 No Data 8.11E-07 1.17E-04 2.63E-05 Pr-144 1.61E-11 4.99E-12 8.10E-13 No Data 2.64E-12 4.23E-07 5.32E-08 Nd-147 2.92E-06 2.36E-06 1.84E-07 No Data 1.30E-06 8.87E-05 2.22E-05 W-187 4.41E-09 2.61E-09 1.17E-09 No Data No Data 1.11E-05 2.46E-05 Np-239 1.26E-07 9.04E-09 6.35E-09 No Data 2.63E-08 1.57E-05 1.73E-05 9-25 Gen. Rev. 13

FNP-0-M-011 Tr.ble 9-9. Inhalation Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 C-14 3.25E-06 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 Na-24 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 P-32 2.36E-04 1.37E-08 8.95E-06 No Data No Data No Data 1.16E-05 Cr-51 No Data No Data 1.69E-08 9.37E-09 3.84E-09 2.62E-06 3.75E-07 Mn-54 No Data 6.39E-06 1.05E-06 No Data 1.59E-06 2.48E-04 8.35E-06 Mn-56 No Data 2.12E-10 3.15E-11 No Data 2.24E-10 1.90E-06 7.18E-06 Fe-55 4.18E-06 2.98E-06 6.93E-07 No Data No Data 1.55E-05 7.99E-07 Fe-59 1.99E-06 4.62E-06 1.79E-06 No Data No Data 1.91E-04 2.23E-05 Co-58 No Data 2.59E-07 3.47E-07 No Data No Data 1.68E-04 1.19E-05 Co-60 No Data 1.89E-06 2.48E-06 No Data No Data 1.09E-03 3.24E-05 Ni-63 7.25E-05 5.43E-06 2.47E-06 No Data No Data 3.84E-05 1.77E-06 h Ni-65 2.73E-10 3.66E-11 1.59E-11 No Data No Data 1.17E-06 4.59E-06 4 Cu-64 No Data 2.54E-10 1.06E-10 No Data 8.01E-10 1.39E-06 7.68E-06 Zn-65 4.82E-06 1.67E-05 7.80E-06 No Data 1.08E-05 1.55E-04 5.83E-06 Zn-69 6.04E-12 1.15E-11 8.07E-13 No Data 7.53E-12 1.98E-07 3.56E-08 Br-83 No Data No Data 4.30E-08 No Data No Data No Data No Data Br-84 No Data No Data 5.41E-08 No Data No Data No Data No Data Br-85 No Data No Data 2.29E-09 No Data No Data No Data No Data Rb-86 No Data 2.38E-03 1.05E-05 No Data No Data No Data 2.21E-06 Rb-88 No Data 6.82E-08 3.40E-08 No Data No Data No Data 3.65E-15 Rb-89 No Data 4.40E-08 2.91E-08 No Data No Data- .No Data 4.222-17 Sr-89 5.43E-05 No Data 1.56E-06 No Data No Data 3.02E-04 4.64E-05 Sr-90 1.35E-02 No Data 8.35E-04 No Data No Data 2.06E-03 9.56E-05 Sr-91 1.10E-08 No Data 4.39E-10 No Data No Data 7.59E-06 3.24E-05

 .w             All values are in (mrom/pci inhaled).                  They. are obtained from
      )         Reference 3 (Table E-8). Neither Reference 2 nor Reference 3 d              contains data for Rh-105, Sb-124, or Sb-125.

l 9-26 Gen. Rev. 13

_ - - _____ - ---- ~ FNP-0-M-011 Table 9-9 (contd). Inhalation Dose Factors for the Teenager Age Group k Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.19E-09 No Data 5.08E-11 No Data No Data 3.43E-06 1.49E-05 Y-90 3.73E-07 No Data 1.00E-08 No Data No Data 3.66E-05 6.99E-05 Y-91m 4.63E-11 No Data 1.77E-12 No Data No Data 4.00E-07 3.77E-09 Y-91 8.26E-05 No Data 2.21E-06 No Data No Data 3.67E-04 5.11E-05 Y-92 1.84E-09 No Data 5.36E-11 No Data No Data 3.35E-06 2.06E-05 Y-93 1.69E-08 No Data 4.65E-10 No Data No Data 1.04E-05 7.24E-05 Zr-95 1.82E-05 5.73E-06 3.94E-06 No Data 8.42E-06 3.36E-04 1.86E-05 Zr-97 1.72E-08 3.40E-09 1.57E-09 No Data 5.15E-09 1.62E-05 7.88E-05 Nb-95 2.32E-06 1.29E-06 7.08E-07 No Data 1.25E-06 9.39E-05 1.21E-05 Ho-99 No Data 2.11E-08 4.03E-09 No Data 5.14E-08 1.92E-05 3.36E-05 Tc-99m 1.73E-13 4.83E-13 6.24E-12 No Data 7.20E-12 1.44E-07 7.66E-07 Tc-101 7.40E-15 1.05E-14 1.03E-13 No Data 1.90E-13 8.34E-08 1.09E-16 Ru-103 2.63E-07 No Data 1.12E-07 No Data 9.29E-07 9.79E-05 1.36E-05 h Ru-105 1.40E-10 No Data 5.42E-11 No Data 1.76E-10 2.27E-06 1.13E-05 Ru-106 1.23E-05 No Data 1.55E-06 No Data 2.38E-05 2.01E-03 1.20E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 1.73E-06 1.64E-06 9.99E-07 No Data 3.13E-06 8.44E-04 3.41E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data To-125m 6.10E-07 2.80E-07 8.34E-08 1.75E-07 No Data 6.70E-05 9.38E-06 Te-127m 2.25E-06 1.02E-06 2.73E-07 5.48E-07 8.17E-06 2.07E-04 1.99E-05 Te-127 2.51E-10 1.14E-10 5.52E-11 1.77E-10 9.10E-10 1.40E-06 1.01E-05 To-129m 1.74E-06 8.23E-07 2.81E-07 5.72E-07 6.49E-06 2.47E-04 5.06E-05 To-129 8.87E-12 4.22E-12 2.20E-12 6.48E-12 3.32E-11 4.12E-07 2.02E-07 To-131m 1.23E-08 7.51E-P) 5.03E-09 9.06E-09 5.49E-08 2.97E-05 7.76E-v5 Te-131 1.97E-12 1.04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.89E-09 I 9-27 Gen. Rev. 13

FNP-0-M-011 k Table 9-9 (contd). Inhalation Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 4.50E-08 3.63E-08 2.74E-08 3.07E-08 2.44E-07 5.61E-05 5.79E-05 I-130 7.80E-07 2.24E-06 8.96E-07 1.86E-04 3.44E-06 No Data 1.14E-06 I-131 4.43E-06 6.14E-06 3.30E-06 1.83E-03 1.05E-05 No Data 8.11E-07 I-132 1.99E-07 5.47E-07 1.97E-07 1.89E-05 8.6FE-07 No Data 1.59E-07 I-133 1.52E-06 2.56E-06 7.78E-07 3.65E-04 4.453-06 No Data 1.29E-06 I-134 1.11E-07 2.90E-07 1.05E-07 4.94E-06 4.58E-07 No Data 2.55E-09 I-135 4.62E-07 1.18E-06 4.36E-07 7.76E-05 1.86E-06 No Data 8.69E-07 Cs-134 6.28E-05 1.41E-04 6.86E-05 No Data 4.69E-05 1.83E-05 1.22E-06 Cs-136 6.44E-06 2.42E-05 1.71E-05 No Data 1.38E-05 2.22E-06 1.36E-06 Co-137 8.38E-05 1.06E-04 3.89E-05 No Data 3.80E-05 1.51E-05 1.06E-06 Cs-138 5.82E-08 1.07E-07 5.58E-08 No Data 8.28E-08 9.84E-09 3.38E-11 l Ba-139 1.67E-10 1.18E-13 4.87E-12 No Data 1.11E-13 8.08E-07 8.06E-07 Ba-140 6.84E-06 8.38E-09 4.40E-07 No Data 2.85E-09 2.54E-04 2.86E-05

• Ba-141 1.78E-11 1.32E-14 5.93E-13 No Data 1.23E-14 4.11E-07 9.33E-14 Ba-142 4.62E-12 4.63E-15 2.84E-13 No Data 3.92E-15 2.39E-07 5.99E-20 La-140 5.99E-08 2.95E-08 7.82E-09 No Data No Data 2.68E-05 6.09E-05 La-142 1.20E-10 5.31E-11 1.32E-11 No Data No Data 1.27E-06 1.50E-06 Ce-141 3.55E-06 2.37E-06 2.71E-07 No Data 1.11E-06 7.67E-05 1.58E-05 Co-143 3.32E-08 2.42E-08 2.70E-09 No Data 1.08E-08 1.63E-05 3.19E-05 Ce-144 6.11E-04 2.53E-04 3.28E-05 No Data 1.51E-04 1.67E-03 1.08E-04 Pr-143 1.67E-06 6.64E-07 8.28E-08 No Data 3.86E-07 6.04E-05 2.67E-05 Pr-144 5.37E-12 2.20E-12 2.72E-13 No Data 1.26E-12 2.19E-07 2.94E-14 Nd-147 9.83E-07 1.07E-06 6.41E-08 No Data 6.28E-07 4.65E-05 2.28E-05 W-187 1.50E-09 1.22E-09 4.29E-10 No Data No Data 5.92E-06 2.21E-05 Np-239 4.23E-08 3.99E-09 2.21E-09 No Data 1.25E-08 8.11E-06 1.65E-05 I

9-28 Gen. Rev. 13

1 I FNP-0-M-011 Table 9-10. Inhalation Dose Factors for the Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 C-14 2.27E-06 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 Na-24 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 P-32 1.65E-04 9.64E-06 6.26E-06 No Data No Data No Data 1.08E-05 Cr-51 No Data No Data 1.25E-08 7.44E-09 2.85E-09 1.80E-06 4.15E-07 Mn-54 No Data 4.95E-06 7.87E-07 No Data 1.23E-06 1.75E-04 9.67E-06 Mn-56 No Data 1.55E-10 2.29E-11 No Data 1.63E-10 1.18E-06 2.53E-06 Fe-55 3.07E-06 2.12E-06 4.93E-07 No Data No Data 9.01E-06 7.54E-07 Fe-59 1.47E-06 3.47E-06 1.32E-06 No Data No Data 1.27E-04 2.35E-05 Co-58 No Data 1.98E-07 2.59E-07 No Data No Data 1.16E-04 1.33E-05 Co-60 No Data 1.44E-06 1.85E-06 No Data No Data 7.46E-04 3.56E-05 Ni-63 5.40E-05 3.93E-06 1.81E-06 No Data No Data 2.23E-05 1.67E-06 Ni-65 1.92E-10 2.62E-11 1.14E-11 No Data No Data 7.00E-07 1.54E-06 Cu-64 No Data 1.83E-10 7.69E-11 No Data 5.78E-10 8.48E-07 6.12E-06 Zn-65 4.05E-06 1.29E-05 5.82E-06 No Data 8.62E-06 1.08E-04 6.68E-06 Zn-69 4.23E-12 8.14E-12 5.65E-13 No Data 5.27E-12 1.15E-07 2.04E-09 Br-83 No Data No Data 3.01E-08 No Data No Data No Data 2.90E-08 Br-84 No Data No Data 3.91E-08 No Data No Data No Data 2.05E-13 Br-85 No Data No Data 1.60E-09 No Data No Data No Data No Data Rb-86 No Data 1.69E-05 7.37E-06 No Data No Data No Data 2.08E-06 Rb-88 No Data 4.84E-08 2.41E-08 No Data No Data No Data 4.18E-19 Rb-89 No Data 3.20E-08 2.12E-08 No Data No Data No Data 1.16E-21 Sr-89 3.80E-05 No Data 1.09E-06 No Data No Data 1.75E-04 4.37E-05 Sr-90 1.24E-02 No Data 7.62E-04 No Data No Data 1.20E-03 9.02E-05 Sr-91 7.74E-09 No Data 3.13E-10 No Data No Data 4.56E-06 2.39E-05 n All values are in (mres/pci inhaled). They are obtained from y Reference for Rh-105,3Sb-124, (Table and E-7),Sb-125. axcept as follows: Reference 2 (Table C-1) 1 9-29 Gen. Rev. 13

 ~                                 __                 _                                                     ._

FNP-0-M-011 Table 9-10 (contd). Inhalation Dose Factors for th1 Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI _ Sr-92 8.43E-10 No Data 3.64E-11 No Data No Data 2.06E-06 5.38E-06 Y-90 2.61E-07 No Data 7.01E-09 No Data No Data 2.12E-05 6.32E-05 Y-91m 3.26E-11 No Data 1.27E-12 No Data No Data 2.40E-07 1.66E-10 Y-91 5.78E-05 No Data 1.55E-06 No Data No Data 2.13E-04 4.81E-05 Y-92 1.29E-09 No Data 3.77E-11 No Data No Data 1.96E-06 9.19E-06 Y-93 1.18E-08 No Data 3.26E-10 No Data No Data 6.06E-06 5.27E-05 tr-95 1.34E-05 4.30E-06 2.91E-06 No Data 6.77E-06 2.21E-04 1.88E-05 Zr-97 1.21E-08 2.45E-09 1.13E-09 No Data 3.71E-09 9.84E-06 6.54E-05 Nb-95 1.76E-06 9.77E-07 5.26E-07 No Data 9.67E-07 6.31E-05 1.30E-05 Mo-99 No Data 1.51E-08 2.87E-09 No Data 3.64E-08 1.14E-05 3.10E-05 Tc-4 "N 1.29E-13 3.64E 13 4.63E-12 No Data 5.52E-12 9.55E-08 5.20E-07 Tc-101 5.22E-15 7.52E-15 7.38E-14 No Data 1.35E-1? 4.99E-08 1.36E-21 Ru-103 1.91E-07 No Data 8.23E-08 No Data 7.29E-07 6.31E-05 1.38E-05 Ru-105 9.88E-11 No Data 3.89E-11 No Data 1.27E-10 1.37E-06 6.02E-06 Ru-106 8.64E-06 No Data 1.09E-06 No Data 1.67E-05 1.17E-03 1.14E-04 Rh-105 9.24E-10 6.73E-10 4.43E .0 No Data 2.86E-09 2.41E-06 1.09E-05 Ag-110m 1.35E-06 1.25E-06 7.43E-07 No Data 2.46E-06 5.79E-04 3.78E-05 Sb-124 3.90E-06 7.36E-08 1.55E-06 9.445-09 No Data 3.10E-04 5.08E-05 [ Sb-125 8.26E-06 8.91E-08 1.66E-06 7.34E-09 No Data 2.75E-04 1.26E-05 e Te-125m 4.27E-07 1.98E-07 5.84E-Oh 1.31E-07 1.55E-06 3.92E-05 8.83E-06 Te-127m 1.58E-06 7.21E-07 1.96E-07 4.11E-07 5.72E-06 1.20E-04 1.87E-05 Te-127 1.75E-10 8.03E-11 3.87E-11 1.32E-10 6.37E 10 8.14E-07

                                                                                                    *7E-06
   . Tc-129m   1.22E-06 5.84E-07     1.98E-07   4.30E-07 4.57E-06                   1.45E-04 4.79E-05 To-129   c.22E-12 2.99E-12     1.55E-12   4.o7E-12 2.34E-11 2.42E-07                     1.96E-08 Te-131m 8.74R-09 5.45E-09      3.63E-09   6.88E-09 3.86E-08 1.82E-05                     6.95E-05 Tsi-131  1.39E-12 7.44E-13     4.49E-13   1.17E-12 5.46E-12                   1.74E-07   2.40E-09 C

9-30 Gen. Rev. 13

                                                                                                               .* 7
                                                           - . _ _ _ _ _ _ _                                   me

FNP-O-M-011 Table 9-10 (contd). Inhalation Dose Factors for the Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 3.25E-08 2.69E-08 2.02E-08 2.37E-08 1.82E-07 3.60E-05 6.37E-05 I-130 5.72E-O' 1.68E-06 6.60E-07 1.42E-04 2.61E-06 No Data 9.61E-07 I-131 3.15E-06 4.47E-06 2.56E-06 1.49E-03 7.66E-06 No Data 7.85E-07 I-132 1.45E-07 4.07E-07 1.45E-07 1.43E-05 6.48E-07 No Data 5.08E-08 1-133 1.08E-06 1.85E-06 5.65E-07 2.69E-04 3.23E-06 No Data 1.11E-06 I-134 8.05E-08 2.16E-07 7.69E-08 3.73E-06 3.44E-07 No Data 1.26E-10 I-135 3.35E-07 8.73E-07 3.21E-07 5.60E-05 1.39E-06 No Data 6.56E-07 Cs-134 4.66E-05 1.06E-04 9.10E-05 No Data 3.59E-05 1.22E-05 1.30E-06 Cs-136 4.88E-06 1.83E-05 1.38E-05 No Data 1.07E-05 1.50E-06 1.46E-06 Cs-137 5.98E-05 7.76E-05 5.35E-05 No Data 2.78E-05 9.40E-06 1.05E-06

      ) Cs-138     4.14E-08 7.76E-08 4.05E-08     No Data  6.00E-08 6.07E-09   2.33E-13 Ba-139    1.17E-10 8.32E-14 3.42E-12     No Data  7.78E-14 4.70E-07   1.12E-07 Ba-140    4.88E-06 6.13E-09 3.21E-07     No Data  2.09E-09 1.59E-C4   2.73E-05

• Ba-141 1.25E-11 9.41E-15 4.20E-13 No Data 8.75E-15 2.42E-07 1.45E-17 Ba-142 3.29E-12 3.38E-15 2.07E-13 No Data 2.86E-15 1.49E-07 1.96E-26 La-140 4.30E-08 2.17E-08 5.73E-09 No Data No Data 1.70E-05 5.73E-05 La-142 8.54E-11 3.88E-11 9.65E-12 No Data No Data 7.91E-07 2.64E-07 Ce-141 2.49E-06 1.69E-06 1.91E-07 No Data 7.83E-07 4.52E-05 1.50E-05 Co-343 2.33E-08 1.72E-08 1.91E-09 No Data 7.60E-09 9.97E-06 2.83E-05 Co-144 4.29E-04 1.79E-04 2.30E-05 No Data 1.06E-04 9.72E-04 1.02E-04

< Pr-143 1.17E-06 4.69E-07 5.80E-08 No Data 2.70E-07 3.51E-05 2.50E-05 Pr-144 3.76E-12 1.56E-12 1.91E-13 No Data 8.81E-13 1.27E-07 2.69E-18 Nd-147 6.59E-07 7.62E-07 4.56E-08 No Data 4.45E-07 2.76E-05 2.16E-05 W-187 1.06E-09 8.85E-10 3.10E-10 No Data No Data 3.63E-06 1.94E-05 Np-239 2.87E-08 2.82E-09 1.55E-09 No Data 8.75E-09 4.70E-06 1.49E-05 9 . m 9-31 Gen. Rev. 13 h

FMP-0-M-011 Tcble 9-11. Ingestion Dose Factors for the Infant Age Group I Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 C-14 2.37E-05 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 Na-24 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 P-32 1.70E-03 1.00E-04 6.59E-05 No Data No Data No Data 2.30E-05 I Cr-51 No Data No Data 1.41E-08 9.20E-09 2.01E-09 1.79E-08 4.11E-07 Mn-54 No Data 1.99E-05 4.51E-06 No Data 4.41E-06 No Data 7.31E-06 Mn-56 No Data 8.18E-07 1.41E-07 No Data 7.03E-07 No Data 7.43E-05 Fe-55 1.39E-05 8.98E-06 2.40E-06 No Data No Data 4.39E-06 1.14E-06 Fe-59 3.08E-05 5.38E-05 2.12E-05 No Data No Data 1.59E-05 2.57E-05 Co-58 No Data 3.60E-06 8.98E-Oo No Data No Data No Data 8.97E-06 Co-60 No Data 1.08E-05 2.55E-05 No Data No Data No Data 2.57E-05 Ni-63 6.34E-04 3.92E-05 2.20E-05 No Data No Data No Data 1.95E-06 Ni-65 4.70E-06 5.32E-07 2.42E-07 No Data No Data No Data 4.05E-05 g Cu-64 No Data 6.09E-07 2.82E-07 No Data 1.03E-06 No Data 1.25E-05 Zn-65 1.84E-05 6.31E-05 2.91E-05 No Data 3.06E-05 No Data 5.33E-05 Zn-69 9.33E-08 1.68E-07' 1.25E-08 Ho Data 6.98E-08 No Data 1.37E-05 Br-83 No Data No Data 3.63E-07 No Data No Data No Data No Data Br-84 No Data No Data 3.82E-07 No Data No Data No Data No Data Br-85 No Data No Data 1.94E-08 No Data No Data No Data No Data Rb-86 No Data 1.70E-04 8.40E-05 No Data No Data No Data 4.35E-06 Rb-88 No Data 4.98E-07 2.73E-07 No Data No Data No Data 4.85E-07 Rb-89 No Data 2.86E-07 1.97E-07 No Data No Data No Data 9.74E-08 Sr-89 2.51E-03 No Data 7.20E-05 No Data No Data No Data 5.16E-05 Sr-90 1.85E-02 No Data 4.71E-03 No Data No Data No Data 2.31E-04 Sr-91 5.00E-05 No Data 1.81E-06 No Data No Data No Data 5.92E-05 All values are in (mrom/pci ingested) . They are obtained from Reference 3 (Tahls F-14). Neither Reference 2 nor Reference 3 contwins- dat1r for Rh-105, Sb-124, or sb-125. I

       ~

9-32 Gen. Rsev. 13

                     ~                                             _.          __          _   . _ .

G FNP-0-M-011 Table 9-11 (cont'). d Ingestion Dose Factors for the Infant Age Group 4 Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.92E-05 No Data 7.13E-07 No Data No Data No Data 2.07E-04 Y-90 84 'fE-08 No Data 2.33E-09 No Data No Data No Data 1.20E-04 Y-91m i ,sJE-10 No Data 2.76E-11 No Data No Data No Data 2.70E-06 Y-91 1.13E-06 No Data 3.01E-08 No Data No Data No Data 8.10E-05 Y-92 7.65E-09 No Data 2.15E-10 No Data No Data No Dat a 1.46E-04 Y-93 2.43E-08 No Data 6.622-10 No Data No Data No Data 1.92E-04 Zr-95 2.06E-07 5.02E-08 3.56E-08 No Data 5.41E-08 No Data 2.50E-05

Zr-97 1.48E-08 2.54E-09 1.16E-09 No Data 2.56E-09 No Data 1.62E-04 Nb-95 4.20E-08 1.73E-08 1.00E-08 No Data 1.24E-08 No Data 1.46E-05 I Mo-99 No Data 3.40E-05 6.63E-06 No Data 5.08E-05 No Data 1.12E-05 Tc-99m 1.92E-09 3.96E-09 5.10E-08 No Data 4.26E-08 2.07E-09 1.15E-06 Tc-101 2.27E-09 2.06E-09 2.83E-08 No Data 3.40E-08 1.56E-09 4.86E-07

(- ( Ru-103 1.48E-06 No Data 4.95E-07 No Data 3.08E-06 No Data 1.80E-05 Ru-105 1.36E-07 No Data 4.58E-08 No Data 1.00E-06 No Data 5.41E-05

Ru-106 2.41E-05 No Data 3.01E-06 No Data 2.85E-05 No Data 1.83E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 9.96E-07 7.27E-07 4.81E-07 No Data 1.04E-06 No Data 3.77E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 2.J3E-05 7.795-06 3.15E-06 7.84E-06 No Data No Data 1.11E-05 4

Te-127m 5.85E-05 1.94E-05 7.08E-06 1.69E-05 1.44E-04 No Data 2.36E-05 Te-127 1.00E-06 3.35E-07 2.15E-07 8.14E-07 2.44E-06 No Data 2.10E-05

Te-129m 1.00E-04 3.43E-05 1.54E-05 3.84E-05 2.5vE-04 No Data 5.97E-05 l Te-129 2.84E-07 9.79E-08 6.63E-08 2.38E-07 7.07E-07 No Data 2.27E-05 i

!      Te-131m  1.52E-05 6.12E-06 5.05E-06 1.24E-05 4.21E-05            No Data     1.03E-04 Te-131  1.76E-07   6.50E-08 4.94E-08 1.57E-07 4.50E-07          No Data     7.11E-06 l O

V 9-33 Gen. Rev. 13

o y FNP-0-M-011

   )  Tcble 9-11 (contd).
 ,                              Ingestion Dose Factors for the Infant Age Group se Nuclide    Bone       Liver     T. Body    Thyroid    Kidney    Lung        GI-LLI Te-132  2.08E-05  1.03E-05 9.61E-06 1.52E-05 6.44E-05         No Data      3.81E-05 I-130  6.00E-06 1.32E-05 5.30E-06       1.48E-03 1.45E-05   No Data      2.83E-06        l I-131  3.59E-05 4.23E-05 1.86E-05 1.39E-02 4.94E-05         No Data      1.51E-06 I-132  1.66E-06 3.37E-06     1.20E-06 1.58E-04 3.766-06     No Data      2.73E-06 I-133  1.25E-05 1.82E-05 5.33E-06 3.31E-03 2.14E-05         No Data      3.08E-06 I-134  8.69E-07 1.78E-06 6.33E-07 4.15E-05 1.99E-06         No Data      1.84E-06 I-135  3.64E-06 7.24E-06 2.64E-06 6.49E-04 8.07E-06         No Data      2.62E-06 Cs-134  3.77E-04 7.03E-04 7.10E-05        No Data   1.81E-04 7.42E-05     1.T1E-06 Cs-136  4.59E-05 1.35E-04 5.04E-05        No Data   5.38E-05 1.10E-05     2.05E-06 l

Cs-137 5.22E-04 6.11E-04 4.33E-05 No Data 1.64E-04 6.64E-05 1.91E-06 Cs-138 4.81E-07 7.82E-07 3.79E-07 No Data 3.90E-07 6.09E-08 1.25E-06 l Ba-139 8.81E-07 5.84E-10 2.55E-08 No Data 3.51E-10 3.54E-10 5.58E-05 Ba-140 1.71E-04 1.71E-07 8.81E-06 No Data 4.06E-08 1.05E-07 4.20E-05 'l Ba-141 4.25E-07 2.91E-10 1.34E-08 No Data 1.75E-10 1.77E-10 5.19E-06 Ba-142 1.84E-07 1.53E-10 9.06E-09 No Data 8.81E-11 9.26E-11 7.59E-07 1 La-140 2.11E-08 8.32E-09 2.14E-09 No Data No Data No Data 9.77E-05 l l La-142 1.10E-09 4.04E-10 9.67E-11 No Data No 3ata No Data 6.86E-05 Ce-141 7.87E-08 4.80E-08 5.65E-09 No Data 1.48E-08 No Data 2.48E-05 l Ce-143 1.48E-08 9.82E-06 1.12E-09 No Data 2.86E-09 No Data 5.73E-05 Ce-144 2.98E-06 1.22E-06 1.67E-07 No Data 4.93E-07 No Data 1.71E-04 Pr-143 8.13E-08 3.04E-08 4.03E-09 No Data 1.13E-08 No Data 4.29E-05 i Pr-144 2.74E-10 1.06E-10 1.38E-11 No Data 3.84E-11 No Data 4.93E-06 Nd-147 5.53E-08 5.68E-08 3.48E-09 No Data 2.19E-08 No Data 3.60E-05 W-187 9.03E-07 6.28E-07 2.37E-07 No Data No Data No Data 3.69E-05 Np-239 1.11E-08 9.93E-10 5.61E-10 No Data 1.98E-09 No Data 2.87E-05 h 9-34 Gen. Rev. 13

Tcble 9-12. Ingestion Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03E-07 C-14 1.21E-05 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 Na-24 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 P-32 8.25E-04 3.86E-05 3.18E-05 No Data No Data No Data 2.28E-05 Cr-51 No Data No Data 8.90E-09 4.94E-09 1.35E-09 9.02E-09 4.72E-07 Mn-54 No Data 1.07E-05 2.85E-06 No Data 3.00E-06 No Data 8.98E-06 Mn-56 No Data 3.34E-07 7.54E-08 No Data 4.04E-07 No Data 4.84E-05 Fe-55 1.15E-05 6.10E-06 1.89E-06 No Data No Data 3.45E-06 1.13E-06 Fe-59 1.65E-05 2.67E-05 1.33E-05 No Data No Data 7.74E-06 2.78E-05 Co-58 No Data 1.80E-06 5.51E-06 No Data No Data No Data 1.05E-05 l Co-60 No Data 5.29E-06 1.56E-05 No Data No Data No Data 2.93E-05 l Ni-63 5.36E-04 2.88E-05 1.83E-05 No Data No Data No Data 1.94E-06 i Ni-65 2.22E-06 2.09E-07 1.22E-07 No Data No Data No Data 2.56E-05 Cu-64 No Data 2.45E-07 1.48E-07 No Data 5.92E-07 No Data 1.15E-05 l Zn-65 1.37E-05 3.65E-05 2.27E-05 No Data 2.30E-05 No Data 6.41E-06 Zn-69 4.38E-08 6.33E-08 5.85E-09 No Data 3.84E-08 No Data 3.99E-06 Br-83 No Data No Data 1.71E-07 No Data No Data No Data No Data Br-84 No Data No Data 1.98E-07 No Data No Data No Data No Data Br-85 No Data No Data 9.12E-09 No Data No Data No Data No Data Rb-86 No Data 6.70E-05 4.12E-05 No Data No Data No Data 4.31E-06 Rb-08 No Data 1.90E-07 1.32E-07 No Data No Data No Data 9.32E-09 Rb-89 No Data 1.17E-07 1.04E-07 No Data No Data No Data 1.02E-09 Sr-89 1.32E-03 No Data 3.77E-05 No Data No Data No Data 5.11E-05 Sr-90 1.70E-02 No Data 4.31E-03 No Data No Data No Data 2.29E-04 Sr-91 2.40E-05 No Data 9.06E-07 No Data No Data No Data 5.30E-05 All values are in (mrem /pci ingested) . They are obtained from l Reference 3 (Table E-13). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125. 9-35 Gen. Rev. 13

FNP-0-M-011 Tcble 9-12 (con d). Ingestion Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 9.03E-06 No Data 3.62E-07 No Data No Data No Data 1.71E-04 Y-90 4.11E-08 No Data 1.10E-09 No Data No Data No Data 1.17E-04 Y-91m 3.82E-10 No Data 1.39E-11 No Data No Data No Data 7.48E-07 Y-91 6.02E-07 No Data 1.61E-W No Data No Data No Data 8.02E-05 Y-92 3.60E-09 No Data 1.03E-10. No Data No Data No Data 1.04E-04 Y-93 1.14E-08 No Data 3.13E-10 No Data No Data No Data 1.70E-04 l Zr-95 1.16E-07 2.55E-08 2.27E-08 No Data 3.65E-08 No Data 2.66E-05 Zr-97 6.99E-09 1.01E-09 5.96E-10 No Data 1.45E-09 No Data 1.53E-04 Nb-95 2.25E-08 8.76E-09 6.26E-09 No Data 8.23E-09 No Data 1.62E-05 l Mo-99 No Data 1.33E-05 3.29E-06 No Data 2.84E-05 No Data 1.10E-05 l Tc-99m 9.23E-10 1.81E-09 3.00E-08 No Data 2.63E-08 9.19E-10 1.03E-06 Tc-101 1.07E-09 1.12E-09 1.42E-08 No Data 1.91E-08 5.92E-10 3.56E-09 s l Ru-103 7.31E-07 No Data 2.81E-07 No Data 1.84E-06 No Data 1.89E-05

 /

Ru-105 6.45E-08 No Data 2.34E-08 No Data 5.67E-07 No Data 4.21E-05 Ru-106 1.17E-05 No Data 1.46E-06 No Data 1.58E-05 No Data 1.82E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 5.39E-07 3.64E-07 2.91E-07 No Data 6.78E-07 No Data 4.33E-05 Sb-124 No Data No Data No Cata No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 1.14E-05 3.09E-06 1.52E-06 3.20E-06 No Data No Data 1.10E-05 Te-127m 2.89E-05 7.78E-06 3.43E-06 6.91E-06 8.24E-05 No Data 2.34E-05 Te-127 4.71E-07 1.27E~07 1.01E-07 '.T.2EE-07 'r.34E-06 No Data 1. ts 4 E-05' Te-129m 4.87E-05 1.36E-05 7.56E-06 1.57E-05 1.43E-04 No Data 5.94E-05 Te-129 1.34E-07 3.74E-08 3.18E-08 9.56E-08 3.92E-07 Ne Data 8.34E-06 Te-131m 7.20E-06 2.49E-06 2.65E-06 5.12E-06 2.41E-05 No Data 1.01E-04 Te-131 8.30E-08 2.53E-08 2.47E-08 6.35E-08 2.51E-07 No Data 4.36E-07 l 9-36 Gen. Rev. 13

1 1 FNP-0-M-011 Tcble 9-12 (contd). Ingestion Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 1.01E-05 4.47E-06 5.40E-06 6.51E-06 4.15E-05 No Data 4.50E-05 I-130 2.92E-06 5.90E-06 3.04E-06 6.50E-04 8.82E-06 No Data 2.76E-06 I-131 1.72E-05 1.73E-05 9.83E-06 5.72E-03 2.84E-05 No Data 1.54E-06 I-132 8.00E-07 1.47E-06 6.76E-07 6.82E-05 2.25E-06 No Data 1.73E-06 I-133 5.92E-06 7.32E-06 2.77E-06 1.36E-03 1.22E-05 No Data 2.95E-06 I-134 4.19E-07 7.78E-07 3.58E-07 1.79E-05 1.19E-06 No Data 5.16E-07 I-135 1.75E-06 3.15E-06 1.49E-06 2.79E-04 4.83E-06 No Data 2.40E-06 Cs-134 2.34E-04 3.84E-04 8.10E-05 No Data 1.19E-04 4.27E-05 2.07E-06 Cs-136 2.35E-05 6.46E-05 4.18E-05 No Data 3.44E-05 5.13E-06 2.27E-06 Cs-137 3.27E-04 3.13E-04 4.62E-05 No Data 1.02E-04 3.67E-05 1.96E-06 Cs-138 2.28E-07 3.17E-07 2.01E-07 No Data 2.23E-07 2.40E-08 1.46E-07

 )   Ba-139   4.14E-07 2.21E-10 1.20E-08          No Data  1.93E-10 1.30E-10   2.39E-05 Ba-140   8.31E-05 7.28E-08 4.85E-06          No Data  2.37E-08 4.34E-08   4.21E-05

• Ba-141 2.00E-07 1.12E-10 6.51E-09 No Data 9.69E-11 6.58E-10 1.14E-07 Ba-142 8.74E-08 6.29E-11 4.88E-09 No Data 5.09E-11 3.70E-11 1.14E-09 La-140 1.01E-08 3.53E-09 1.19E-09 No Data No Data No Data 9.84E-05 La-142 5.24E-10 1.67E-10 5.23E-11 No Data No Data No Data 3.31E-05 Ce-141 3.97E-08 1.98E-08 2.94E-09 No Data 8.68E-09 No Data 2.47E-05 Ce-143 6.99E-09 3.79E-06 5.49E-10 No Data 1.59E-09 No Data 5.55E-05 Ce-144 2.08E-06 6. 5;2 E-07 1.11E-07 No Data 3.61E-07 No Data 1.70E-04 Pr-143 , 3.93E-08 1.18E-08 1.95E-09 No Data 6.39E-09 No Data 4.24E-05 Pr-144 1.29E-10 3.99E-11 6.49E-12 No Data 2.11E-11 No Data 8.59E-08 Nd-147 2.79E-08 2.25E-08 1.75E-09 No Data 1.24E-08 No Data 3.58E-05 W-187 4.29E-07 2.54E-07 1.14E-07 No Data No Data No Data 3.57E-05 Np-239 5.25E-09 3.77E-10 2.65E-10 No Data 1.09E-09 No Data 2.79E-05 h

9-37 Gen. Rev. 13

FNP-0-M-011 Tcble 9-13. Ingestion Dose Factors for the Teenager Age Group 1 Nuc3ide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 C-14 4.06E-06 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 Na-24 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 P-32 2.76E-04 1.71E-05 1.07E-05 No Data No Data No Data 2.32E-05 Cr-51 No Data No Data 3.60E-09 2.00E-09 7.89E-10 5.14E-09 6.05E-07 Mn-54 No Data 5.90E-06 1.17E-06 No Data 1.76E-06 No Data 1.21E-05 Mn-56 No Data 1.58E-07 2.81E-08 No Data 2.00E-07 No Data 1.04E-05 Fe-55 3.78E-06 2.68E-06 6.25E-07 No Data No Data 1.70F.-06 1.16E-06 Fe-59 5.87E-06 1.37E-05 5.29E-06 No Deta No Data 4.3T.E-06 3.24E-05 Co-58 No Data 9.72E-07 2.24E-06 No Data No Data No Data 1.34E-05 Co-60 No Data 2.81E-06 6.33E-06 No Data No Data No Data 3.66E-05 Ni-63 1.77E-04 1.25E-05 6.00E-06 No Data No Data No Data 1.99E-06 Ni-65 7.49E-07 9.57E-08 4.36E-08 No Data No Data No Data 5.19E-06 Cu-64 No Data 1.15E-07 5.41E-08 No Data 2.91E-07 NoDataj8.92E-06 Zn-65 5.76E-06 2.00E-05 9.33E-06 No Data 1.28E-05 No Data b.4?S-06 2n-69 1.47E-08 2.80E-08 1.96E-09 No Data 1.83E-08 No Data 5.16E-08 Br-83 No Data No Data 5.74E-08 No Data No Data No Data No Data Br-84 No Data No Data 7.22E-08 No Data No Data No Data No Data Br-85 No Data No Data 3.05E-09 No Data No Data No Data No Data Rb-86 No Data 2.98E-05 1.40E-05 No Data No Data No Data 4.41E-06 Rb-88 No Data 8.52E-08 4.54E-08 No Data No Data No Data 7.30E-15 Rb-89 No Data 5.50E-08 .3.89E-08 No Data No Data No Data 8.43E-17 Sr-89 4.40E-04 No Data 1.26E-05 No Data No Data No Data 5.24E-05 Sr-90 8.30E-03 No Data 2.05E-03 No Data No Data No Data 2.33E-04 Sr-91 8.07E-06 No Data 3.21E-07 No Data No Data No Data 3.66E-05 All valuaa are in (mram/pci ingested) . They are obtained from Reference 3 (Table E-12). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125. 9-38 Gen. Rev. 13

FNP-0-M-011 I Tcble 9-13 (contd). Ingestion Dose Factors for the Teenager Age Group l Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l Sr-92 3.05E-06 No Data 1.30E-07 No Data No Data No Data 7.77E-05 Y-90 1.37E-08 No Data 3.69E-10 No Data No Data No Data 1.13E-04 l l Y-91m 1.29E-10 No Data 4.93E-12 No Data No Data No Data 6.09E-09 1 Y-91 2.01E-07 No Data 5.39E-09 No Data No Data No Data 8.24E-05 l Y-92 1.21E-09 No Data 3.50E-11 No Data No Data No Data 3.32E-05 Y-93 3.83E-09 No Data 1.05E-10 No Data No Data No Data 1.17E-04 Zr-95 4.12E-08 1.30E-08 8.94E-09 No Data 1.91E-08 No Data 3.00E-05 Zr-97 2.37E-09 4.69E-10 2.16E-10 No Data 7.11E-10 No Data 1.27E-04 Nb-95 8.22E-09 4.56E-09 2.51E-09 No Data 4.42E-09 No Data 1.95E-05 Mo-99 No Data 6.03E-06 1.15E-06 No Data 1.38E-05 No Data 1.08E-05 Tc-99m 3.32E-10 9.26E-10 1.20E-08 No Data 1.38E-08 5.14E-10 6.08E-07 Tc-101 3.60E-10 5.12E-10 5.03F-09 No Data 9.26E-09 3.12E-10 8.75E-17 Ru-103 2.55E-07 No Data 1.09E-07 No Data 8.99E-07 No Data 2.13E-05 Ru-105 2.18E-08 No Data 8.46E-09 No Data 2.75E-07 No Data 1.76E-05 Ru-100 3.92E-06 No Data 4.94E-07 No Data 7.56E-06 No Data 1.88E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 2.05E-07 1.94E-07 1.18E-07 No Data 3.70E-07 No Data 5.45E-05 Sb-124 No Data No Data No Data No Dat- No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 3.83E-06 1.38E-06 5.12E-07 1.07E-06 No Data No Data 1.13E-05 Te-127m 9.67E-06 3.43E-06 1.15E-06 2.30E-06 3.92E-05 No Data 2.41E-05 Te-127 1.58E-07 5.60E-08 3.40E-08 1.09E-07 6.40E-07 No Data 1.22E-05 Te-129m 1.63E-05 6.05E-06 2.58E-06 5.26E-06 6.82E-05 No Data 6.12E-05 Te-129 4.48E-08 1.67E-08 1.09E-08 3.20E-08 1.88E-07 No Data 2.45E-07 Te-131m 2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 No Data 9.39E-05 Te-131 2.79E-08 1.15E-08 8.72E-09 2.15E-08 1.22E-07 No Data 2.29E-09 0 . . 9-39 Gen. Rev. 13

I FNP-0-H-011 h Table 9-13 (contd). Ingestion Dose Factors for the Teenager Age Group i Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 3.49E-06 2.21E-06 2.08E-06 2.33E-06 2.12E-05 No Data 7.00E-05 I-130 1.03E-06 2.98E-06 1.19E-06 2.43E-04 4.59E-06 No Data 2.29E-06 I-131 5.85E-06 8.19E-06 4.40E-06 2.39E-03 1.41E-05 No Data 1.62E-06 I-132 2.79E-07 7.30E-07 2.62E-07 2.46E-05 1.15E-06 No Data 3.18E-07 I-133 2.01E-06 3.41E-06 1.04E-06 4.76E-04 5.98E-06 No Data 2.58E-06 I-134 1.46E-07 3.87E-07 1.39E-07 6.45E-06 6.10E-07 No Data 5.10E-09 I-135 6.10E-07 1.57E-06 5.82E-07 1.01E-04 2.48E-06 No Data 1.74E-06 Cs-134 8.37E-05 1.97E-04 9.14E-05 No Data 6.26E-05 2.39E-05 2.45E-06 Cs-136 8.59E-06 3.38E-05 2.27E-05 No Data 1.84E-05 2.90E-06 2.72E-06 Cs-137 1.12E-04 1.49E-04 5.19E-05 No Data 5.07E-05 1.97E-05 2.12E-06 Cs-138 7.76E-08 1.49E-07 7.45E-08 No Data 1.10E-07 1.28E-08 6.76E-11

  }      Ba-139    1.39E-07 9.78E-11   4.05E-09   No Data  9.22E-11 6.74E-11   1.24E-06 Ba-140    2.84E-05 3.Y5E-05 1.83E-06     No Data  1.18E-08 2.34E-08   4.38E-05 Ba-141    6.71E-08 5.01E-11 0.24E-09     No Data  4.65E-11 3.43E-11   1.43E-13 Ba-142    2.99E-08 2.99E-11   1.84E-09   No Data  2.53E-11 1.99E-11   9.18E-20 La-140    3.48E-09 1.71E-09 4.55E-10     No Data   No Data  No Data   9.82E-05 La-142    1.79E-10 7.95E-11 1.98E-11     No Data   No Data  No Data   2.42E-06 Ce-141     1.33E-08 8.88E-09   1.02E-09   No Data  4.18E-09  No Data   2.54E-05 Ce-143     2.35E-09  1.71E-06 1.91E-10    No Data  7.67E-10  No Data   5.14E-05 Ce-144     6.96E-07 2.88E-07 3.74E-08     No Data  1.72E-07  No Data   1.75E-04 Pr-143     1.31E-08 5.23E-09 6.52E-10     No Data  3.04E-09  No Data   4.31E-05 Pr-144     4.30E-11 1.76E-11 2.18E-12     No Data  1.01E-11  No Data   4.74E-14    ,

Nd-147 9.38E-09 1.02E-08 6.11E-10 No Data 5.99E-09 No Data 3.68E-05 W-187 1.46E-07 1.19E-07 4.17E-08 No Data No Data No Data 3.22E-05 Np-239 1.76E-09 1.66E-10 9.22E-11 No Data 5.21E-10 No Data 2.67E-05 I 9-40 Gen. Rev. 13

                                   ~ >          s c

pr s , ,. ,, , FNP-0-M-011 A Table 9-14. Ingestion Dose Factors for the Adult Age Group v) Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 P-32 1.93E-04 1.20E-05 7.46E-06 No Data No Data No Data 2.17E-05 Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 Mn-56 No Data 1.15E-07 2.04E-08 No Data 1.46E-07 No Data 3.67E-06 Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91E-06 No Data No Data 2.85E-06 3.40E-05 Co-58 No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51E-05 1 I Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 ' Ni-63 1.30E-04 9.01E-06 4.36E-06 No Data No Data No Data 1.88E-06 Ni-65 5.28E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 r Cl

                                                                                                           \

Cu-64 No Data B.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 l l Zn-65 4.84E-06 1.54E-05 6.96E-06 ' , ' - Sata 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-08 1.97E-08 1.37E-09 No Data 1.28E-08 No Data 2.96E-09 Br-83 No Data No Data 4.02E-08 No Data. No Data. No Data 5.79E-08 Br-84 No Data No Data 5.21E-08 No Data No Data No Data 4.09E-13 Br-85 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.21E-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4.01E-08 2.82E-08 No Data No Data No Data 2.33E-21 Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data No Data 4.94E-05 Sr-90 7.50E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 All values are in (mrem /pci ingested). They are obtained from , Reference 3 (Table E-11), except as follows: Reference 2 (Table A-3) for Rh-105, sb-124, and Sb-125. G/ k-9-41 Gen. Rev. 13

FNP-0-M-011 Tcble 9-14 (contd). Ingestion Dose Factors for the Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-11 No Data No Data No Data 1.48E-05 Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-05 Mo-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 60-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 I 9-42 Gen. Rev. 13

                    ~

FNP-0-M-011 Tcble 9-14 (contd). Ingestion Dose Factors for the Adult Age Group

m. . .

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 Cs-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 2.59E-06 Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 2.11E-06 Cs-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Bt-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 Bn-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 La-142 1.28E-10 5.82E-11 1.45E-11' No Data No Data No Data 4.25E-07 C3-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05 C3-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Co-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.03E-07 8.61E-08 3.012-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 9-43 Gen. Rev. 13

FNP-0-M-011

/   Table 9-15. External Dose Factors for Standing on Contaminated Ground s

Nuclide T. Body Skin Nuclide T. Body Skin H-3 0.00 0.00 Sr-91 7.10E-09 8.30E-09 C-14 0.00 0.00 Sr-92 9.00E-09 1.00E-08 Na-24 2.50E-08 2.902-08 Y-90 2.20E-12 2.60E-12 P-32 0.00 0.00 Y-91m 3.80E-09 4.40E-09 Cr-51 2.20E-10 2.60E-10 Y-91 2.40E-11 2.70E-11 Mn-54 5.80E-09 6.80E-09 Y-92 1.60E-09 1.90E-09 Mn-56 1.10E-08 1.30E-08 Y-93 5.70E-10 7.80E-10 Fe-55 0.00 0.00 Zr-95 5.00E-09 5.80E-09 Fe-59 8.00E-09 9.40E-09 Zr-97 5.50E-09 6.40E-09 Co-58 7.00E-09 8.20E-09 Nb-95 5.10E-09 6.00E-09 Co-60 1.70E-08 2.00E-08 Mo-99 1.90E-09 2.20E-09 Ni-63 0.00 0.00 Tc-99m 9.60E-10 1.10E-09 Ni-65 3.70E-OS 4.30E-09 Tc-101 2.70E-09 3.00E-09 Cu-64 1.50E-09 1.70E-09 Ru-103 3.60E-09 4.20E-09 Q Zn-65 4.00E-09 4.60E-09 Ru-105 4.50E-09 5.10E-09 U Zn-69 0.00 0.00 Ru-106 1.50E-09 1.80E-09 , Br-83 6.40E-11 9.30E-11 Rh-105 6.60E-10 7.70E-10 Br-84 1.20E-08 1.40E-08 Ag-110m 1.80E-08 2.10E-08 Br-85 0.00 0.00 Sb-124 1.30E-08 1.50E-08 Rb-86 6.30E-10 7.20E-10 Sb-125 3.10E-09 3.50E-09 Rb-88 3.50E-09 4.00E-09 Te-125m 3.50E-11 4.80E-11 Rb-89 1.50E-08 1.80E-08 Te-127m 1.10E-12 1.30E-12 Sr-89 5.60E-13 6.50E-13 Te-127 1.00E-11 1.10E-11 Sr-90 0.00 0.00 Te-129m 7.70E-10 9.00E-10 All values are in (mrem /h) per (pCi/m 2). They are obtained from Reference 3 (Table E-6), except as follows: Reference 2 (Table A-7) for Rh-105, sb-124, and Sb-125.

D G

9-44 Gen. Rev. 13

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l FNP-0-M-011 I T ble 9-15 (contd). External Dose Factors for Standing on Contaminated Ground l Nuclide T. Body Skin ) Te-129 7.10E-10 8.40E-10 Te-131m 8.40E-09 9.90E-09 i Te-131 2.20E-09 2.60E-06 Te-132 1.70E-09 2.00E-09 I-130 1.40E-08 1.70E-08 ] I-131 2.80E-09 3.40E-09 I-132 1.70E-08 2.00E-08 I-133 3.70E-09 4.50E-09 I-134 1.60E-08 1.90E-08 I-135 1.20E-03 1.40E-08 Cs-134 1.1DE-08 1.40E-08 Cs-136 1.50E-08 1.70E-08

 -g       Cs-137     4.20E-09         4.90E-09 m- /     Cs-138     2.10E-08         2.40E-08 Ba-139     2.40E-09         2.70E-09 Ba-140     2.10E-09         2.40E-09 Ba-141     4.30E-09         4.90E-09 Ba-142     7.90E-09         9.00E-09 La-140     1.50E-08         1.70E-08 La-142     1.50E-08         1.80E-08 Ce-141     5.50E-10         6.20E-10 co-143     2.20E-09         2.50E-09 Ce-144     3.20E-10         3.70E-10 Pr-143        0.00             0.00 Pr-144-    2.00E-10         2.30E-10 Nd-147     1.00E-09         1.20E-09 W-187     3.10E-09         3.60E-09 Np-239     9.50E-10         1.10E-09 k

9-45 Gen. Rev. 13

, , - _ _ . - _ _ _ _ _ ~ _ . _ . . - -..-_____..._._.m-_._ m_ _ _ _ _ ..._ _~ . . . _ _ _ _ _ _ t FNP-0-M-011 + CHAPTER 10 DEFINITIONS OF EFFLUENT CONTROL TERMS f e j The terms defined in this chapter are used in the presentation of the above l a chapters. These terms are shown in all capital letters to indicate that they are

epecifically defined.

i: 1 10.1 TERMS SPECIFIC TO THE ODCM The following terms are used in the ODCM, but are not found in the Technical j specifications: 4 BATCH RELEASE A BATCH RELEASE is the discharge of wastes of a discrete volume. Prior to { l sampling for analyses, each liquid batch shall be isolated and then l thoroughly mixed by a method described in the ODCM to assure representative sampling.

                                                                                                                          )

l i l COMPOSITE SAMPLE A COMPOSITE SAMPLE is one which contains material from multiple waste releases, in which the quantity of sample is proportional to the quantity ') D of waste discharged, and in which the method of sampling employed results [V ' in a specimen that is representative of the wastes released. Prior to analyses, all liquid samples that are to be aliquetted for a COMPOSITE' ) SAMPLE shall be mixed thoroughly, in order for the COMPOSITE SAMPLE to be representative of the effluent release. When assessing the consequences of a waste release at the pre-release or post-release stage, the most recent availsble COMPOSITE SAMPLE results for

                                                                                                                         ]

the applicable release pathway may be used.

                                                                                                                         ]

CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of wastes of a non-discrete volume, e.g., from a volume within a system that has an input flow during the continuous release. To be representative of the quantities and concen-tratione of radioactive meterials in CONTINUOUS RELEASES of liquid effluents, samples shall be collected in proportion to the rate of flow of , the affluent stream, or to the quantity of waste discharged. CASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed I to reduca radiamet hra. gaseous affluents by collecting primary coolant system offgasee.from the primary system and providing for delay or holdup 10-1 Gen. Rev. 13

 ~

FNP-0-M-011 (m) for the purpose of reducing the total radioactivity prior to release to the environment. This system consists of at least one gas compressor, waste gas decay tanks, and associated components providing for treatment flow and functional control. LIOUID RADWASTE TREATMENT SYSTEM A LIQUID PADWASTE TREATMENT SYSTEM is any system designed and installed to I reduce radioactive materials in limaid effluents by systematic collection, retention, and processing through filtration, evaporation, separation and/or ion exchange treatment. This system consists of at ' least one collection tank, one evaporator or demineralizer system, one post-treatment tank and associated components providing for treatment flow and functional control. MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS i For the purposes of the ODCM, MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS include the following changes to such systems: I t (1) Major changes in process equipment, components, structures, or effluent monitoring instrumentation as described in the Final

      ,]                   Safety Analysis Report (FSAR) or as evaluated in the Nuclear
 'k)                       Regula*.ory Commission staf f's Safety Evaluation Report (SER) (e.g. ,

deletion of evaporators and installation of demineralizers); (2) Changes in the design of radwasce treatment systems that could  ; significantly increase quantities of effluents released from those previously considered in the FSAR and SER; (3) Changes in system design which may invalidate the accident analysis  ! as described in the SER (e.g. , changes in tank capacity that would l alter the curies released); or (4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g., use of temporary equipment without adequate j shielding provisions). l MEMBER (S) OF THE PUBLIC I A MEKBER OF THE PUBLIC shall be an individual in a controlled area or an UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC

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L,/ l 1 The italicized terms in this definiti.on, which are not otherwise used in  ! this ODCM, shall have the definitions assigned to them by 10 CFR 20.1003. 10-2 Gen. Rev. 13

4 , q. FNP-0-M 011 during any period in which the individual receives an occupational dose. This category may include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant. 4

i MINIMUM DETECTABLE CONCENTRATION i

The MINIMUM DETECTABLE CONCENTRATION (MDC) is defined, for purposes of the controle in this ODCM, as the smallest concentration of radioactive material in a sample that will yield a net count above system background and that will be detected with 95-percent probability, with only 5-percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation, the MDC for a given radionuclide is determined as follows

  ,                  (Reference 12):

2. 71 + 3. 29 Rb

• Cs Cs

                                                                   %        '      Cb'
    }}
    . ss' MDC =

Ea V 2.22 x 106 , y , ,-A & n, .. where MDC = the a priori MINIMUM DETECTABLE CONCENTRATION (yCi per unit mass or volume). 2.71 = the square of the standard normal variate (1.645) for the 95 percent confidence level (Ref. 12, Section II.D). 3.29 = Two times the standard normal variate (1.645) for the 95 parcent confidence level (Ref. 12, Section II.C). Rb= the background counting rate, or the counting rate of a blank aimple, as appropriate (counts per minute), t, = the length of the sample counting period (minutes). tb= the length of the background counting period.(minutes). E= the counting efficiency (counts per disintegration) V= the sample size (units of mass or volume). 2.22 x 100 = the number of disintegrations per minute per C1. Y =- the fractional radiochemical yield, when applicable. 1 =. the radioactive decay constant for the given radionuclide ( h-I) . Values of i used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15. 10-3 Gen. Rev. 13

FNP-G-M-011 At = for effluent samples, the elapsed time between the midpoint of sample collection and the time of counting (h); for environmental samples, the elapsed time between the end of sample collection and the time of counting (h). Typical values of E, V, Y, and At should be used in the calculation. It should be recognized that the MDC 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. PRINCIPAL GAMMA EMITTERS The PRINCIPAL GAMMA EMITTERS for which the MINIMUM DETECTABLE CONCENTRATION (MDC) limit applies include exclusively the following radio-nuclides: e For liquid radioactive effluents: Mn-54, Fe-59, Co- 5 8, Co-60, Zn-65, Mo- 99 , Cs-134, Co-137, and Ce-141. Ce-144 shall also be measured, but with an MDC of 5 x 10-6 yci/mL.

• For gaseous radioactive effluents: In noble gas releases, Kr-87, l

Kr-88, Xe- 13 3, Xe-133m, Xe-135, Xo-138; and in particulate releases, Mn-54, Fe-59, Co-58, co-60, Zn-65, Mo-99, Cs-134, Cs-137, . Ce- 141, and Ce-144. e For environmental media: The gamma emitters specifically listed in Table 4-3. These liste do not mean that only these nuclides are te be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report, the Annual Radiological Environmental Operating Report, or other applicable report (s). SITE BOUNDARY For the purpose of effluent control 1 defined in the ODCM, at- r!TE BOUNDARY shall be as shown in Figure 10-1. MFRESTRICTED AREA The UNRESTRICTED AREA shall be any area access to which is neither 1raiad nor controlled by the licensee or any area.withis. the. SITE.BOUNDIXY ured for residential quarters or for industrial, commercial, instir.utional, and/or recreational purposes. 10-4 Gen. Rev. 13

1

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FNP-0-M-011

   '10.2     TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS

• I Th3 following terms are defined in the Technical Specifications, Section 1.0.

B3cGuse they are used throughout the Limits of Operation sections of the ODCH, th y are presented here for convenience. In the event of discrepancies between , tho definitions below and those in the Technical Specifications, the Technical Specification definitions shall take precedence. ACTIONfS) An ACTION shall be that part of a control that prescribes remedial measures required under designated conditions. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel, such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock, and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps, such that the entire channel is calibrated. E ,; CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comperison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST thall bei e Analog Channels - the injection of a simulated signal into the channel.as close to the sensor as practicable to verify OPERABILITY including alarm, and/or trip functions.

• Bistable Channels - the injection of a simulatad signal into the sensor to verify OPERABILITY including alarm, and/or trip functions.

DOSE EOUIVALENT I-131 , DOSE EQUIVALENT I-131 shall be that concentration of I-131 ( Ci/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-13b I-14 I-133, I-134, and I-135 actually g __..;.. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, 1977. ' 10-5 Gen. Rev. 13

FNP-0-M-Oll } .FREOUENCY NOTATION d The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined below, with a maximum allowable extension not to exceed 25% of the surveillance interval. NOTATION FREOUENCY S (Once per shift) At least once per 12 hours. D (Daily) At least once per 24 hours. W (Weekly) At least once per 7 days. M (Monthly) At least once per 31 days. Q (Quarterly) At least once per 92 days. SA (Semi-annually) At least once per 184 days. R (Refueling) At least once per 18 months. S/U (Startup) Prior to each reactor startup. NA Not applicable. P (Prior) Completed prior to each release. MODE Ior OPERATIONAL MODE) An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant j temperature specified in Section 1.0 of the Technical Specifications.

  ,j     OPERABLE for OPERABILITY)
  ;             OPERABILITY exists when a system, subsystem, train, component or device is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2652 MWt. SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. THERMAL POWER THERMAI POWER shall be the total reactor core heat transfer rate to the reactor coolant. 1J-6 Cen. Rev. 13

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FNP-O-M-011 j VENTILATION EXHAUST TREATMENT SYSTEM The VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulaces from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on any noble gas effluents). Engineered Safety Feature (ESF) atmowpheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. This system consists of the radwaste filtration unit, fuel pool exhaust filtration units and associated components providing for treatment flow and functional control. I A l v, 7 I 10-7 Gen. Rev. 13

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