ML20112D242

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Rev 1 to Limerick Generating Station Offsite Dose Calculation Manual
ML20112D242
Person / Time
Site: Limerick Constellation icon.png
Issue date: 02/28/1985
From: Leitch G, Mulford R, Murphy G
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20112D202 List:
References
PROC-850228-01, NUDOCS 8503220287
Download: ML20112D242 (62)
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5 ~~ PHILADELPHIA ELECTRIC COMPANY PHILADELPHIA -

ATTACI:::E::T D' I s EO P:lILADCLPIIIA CLECTRIC CO:!PANY'S SE: I-A:::!UAL CFTLUC!!T ROLEASCS RCPORT NO. '

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'l PERTAI!;I!!G TO 9 LII!CRICI: CCNCRATING STATIO:: U:IT 1 DCCI;CT !!O.50-35C CI ANGES TO TIIE OFFSITE DOSE CALCULATIO: ::A::UAL (ODC:1)

PE3RUARY, 1985 I

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ATTACHMENT D LIMERICK GENERATING STATION ODCM, REVISION 1 1

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I Attachment D1 Technical Evaluation and Rational for Change Attachment D2 Revision 1 of the ODCM with PORC Approval Attachment D3 Statement on Accuracy and Reliability Attachment D4 Documentation of Engineer-in-Charge, Nuclear and Environmental Section I

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TECHNICAL EVALUATION AND RATIONALE OF CHANGE 1

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E9CCM Og(o PHILADELPHIA ELECTRIC COMPANY ELECTRIC PRODUCTION DEPARTMENT LIMERICK GENERATING STATION I

January 28, 1985 From: G. M. Leitch To: .R. A. Mulford

Subject:

LGS PORC Approved Changes to the ODCM In accordance with LGS Technical Specifications Section 6.5.1.6 m, PORC has reviewed and approved the changes to the ODCM listed below:

1. The containment purge isolation setppint basis has been modified to eliminate the reference to ,

10CFR20, in order to achieve consistency with Tech.

Specs. Section T3.3.2-2C. Tech Specs allows containment isolation on high radiation to occur at 3,2.1 uCi/cc. The ODCM modification will not only allow LGS flexibility in determining the appropriate setpoint for isolation, but also provide a necessary uC1/cc to uCi/sec conversion.

2. The determination of the VFi factor (release point weighting factor) has been modified to weight the release points on dose rate contribution rather than an MPC weighted basis. Both methodologies accomplish the same objective, however, the former I not only allows a more expedient determination of VFi since the dose rate per uCi/cc is inherent in Chemistry's computer based setpoint methodology, I but also provides a more even release point weighting. (The latter methodology greatly penalized the South vent).

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I Mr. R. A. Mulford Page 2 January 28, 1985 I 3. The methodology for determining the alarm setpoints from the " previous month's data" was changed to I- have the alarm setpoints based on the "first scheduled sample of the month". This change more accurately reflects Chemistry's program and the Surveillance Tests which schedule the samples.

4. Setpoints have been determined for the Hot I Maintenance Shop. The " worst case" assumptions used in the determination result in setpoints which are highly conservative. However, releases from the Hot Maintenance Shop are anticipated to be I small by comparison and the need to consider this release point in the VFi determination has been eliminated.

Per Tech Spec Section 6.14.2-b, "these changes shall become effective upon review and acceptance by the Engineer-I in-charge, Nuclear and Environmental Section". We request you review the attached changes and submit your reply to LGS as soon as possible in order that we provide these changes to the NRC with the Semi-Annual Radioactive Effluent Release I Report.

2210.

Please address any questions to G. W. Murphy, Ext.

Thank you.

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. M. Leitch h Superintenden - LGS GWM/ nip Attachment cc J. Doering I P. J. Duca J. B. Cotton R. W. Dubiel J. S. Wiley I File 14-1 I

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ATTACHMENT D2

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I REVISION 1 OF THE ODCM WITH PORC APPROVAL

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I I OFFSITE DOSE CALCULATION MANUAL Revision 1 I l LIMERICK GENERATING STATION UNITS 1 AND 2 I

I PIIILADELPIIIA ELECTRIC COMPANY I Docket Nos. 50-352 & 50-353 PORC Approval:

Sta' tion Wperintendent I

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LGS Ilealth Physics Representative: /

NuclearandEnvironment[al Representative: [ JVM I

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4 Table of Contents c

Page I. Purpose II. Liquid Pathway Dose Calculations I A.

B.

Surveillance Requirement 4.11.1.1.2 Surveillance Requirement 4.11.1.2 1

2 C. Surveillance Requirement 4.11.1.3.1 3 III. Gaseous Pathway Dose Calculations A. Surveillance Requirement 4.11.2.1.1 5 B. Surveillance Requirement 4.11.2.2 8 C. Surveillance Requirement 4.11.2.3 11 D. Surveillance Requirement 4.11.2.5.1 12 IV. Nuclear Fuel Cycle Dose Assessment - 40 CFR 190 A. Surveillance Requirement 4.11.4.1 14 B. Surveillance Requirement 4.11.4.2 14 V. Calendar Year Dose Calculations A. Unique Reporting Requirement 6.9.1.12 15 VI. Radiological Environmental Monitoring Program A. Surveillance Requirement 4.12.1 22 VII. Effluent Radiation Monitor Setpoints 24 VIII. Bases 34 IX. Liquid and Gaseous Effluent Flow Diagrams 39 4

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I DATE:J/ar/tra HP: 4 ENGR: AdC I

I. Purpose The purpose of the Offsite Dose Calculation Manual is to establish methodologies and procedures for calculating doses to individuals in areas at and beyond the SITE I BOUNDARY due to radioactive effluent from Limerick Generating Station and establishing setpoints for radioactive effluent monitoring instrumentation. The i

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results of these calculations are required to determine Il compliance with Appendix A to Operating License NPF-27,

" Technical Specification and Bases for Limerick Generating Station Units No. 1 and 2.

II. Liquid Pathway Dose Calculations I A. Surveillance Requirement 4.11.1.1.2 - Liquid Radwaste Release Compliance with 10CFR20 Limits I Limerick Generating Station Units 1 and 2 have one common discharge point for liquid releases under normal circumstances. In the event of heat exchanger leakage, additional release pathways are possible I through the plant service water system and the RHR service water system. The following calculation assures that the radwaste release limits are met.

The flow rate of liquid radwaste released from the site to areas at and beyond the SITE BOUNDARY shall be I such that the concentration of radioactive material after dilution shall be limited to the concentration specified in 10 CFR 20.106(a) for radionuclides other than the dissolved or entrained noble gases and the I concentration listed on Technical Specification Table 3.11.1.1-1 for all dissolved or entrained noble gases as specified in Technical Specification 3.11.1.1.

I Each tank of radioactive waste is sampled prior to release and is quantitatively analyzed for identifiable gamma emitters as specified in Table I

4.11-1 of the Technical Specification. From this gamma isotopic analysis the maximum permissible release flow rate is determined as follows:

Determine a Dilution Factor by:

Dilution Factor = uCi/ml i

{i MPCi uCi/ml i = the activity of each identified gamma emitter in uC1/ml I .

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I MPCi = The MPC specified in 10 CFR 20, Appendix B, I Table II, Column 2 for radionuclides other than dissolved or entrained noble gases or the concentrations listed on Technical Specification Table 3.11.1.1-1 for dissolved I I or entrained noble gases. Any unidentified concentration is assigned an MPC value of lE-07 uCi/ml.

Determine the Maximum Permissible Release Rate with this Dilution Factor by:

Release Rate (gpm) = A B X Dilution Factor I A = The cooling tower blowdown volume which will provide dilution. Maximum flow rate is 10,000 gpm.

B = margin of assurance which includes consideration of the maximum error in the activity setpoint and the maximum error in the flow setpoint and the possibility of multiple release pathways.

B. Surveillance Requirement 4.11.1.2 The primary method of calculating dose contributions from liquid effluents released to areas at or beyond the SITE COUNDARY will be by using a computer-based I calculatior.al program developed using the equations and paramete.s of R.G. 1.109, Rev. 1, October, 1977 (see bases NoNe 4) for all organs and age groups. The Ai values used for this calculation are located in the Appendix, Table 1..

I Dose contributions

  • rom liquid effluents released to areas at and beyond the SITE BOUNDARY shall be calculated using the equation below. This dose calculation uses as a minimum those appropriate I radionuclides listed in Table II.A.l. These radionuclides account for virtually 100 percent of the total body dose and bone dose from liquid effluents.

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Dq = k Rik i Ai i=1 E d tl (Cil) Fi ~

where:

I D { = the cumulative body dose Tcommitment or any organ, , from liquid for the total time period to the total effluents in mrem Zm i=1 gtl I

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Ri = reported release points d t1 = the length of the lth time period over which Cil and F1 are averaged for tne liquid release, in hours.

Cil = the average concentration of radionuclide, i, in undiluted liquid effluent during time period I t from any liquid release, (determined by the effluent sampling analysis program, Technical Specification Table 4.11.1.1-1), in uCi/ml.

A1 = the site related ingestion dose commitment factor to the total body or organ, , for each radionuclide listed in Table I .A.1, in mrem-m1 per hr-uC1. See Site Specific Data.**

Fl = the near field average dilution factor for I Cil during any liquid effluem release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to the average flow from the discharge structure to the I II.C Schuylkill River.

Surveillance Requirement 4.11.1.3.1 I Projected dose contributions from liquid effluents shall be calculated using the methodology described in Section II.B.

To estimate expected concentration of the various radionuclides (Cil) in the undiluted liquid effluent, the duration of liquid release (4t), and the near field average dilution factor (F1), the expected plant operating status shall be reviewed. If no operational changes are expected which would af fect Cil,6 t, or F1 the same values as used to evaluate Section II.B may be used.

If any operational changes are expected during the following 31 days which could affect Cil,6t or F1, the values used shall be based on plant history. During the initial stages of plant operation, the values for Cil,4 t, and ?1 as given in LGS FSAR Section 11.2 and EROL Section 5.2 may be used.

I ** See Note 1 in Bases I

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TABLE II.A.1 LIQUID EFFLUENT INGESTION DOSE FACTORS (Decay Corrected)

Ai Dose Factor (mrem-ml per hr-uci)

Radionuclide Total Body Bone Cs-137 3.42E+05 3.82E+05 Cs-134 5.79E+05 2.98E+05 P-32 5.llE+04 2.05E+05 I Cs-136 Zn-65 Sr-90 8.42E+04 3.32E+04 1.35E+05 2.97E+04 2.31E+04 5.52E+05 I H-3 Na-24 I-131 3.29E-01 1.35E+02 1.16E+02 1.35E+02 1.40E+02 Co-60 5.70E+02

  • I-133 1.23E+01 2.31E-01 Fe-55 1.06E+02 6.61E+02 Sr-89 6.36E+02 2.21E+04 I Te-129m Mn-54 Co-58 1.70E+03 8.34E+02 2.00E+02 1.08E+04 8.34E+02 I Fe-59 Te-131m Ba-140 9.26E+02 3.88E+02 1.33E+01 1.02E+03 9.53E+02 2.03E+02 Te-132 1.21E+03 1.99E+03 I NOTE: The listed dose factors are for radionuclides that may be I detected in liquid effluents and have significant dose consequences. These factors are decayed for one day to account for the time between effluent release and ingestion of fish by the maximum exposed individual, an adult.
  • There is no bone dose factor given in R.G. 1.109 for these nuclides.

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I III. Gaseous Pathway Dose Calculations I The controling receptor locations for the gaseous pathway dose calculations are based on a land-use census performed in 1975 to 1976 which has been periodically updated. The most recent update was in 1983.

A. Surveillance Requirement 4.11.2.1.1 The dose rate in areas at and beyond the SITE BOUNDARY due to radioactive materials released in gaseous effluents shall be determined by the expressions below:

1. Noble Gases The dose rate from radioactive noble gas releases shall be determined by either of two methods. Method (a), the Isotopic Analysis Method, utilizes the I results of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2. Method (b),

the Gross Release Method, assumes that all noble gases I released are the most limiting nuclide-Kr-88 for total body dose and Kr-87 for skin dose.

.g For normal operations, it is expected that method (a)

E wi 1 be " Sed- However' if iSotoP io rele Se d^t^ ^re not available method (b) can be used. Method (a) allows more operating flexibility by using data that more accurately reflect actual releases.

a. Isotopic Analysis Method DTD =

2 i(Ki (X/Q)v Qiv)

Ds =ki (Li + 1.lMi) (X/Q)vj I where:

I The location is the site boundary, 790m NE from the vents. This location results in the highest calculated dose to an individual from noble gas releases.

DTB = total body dose rate, in mrem /yr.

Ds = skin dose, in mrem /yr.

Ki = the total body dose factor due to gamma I emissions for each identified noble gas radionuclide. Values are listed on Table III.A.1 and are taken from R.G. 1.109, I in mrem /yr per uCi/m3.

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I (X/0)v = 1.lE-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all I_ vent releases (NE boundary).

I Qiv = the release rate of noble gas radionuclide, i, in gaseous effluents from all vent releases determined by isotopic analysis averaged over one hour, in uCi/sec.

Li = the skin dose factor due to beta emissions for each identified noble gas radionuclide.

I Values are listed on Table III.A.1 and are taken from R.G. 1.109, in mrem /yr per uCi/m3.

Mi I the air dose factor due to gamma emissions

=

for each identified noble gas radionuclide.

Values are listed on Table III.A.1 and are taken from R.G. 1.109, in mrad /yr per uCi/m3.

1.1 = unit conversion, converts air dose to skin dose, mrem / mrad.

b. Gross Release Method I DTB = K (X/Q)V hnv Ds =

(L + 1.lM) (X/0)hnv where:

The location is the site boundary, 790m NE from the vents. This location results in the highest calculated dose to an individual from noble gas I releases.

DTB = total body dose rate, in mrem /yr.

Ds = skin dose rate, in mrem yr.

l I K = 1.47E04 mrem /yr per uCi/m3; the total body dose factor due to gamma emissions for Kr-88 (Reg. Guide 1.109, Table B-1).

(X/Q)v = 1.lE-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary). ,

(b)nv=

I the gross release rate of noble gases in gaseous effluents from vent releases determined lI 6

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I by gross activity vent monitors averaged over one hour, in uCi/sec.

L = 9.73E03 mrem /yr per Ci/m3; the skin dose factor due to beta emissions for Kr-87 (Reg. I Guide 1.109, Table B-1).

M = 6.17E03 mrad /yr per uCi/m3; the air dose factor due to gamma emissions for Kr-87 (Reg. Guide 1.109, Table B-1).

2. The primary method of calculating dose contribution from Iodine-131, Iodine-133, tritium, and radioactive material in particulate form, other than noble gases, with half-lives greater than eight days will be by using a computer-based calculational program developed I using the equations and parameters of R.G. 1.109, Rev.

1, October, 1977 (see bases Note 4) for all organs and age groups.

If the computer model is not available, the dose contributions from Iodine-131, Iodine-133, tritium, I and radioactive materials in particulate form, other than noble gases, with half-lives greater than eight days will be calculated using the equation below:

I where:

DT =

(CF)i Pi WV (Qiv) .

6 The location is the site boundary, 762m ESE from the vents.

DT = dose rate to the thyroid, in mrem /yr.

CF = 1.02; the correction factor accounting for I the use of iodine-131 and iodine-133 in lieu of all radionuclides released in gaseous effluents.

P = 1.62E07 mrem /yr per uC1/m3; the inhalation I-131 dose parameter for I-131 inhalation pathway.

I The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, child. All values are from Reg.

Guide 1.109 (Tables E-5 and E-9).**

P = 3.85E06 mrem /yr per uC1/m3; the inhalation I-133 dose parameter for I-133 inhalation pathway.

I The dose factor is based on the critical .

individual organ, thyroid, and most restrictive age group, child. All values are from Reg.

Guide 1.109 (Tables E-5 and E-9).**

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I Wv = 1.00E-05 sec/m3; the highest calculated I annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).

I hiv = the release rate of iodine-131 and/or I iodine-133 in gaseous effluents from all vent releases, determined by the effluent sampling and analysis program (Technical Specification Table 4.8.2) in I \

uCi/sec.

III.B Surveillance Requirement 4.11.2.2 The air dose in areas at and beyond the SITE BOUNDARY due to noble gases released in gaseous effluents shall be determined by the expressions below.

The dose rate from radioactive noble gas raleases shall be determined by either of two methods. Method (a), the Isotopic Analysis Method, utilizes the results of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2, Method (b), the Gross Release Method, assumes I that all noble gases released are the most limiting nuclide

- Kr-88 for total body dose and Kr-87 for skin dose.

I For normal operations, it is expected that method (a) will be used. However, if isotopic release data are not available method (b) can be used. Method (a) allows more operating flexibility by using data that more accurately reflects actual releases.

I ** See Note 2 in Bases I 1. for gamma radiation l

I a) Isotopic Analysis Method 7

Dy = 3.17E-08 Mi(X/Q)v Qiv where:

The location is the SITE BOUNDARY, 762m ESE from the vents. This location results in the highest calculated gamma air dose from noble gas releases.

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whero:

Dy = gamma air dose, in mrad.

3.17E-08= years per second.

Mi = the air dose factor due to gamma emissions for each identified noble gas radionuclide.

Values are listed on Table III.A.1 and are taken from R.G. 1.109 in mrad /yr per uCi/m3.

(X/Q)v = 1.1E-05 sec/m3; the highest calculated I average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

Qiv = the release of noble gas radionuclides, i, in gaseous effluents from all vents as determined by isotopic analysis, in uC1.

I Releases shall be cumulative over the calendar quarter or year, as appropriate.

b. Gross Release Method D f = 3.17E-08 M (X/Q)v Qv where:

The location is the SITE BOUNDARY 790m NE from the I vents. This location results in the highest calculated gamma air dose from noble gas releases.

D 'd I = gamma air dose, in mrad.

3.17E-08= years per second.

I M = 1.52E04 mrad /yr per uCi/m3; the air dose factor due to gamma emissions for Kr-88 (Reg. Guide 1.109, Table B-1).

(X/Q)v = 1.lE-05 sec/m3; the highest calculated annual average relative concentration

!E from vent releases for any area at or g beyond the SITE BOUNDARY.

! Qv = the gross release of noble gas radio-l nuclides in gaseous effluents from all vents, determined by gross activity vent monitors, in uCl. Releases shall be

. cumulative over the calendar quarter or .

lE5 year as appropriat.e.

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2. for beta radiation
a. Isotopic Analysis D3 ,

= 3.17E-08 Ni (X/Q)v Qiv where:

The location is the SITE BOUNDARY 790m NE from the I vents. This location is the highest calculated gamma air dose from noble gas releases.

3.17E-08 = years per second.

Ni = the air dose factor due to beta emissions for each identified noble gas radionuclide.

I Values are listed on Table III.A.1 and are taken from Reg. Guide 1.109, in mead /yr per uCi/m3.

(X/0)v = 1.lE-05 sec/m3; the highest calculated annual average relati"e concentration from I vent releases for any area at or beyond the SITE BOUNDARY.

Qiv = the release of noble gas radionuclide, i, in gaseous effluents from all vents as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

b. Gross Release Method D = 3.17E-08 N (X/Q)v Qv where:

The location is the SITE BOUNDARY 790m NE from the vents. This location results in the highest calculated gamma air dose from noble gas releases.

Dp = beta air dose, in mrad.

3.17E-08 = years per second.

N = 1.03E04 mrad /yr per uCi/m3; the air dose I factor due to beta emissions for Kr-87 (Reg.

Guide 1.109, Table B-1).

I (X/Q)v = 1.1E-05 sec/m3; the highest calculated annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

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c Qv = the gross release of noble gas radionuclides in gaseous effluents from all vents determined by gross activity vent monitors, in uCi. Releases shall be cumulative over the calendar quarter or year, as appropriate.

III.C Surveillance Requirement 4.11.2.3 The primary method of calculating dose to an individual from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form, other than noble gases, with half-lives greater than eight days in gaseous effluents I released to areas at and beyond the SITE BOUNDARY, will be by using a computer-based calculational program developed using the equations and parameters of R.G. 1.109, Rev. 1, October, 1977 (see based Note 4) for all organs and age I groups.

If the computer model is not available, the following expression will be used:

D = 3.17E-08 (CF) (0.5) 2 Ri (WV\Qivj I

where:

I Location is the critical pathway dairy 1770m ESE from vents.

D = critical organ dose, thyroid, from all pathways, in mrem.

3.17E-08 = years per second.

CF = 1.00; the correction factor accounting for the use of Iodine-131 and Iodine-133 in lieu of all radionuclides released in gaseous effluents.

0.5 = fraction of iodine releases which are nonelemental.

R = (9.51E+11)/m2 (mrem /yr) per (uCi/sec); the dose I-131 factor for Iodine-131. The dose factor is based I on the critical individual organ, thryoid, and most restrictive age group, infant. See Site Specific Data.**

R =

(8.13E+09)/m2 (mrem /yr) per uCi/sec; the dose I-133 factor for Iodine-133. The dose factor is based I on the critical individual organ, thyroid, and most restrictive age group, infant. See Site Specific Data.**

Wv =

( 1. 82E-09) /m2 (D/Q) for the food ll I

I pathway for vent releases.

I Qiv

= the release of Iodine-131 and/or Iodine-133 determined by the effluent sampling and analypis program (Technical Specification Table 4.11.1.1.2-1)

I in uCin Releases shall be cumulative over the calendar quarter or' yea,r, as appropriate.

III.D Surveillance Requirement 4.11.2.5.1 '

The projected doses from releases of gaseous effluents to '

areas at and beyond the SITE BOUNDARY shall be calculated in accordance with'the following sections of this manual: '

a. gamma air dose - III.B.E'
b. beta air dose - III.B.2
c. organ dose - III.C The projected dose calculation shall be based on expected releases from plant operation. The normal release pathways I result in the maximum releases from the plant. Any alternative release pathways result in lowed releases and7 therefore lower doses.

l To estimate the e::pected releases of noble ~ gases and radiciodines in gaseous effluents, the expected plant operating status shall be reviewed. If no operational I changes are expected which would affect the magnitude or type of releases the same values used to evaluate Sections III.B.1, III.B.2 and III.C may oe used.

If any operational changes are expected during the '

following 31 days which could affect the magnitude or type of releases, the values used shall be based on plant history. Durin.g the initial stages of plant operation the j values for releases expected as given in LGS FSAR Section 11.3 may be used.

.I ** See Note 3 in Bases l

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TABLE III.A.1 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES Nuclide B-air *(Ni) B-Skin **(Li) 8-Air *(Mi) [ Body **(Ki)

Kr-83m 2.88E-04 ---

1.93E-05 7.56E-08 Kr-85m 1.97E-03 1.46E-03 1.23E-03 1.17E-03 Kr-85 1.95E-03 1.34E-03 1.72E-05 1.61E-05 Kr-87 1.03E-02 9.73E-03 6.17E-03 5.92E-03 Kr-88 2.93E-03 2.37E-03 1.52E-02 1.47E-02 Kr-89 1.06E-02 1.01E-02 1.73E-02 1.66E-02 Kr-90 7.83E-03 7.29E-03 1.63E-02 1.56E-02 Xe-131m 1.11E-03 4.76E-04 1.56E-04 9.15E-05 Xe-133m 1.48E-03 9.94E-04 3.27E-04 2.51E-04 Xe-133 1.05E-03 3.06E-04 3.53E-04 2.94E-04 Xe-135m 7.39E-04 7.11E-04 3.36E-03 3.12E-03 Xe-135 2.46E-03 1.86E-03 1.92E-03 1.81E-03 Xe-137 1.27E-02 1.22E-02 1.51E-03 1.42E-03 Xe-138 4.75E-03 4.13E-03 9.21E-03 8.83E-03 Ar-41 3.28E-03 2.69E-03 9.30E-03 8.84E-03

  • mrad-m3 pCi yr
    • mrem-m3 pCi-yr

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REFNRENCE: Regulatory Guide 1.109, Revision 1, October 1977 13

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IV. TOTAL DOSE A. Surveillance Requirement 4.11.4.1 If the doses as calculated by the equations in this I manual do not exceed the limits given in Technical Specifications 3.11.1.2.a, 3.ll.2.b, 3.ll.2.a, 3.ll.2.2.b, 3.ll.2.3.a, or 3.11.2.3.b by more than two times, the conditions of Technical Specification I 3.11.4.2 have been met.

B. Surveillance Requirement 4.11.4.2 If the doses as calculated by the equations in this manual exceed the limits given in Technical I Specifications 3.ll.l.2.a, 3.ll.l.2.b, 3.11.2.2.a, 3.ll.2.2.b, 3.ll.2.3.a, or 3.ll.2.3.b by more than two times, the maximum dose or dose commitment to a real individual shall be determined utilizing the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactot Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977. Any deviations from the methodology provided in Regulatory Guide 1.109 shall be documented in the Special Report to be prepared in I accordance with Technical Specification 3.11.4.1.

The cumulative dose contribution from direct radiation from the two reactors at .the site and from radwaste storage shall be determined by the following methods:

I Cumulative dose contribution from direct radiation =

Total dose at the site of interest (as evaluated by TLD measurement) -

Mean of background dose (as evaluated by TLD's at background sites) -

Effluent contribution to dose (as evaluated above).

The method provided in the second paragraph above is I used only to evaluate the contribution from direct radiation dose. The direct radiation dose is then added to the dose or dose commitment determined in accordance with the methods in the first paragraph above to determine total dose from all pathways.

This evaluation is in accordance with ANSI /ANS 6.6.1-I 1s79 Section 7. The error using this method is estimated to be approximately 8%.

I I 14

I V.A Unique Reporting Requirement (6.9.1.12) - Dose Calculations for the Radioactive Effluent Release Report The assessment of radiation doses for the radiation dose assessment report shall be performed utilizing the I methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, I October 1977. Any deviations from the methodology provided in Regulatory Guide 1.109 shall be documented in the radiation dose assessment report.

The meteorological conditions concurrent with the time of release of radioactive materials (as determined by sampling I frequency of measurement) or approximate methods shall be used as input to the dose model, g The Radioactive Effluent Release Report shall be submitted g within 60 days after January 1 of each year.

VI.A Surveillance Requirement 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table VI.A.1 from the locations shown on Figures VI.A.1, VI.A.2 and VI.A.3 and shall be analyzed I pursuant to the requirements of Table 3.12-1 of the LGS Technical Specifications.

VII.A Surveillance Requirement 4.12.3 Pursuant to Section 4.12.3 of the LGS Technical Specifications, the laboratory performing the radiological I environmental analyses shall participate in an interlaboratory comparison program which has been approved I by the NRC. This program is the Environmental Protection Agency's (EPA's) Environmental Laboratory Intercomparison Studies (cross check) Program. Our participation code is CJ. Participation includes all of the determinations (sample medium-radionuclide combination) that are offered I by the EPA and that are also included in the monitoring program. The results of the analysis of these (cross I check) samples will be included in the Annual Radiological Environmental Operating Report.

E E

I m

m m W W W W W W W W W m W W m W m a m.

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TABLE VI.A.1 RADIOLOGICAL ENVIRONMENTAL MONITORIMS PROGRAM 1

EXF43URE PATMWAT MUMBER OF S AMPLES AND STATION STATION DISTAMCE S MD/CR S AMPLE SAMPLE STATIOM MAME CODE SECTOR (MILE 5) COMMENTS P

Direct 40 LOCATIONS (a) TLD sites were ehesen in assordance Radiation (a) IMMER RIMG LOCATIONS with Limerick Generating Station's 13 Evergreen E Senatoga Road 3651 M 0.6 Technical specifications Table 3.12-7 23 Sanatoga Road 351 MNE 0.6 Item 1. The inner ring and outer 33 Possus Mellow Road 551 NE 0.4 ring stations cover all sectors.

43 LG5 Training Center 751 EME 0.5 53 Keen Road 1953 E 0.5 The control'and special interest 63 LGS Information Center 1151 Est 0.5 stations provide information on 73 Longview Road. SE Sector 1951 SE 0.6 population centers and other special site Boundary interest locations.

93 Longview Road SSE Sector 1652 SSE 0.6 Site Boundary 93 Railroad Track Along 1951 5 0.3 Longview Road

$_ 103 Impounding Basin SSN 2151 SSN 0.5 sector Site Boundary 111 Transmission Tower. SW ~ 2332 SM 0.5 '

sector Site Boundary 123 WSW Sector. Site Boundary 1551 WSW 0.5 133 Meteorological Tever 2. Site 2653 W 0.4 143 WNW 5ector site Boundary 1951 WNW 0.5 153 MW Sector Site Boundary 3251 MW 0.6 163 Meteorological Tower 1 Site 3452 MMW 0.6 CUTER RIMS LOCATIOMs 13 Ringing RocA Substation 35y1 M 4.2 23 Laughing Waters Gsc 2E1 MME 5.1 33 Meiffer Road 4E1 NE 4.6 43 Pheasant Road Game rara .7El EME 4.1 Site 33 Transmission corrider. 10E1 E 3.9 63 Trappe substation 10r3 ESE 5.5 73 Vaughn Substation 13E1 SE 4.3 ,

83 Fikeland substation 16rt SSE 4.9 93 Shouden substation 19D1 5 3.6 103 sheeder substation 20rt 55W 5.1 113 Porter's Mill substation 24D1 SW 3.9 123 Transmission Corrider. 25D1 WSW  %.0 Moffecker and Keim streets 133 Tr.ansmission corrider. 20D2 W 3.0 W. Cedarville Road 193 Prince Street 29El WMW 4.9 153 Poplar Substation 31D2 MW 3.9 163 Yarueli Road 34E1 MMW  %.6

m M M M M OM b M 'M M M MM E. .

CONTROL AND SPECIAL INTEREST -

) 10CATICMS 25.8 13 stroh Substation (control) 5N1 NE EME 2.1 23 Fettstown Landing Field SC1

)

3) Reed Reed 9C1 E 1.1
4) Einy Road 13C1 SE 2.9
5) Spring City Substation . 1591 SE 3.2 3 , 6) Linfield Substation 17s1 3 1.6 73 Ellis Woods Road test SSE 3.1
8) 11nooln Substation 3131 NW 3.9 .

) 5 10 CATIONS 1053 E 9.5 (b) These stations. provide for coverage

2. Airborne 13 Keen Road of the highest annual ground level

) 2) 105 Infora~a tion Center Radiciodine and 3) Zongview Road 1951 1431, ESE SE 0.5 e.6 B/t. and a control location. Radio-13C1 SE 2.9 iodine cartridges which have been Ferticulates 4) Eing Road SE 28.8 tested for perforsance by the I (b) 5) 2301 Market Street. 13M4 manufacturer are used at all times Philadelphia. FA (control) te) All surface and drinking stations have

) 3. Wat'orborne (c) 9 LOCATIONS - continuous samplers.

Surface' 1) 11serick Intake (control) 2451 WSW 0.3 J - il 11nfield Brid5e -

1632 SSE 1.1 .

Ground 13 103 Information Centet 1131 ESE 0.5 23 South Sector Tara Near Site 18A1 -S 1.0 l'~ ' Drinking 13 Phoenizville Water Works '

23 Fettstown Water Authority' 15F7

'28F3 SSE NMW 5.2 5.9 * *

~

(control) 3 33 Philadelphie Suburban Water 15F4 ,SSE 7.8 Company . .?. ,,

4) Citisans Home Water Company 16C2 SSE 2.4

) Sodiment*From 13 Vincent Den Pool Area 16C4 3 1.9 .

Shoreline .

s 4. Ingestion S LOCATIONS 22F1 (d) Milk saa'ples are taken free several

") Milk (d) 13 Control Station

  • farms surrounding LOS. These farms I 2) - . 3C1 1031 include those with the highest dose 33 2551 potential from which samples are 3 41 routinely availabid. as well as a control station. The locations of the

)* farms is not listed herein due to 4 longstanding agreement with the farmers involved. In return for being allowed

) to sample and analFae the milk. FEco has agreed not to divulge the location of the farms. .

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33 speteen of tes. Esim sizeot 2fCt unu 3.3 Bridge to Renetet street teidge (contre 13 9.5 f f 7 reos produete are to be esoples es Feed Fredeste 13 Les Information Centet list tst y:?t of the Les Technteet speelf t.

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  • l 1

I l VII. Effluent Radiation Monitor Setpoint Calculations A. Liquid Effluents

1. Radwaste Discharge Line Radiation Monitor -

Monitor alarm setpoints will be determined in order to assure compliance with 10CFR20. The setpoints will indicate if the concentration of I radionuclides in the liquid effluent at the site boundary is approaching the concentrations specified in 10CFR20, Appendix B, Table II, l

Column 2 for radionuclides other than dissolved 8 or entrained noble gases. The setpoints will also assure that a concentrations listed on Technical Specification Table 3.11.1.1-1 for dissolved or entrained noble gases is not i exceeded. The following method applies to liquid releases from the plant via the cooling tower i blowdown line when determining the high-high alarm setpoint for the Liquid Radwaste Effluent Monitor during all operational conditions. When the high-high alarm setpoint is reached or exceeded, the releases will be automatically 8 terminated.

I a. The setpoint for the Liquid Radwaste Effluent monitor will be calculated as follows:

1) Determine Ct Ct = 1 CiD (Fi)

SZ(Ci/MPCi) where:

Ct = concentration at the liquid radwaste discharge line monitor (prior to dilution to assure 10CFR20.106 limits are not exceeded; uCi/cc 2 Ci = total concentration of liquid effluent discharge prior to dilution with cooling tower blowdown; uCi/cc 5 = margin of safety factor including F uncertainty, I to assure that the high-high alarm will terminate the discharge before 10CFR20 limits are exceeded.

T Ci = sum of the ratio of the isotopic concentrations I 4 MPCi divided by their respective MPC. .

I I

m

s

.. ~

I D = dilution factor due to blowdown from the cooling tower; calculated by dividing the total flow 8 (cooling tower blowdown plus radwaste discharge flow) by the radwaste discharge flow.

Fi = Ratio of MPC-weighted releases in the liquid

.radwaste effluent monitor flow path divided by the total MPC-weighted liquid releases; l e.g. Ci I release of flow path of interest MPCi ,

Ci k MPCi all release flow paths

2) Determine C.R.

C.R. =C_t_

E where:

C.R. = the calculated monitor count rate above background attributable to the radionuclides; CPS E = the detection efficiency of the monitor; uCi/cc/ cps.

3) The monitor high-high alarm setpoint above background should be set at the I ,

b.

C.R. value, The monitor high-high alarm setpoint will be calculated monthly. The calculation will be I based on isotopes detected in the liquid radwaste sample tanks during the previous I month. If there were no isotopes detected during the previous month then the annual average concentrations (EROL Table 3.5-3) of those isotopes listed in Table II.A.1 will be used to determine the setpoint.

I If the calculated setpoint is less than the I existing monitor setpoint, the setpoint will be reduced to the new value. If the calculated setpoint is greater than the 3' existing monitor setpoint, the setpoint may

.g remain at the lower value or increased to ,

the new value.

2. Plant Service Water Monitor - Monitor alarm

= setpoint will be determined in order to be able I

23 I .

I to identify and rectify any potential problem due to excessive leakage of heat exchangers. This I setpoint results in concentrations at the site boundary far below 10CFR20, Appendix B, Table II limits. The service water side of the fuel pool I heat exchangers is kept at higher pressure than the shell side to prevent potential radioactive contamination of the service water.

a. The setpoint for the Plant Service Water monitor will be calculated as follows:
1) Determine C.R.s C.R.s = Z (CRb)

I C.R.s = the calculated monitor setpoint count rate attributable to system leakage plus background; CPM Z = multiplier to establish monitor setpoint I count rate above background count rate C.R.b = monitor count rate attributable to background I radiation; CPM b.

I The monitor high alarm setpoint will be calculated monthly. The calculation will be based on the background count rate during the previous month. If the calculated I setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value. If the calculated setpoint is greater than the existing setpoint, the

,- I setpoint may remain at the lower value or I increased at the new value.

3. RHR Service Water Monitor - Monitor alarm l setpoints will be determined in order to be able to identify and rectify any potential problem due to excessive leakage of heat exchangers. This setpoint results in concentrations at the site boundary far below 10CFR20, Appendix B, Table II limits. The following method applies to liquid

'I releases from the plant to the spray pond when determining the high-high alarm setpoint for the l

RHR Service Water Monitor during all operational conditions. When the high-high alarm setpoint is .

reached or exceeded, the releases will be

! automatically terminated.

'I t

i I

I a. The setpoint for the RHR Service Water monitor will be calculated as follows:

1) Determine C.R.

C.R.s = Z x C.R.b where:

C.R.s = the calculated monitor count rate above background attributable to system leakage plus background; CPM Z = multiplier to establish monitor setpoint count rate above background count rate.

C.R.b = monitor count rate attributable to background radiation; CPM I E = the detection efficiency of the monitor; uCi/cc/ CPM.

I 3) The monitor high-high alarm setpoint above background should be set at the C.R. value.

b. The monitor high-high alarm setpoint will be calculated monthly. The calculation will be based on the background count rate during I the previous month. If the calculated setpoint is less than the existing monitor setpoint, the setpoint will be' reduced to the new value. If the calculated setpoint I is greater than the existing monitor setpoint, the setpoint may remain at the lower value or increased to the new value.

B. Gaseous Effluents I 1. North and South Stack Vent Radiation Monitors -

Monitor alarm setpoints will be determined in order to assure compliance with 10CFR20. The setpoints will indicate if the dose rate at or beyond the site boundary due to radionuclides in I the gaseous effluent released from the site is approaching 500 mrem /yr to the whole body and I 3000 mrem /yr to the skin from noble gases, or 1500 mrem /yr to the thyroid from I-131 and I-133 (inhalation pathway only). The alarm setpoint I for the gaseous effluent radiation monitors will be calculated as fo.llows: .

a. North and South Stack Vent Noble Gas Channel
1) Determine Ct I

I .

. 1 I l Ct = 2.12E-03 Ot e

I where:

)

Ct = the concentration at the vent noble gas radiation ,

monitor which indicates that the 10CFR20 dose I rate limit at the site boundary has been reached; uCi/cc 2.12E-03 = unit conversion factor to convert uCi/sec/CFM to uCi/cc.

Qt = the total release rate of all noble gas radio-nuclides in the gaseous effluent (uCi/sec) based I on the lower of either the whole body exposure limit (500 mrem /yr) or the skin exposure I F (3000 mrem /yr) Ot will be calculated as shown in Attachment 1.

= anticipated maximum vent flow rate; CFM

2) Determine the noble. gas channel alarm setpoint (Sn)

Sn = VFi Ct where:

I VFi = fractional contribution to site boundary dose rate from the release point of interest; I i.e. noble gas dose rate contribution from North Vent divided by the total noble gas dose rate contributien from the North and South Vents.

Normally the VFi values will be l

determined on a monthly basis but may be performed more often in response to plant conditions.

I

'I lE -

I I

i 26 I

b. North and South Stack Vent Iodine Channel
1) Determine Ct Ct = 2.12E-03 Ot e

I where:

Ct = the concentration at the vent iodine radiation monitor which indicates that the 10CFR20 dose rate liinit at the site boundary has been reached; I uCi/cc.

2.12E-03 = unit converstion factor to convert uCi/sec/CFM to uCi/cc.

Qt = the total release rate of radiciodines in the gaseous effluents (uCi/sec) Qt will be ,

I calculated as shown in Attachment 1.

F = maximum antcipated vent flow; CFM.

2) Determine the iodine channel alarm setpoint (Si)

I Si = VFi Ct where:

VFi = fractional contribution to site boundary dose rate from the release point of interest; i.e. iodine dose rate I contribution from North Vent divided by total iodine dose rate contribution from the North and South Vents.

Normally, the VFi values will be determined on a monthly basis but may be performed more often in I response to plant conditions.

2. The monitor alarm setpoints will be calculated I monthly. These routine calculations will be based on isotopic analysis of the first scheduled sample of the month. The monitor alarm setpoint I calculations may be performed more often in response to plant conditions. If there were no isotopes detected in the sample then isotopic concentrations calculated from the expected I annual average noble gas and iodine-131 and 133 .

isotopic release rates (EROL Table 3.5-6) will be used to determine the setpoint. If any I calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to I

2, S:$5 ENGR: Ad4

the new value. If the calculated setpoint is I greater than the existing value, the setpoint may remain at the lower value or increased to the new value.

Due to the fact that I-131 and I-133 comprise 98.5% of the total dose based on expected annual average releases (LGS FSAR Table 11.3-1) and I particulates contribute a minor fraction of the total dose, a particulate channel setpoint will not be calculated for purposes of the ODCM.

3. Containment Purge Isolation
a. Monitor alarm setpoints will be determined I for the North Stack Vent Wide Range Gas Monitor to initiate closure of the containment purge supply and exhaust lines in the event that high radioactivity I releases are detected. The setpoint will be determined to alarm and isolate containment at the minimum release rate from the North Vent which corresponds to a value less than I or equal to 2.1 uCi/cc. The total effluent high alarm setpoint for the Wide Range Gas Monitor will be calculated as follows:

l 1) Determine Si a l Si = Ci x F (min.) x 472 where:

Il l Si = containment purge isolation setpoint (uci/sec)

I Ci = a value < 2.1 uCi/cc determined by the plant staff.

I F(min) = minimum anticipated vent flow rate during purge 472 = units conversion factor to convert uCi/cc per CFM to uCi/sec l 4. Containment Purge During Routine Operations

a. Monitor alarm setpoints will be determined for the North Stack Vent Wide Range Gas Monitor to indicate to Control Room personnel that unanticipated high

'I l radioactivity releases are detected. The l setpoint will be determined to alarm in the l event that 10CFR20 dose rates at the site boundary are approached or exceeded. The l

f D A E: a/r//tf 28 HP: k ENGR: M

I total effluent alert alarm setpoint for the Wide Range Gas Monitor will be calculated as follows:

I l 1) Determine Sn l Sn = VFiQt l where:

l Sn = Containment purge 10CFR20 alert alarm limit (uCi/sec)

Ot = the total release rate of all noble gas radionuclides in the gaseous effluent (uCi/sec) based on the I lower of either the whole body exposure limit (500 mrem /yr) or the skin exposure limit (3000 mrem /yr)

VFi = fractional contribution to site boundary dose rate from the release point of interest 1.e. noble gas dose rate contribution from I the north vent divi,ded by total noble gas dose rate contribution from the North and South Vents Normally the VFi values will be determined on a monthly basis but may be performed more often in response to plant conditions, b) Prior to containment purge and venting, the monitor setpoint will be recalculated. The I calculations will bs based on the noble gases detected by isotopic analysis of the containment atmosphere. If the calculated I setpoint is less than the exisiting monitor setpoing, the setpoint will be reduced to the new value. If the calculated setpoint is greater than the existing value, the I setpoint may remain at the lower value or increased to the new value.

l S. Hot Maintenance Shop Setpoint Determination

a. The Hot Maintenance Shop Particulate and I Iodine setpoints are based on a worst case isotope assumption. Although the application of the worst case isotope results in a highly conservative setpoint, I

releases from the Hot Maintenance Shop are .

expected to be small by comparison. In addition, a sufficient margin of safety factor is built in to the calculation, to DATE: */ar/d 29 HP: M ENGR:M

I' preclude the application of a VFi for the reles.se point.

1. The iodine high alarm setpoint is set to alarm in the event that 10CFR20 dose I rates at the site boundary are approached or exceeded. The ittethodology is as follows:

C =

1500 mR/hr t (1.0E-05 sec/m3)(7000CFM)(472)(1.62E07 mrem /yr uCi/m3 I l where:

Ct = the concentration at the iodine monitor which I indicates that the 10CFR (.ose rate limit at the site boundary has been reached, 2.8E-06 uCi/cc 1500mR/yr= 10CFR20 dose rate limit for iodine, tritium and particulates with half lives greater than 8 days.

Il1.0E-05sec/m3=annualaveragedepletedChi/Q l 7000 CFM = maximum vent flow rate 472 =

conversion factor to convert uC1/sec per CFM to uCi/cc 1.62E07 mrem /yr = inhalation dose factor, I - 131 for uCi/m3 child, per Reg. Guide 1.109

2. Determine the Hot Maintenance Shop high alarm setpoint for iodine as follows:

l Si = 0.01(2.8E-06 uCi/cc)

Ilwhere:

Si = Hot Maintenance Shop Iodine high alarm setpoint; 2.8E-08 uCi/cc

.01 = Margin of Safety Factor to encompass I possible contribution from all other release points.

3. The particulate high alarm setpoint is set to alarm in the event that 10CFR20 l dose rates at the site boundary are approached or exceeded. The methodology is as follows:

1 DATE: J/ar//f 30 HP: &

ENGR : .$g/C.

I Sp = .01(2.8E-06 uCi/cc) l where:

I Sp = Hot Maintenance Shop Particulate high alarm setpoint; 4.59E-09 uCi/cc I .01 = Margin of Safety Factor to encompass possible contribution from all other release points .

2.8E-06 uCi/cc = The concentration at the iodine I monitor which indicates that 10CFR20 dose rate limits at the site boundary have been reached.

6.1 = ratio of the adult inhalation dose factor for Sr-90 x breathing rate for adult to the child I inhalation dose factor for I-131 x breathing rate for child. This ratio may be modified by plant personnel if the isotopes available for release are identified and a new ratio based on dose we.ighted averages is established I

I I ,

I I

I lI i

I

' DA E: F/3#/f/

31 HP: M4.--

ENGR: A&C I .

ATTACHMENT 1 -

Ot calculations

1. Ot(whole body = 500 (X/Q)v KiSi where:

Ot = the total release rate of all noble gas I radionuclides in the gase.oc effluent; uCi/sec.

(X/Q)v = 1.1E-05 sec/m3; the highest calculated

.I annual average relative concentration for an area at or beyond the site boundary for all vent releases (NE boundary).

Ki = whole body gamma dose factors due to noble gases listed on Table III.A.1 (from Reg.

Guide 1.109, Table B-1).

Si = the fraction of the total radioactivity in the I gaseous effluent comprised by noble gas radionuclide "i".

2. Q(t(skin))= 3000 (X70)vEi{Li +1.lMi)Si]

(X/Q)v = 1.lE-05 sec/m3; the highest calculate'd I annual average relative concentration for an area at or beyond the site boundary for all vent releases (NE boundary).

Li = beta skin dose factor due to noble gases, listed on Table III.A.1 (from Reg. Guide 1.109, l Table B-1).

l l Mi = air dose factor due to noble gases, l listed on Table III.A.1 (from Reg.

Guide 1.109, Table B-1).

Si = the fraction of the total radioactivity in the gaseous effluent comprised by noble gas l radionuclide "i".

! 3. Ot(thyroid)= 1500 l (X/Q)d {PiAi ,

lI lI 32 lI

4 I where:

Ot = the total release rate of radioiodines in the gaseous effluent; uCi/sec.

(X/Q)d = 1.0E-05 sec/m3; the highest calculated annual average depleted concentration for an area at or beyond the site boundary for all vent releases (NE boundary).

Pi = inhalation dose factor for child thyroid for I radiciodines mrem-m3/uCi-yr; 1.62E07 for I-131 and 3.85E06 for I-133 Ai = the fraction of the total radioactivity in the I gaseous effluent (iodine channel) comprised by radionuclide "i".

I I

I I

I I

I

I l

I .

I .

1 1

lI 33 8

O

  • I VII. BASES Site Specific Data Liquid dose factors, A , for section III.A were I

Note 1:

1 developed using the following site specific data.

The liquid pathways involved are drinking water I and fish. The maximum exposed individual is an adult.

A1( = (Uw/DW + UF x BFi) KO x DFi Uw = 730 liters per year; maximum adult usage of drinking water (Reg. Guide 1.109, Table 3-5).

Dw = 85; average annual dilution at Phoenixville Water Authority intake.

UF = 21 kg per year; maximum adult usage of fish (Reg.

Guide 1.109, Table E-5).

BFi = bioaccumulation factor for nuclide, i, in fresh-water fish. Reg. Guide 1.109, Table A-1, except P-32 which uses a value of 3.0E03 pCi/kg per pCi/ liter.

KO = 1.14E05 lE06pCi/uCi (lE03 ml/Kg)8760 hr/yr units conversion factor.

DFi = dose conversion factor for nuclide, i, for adults in total body or bone, as applicable. Reg. Guide I 1.109, Table E-ll, except P-32 bone which uses a value of 3.0E-05 mrem /pCi ingested.

I The data for D was taken from data published in Limerick Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3. All other data except I P-32 BF and DFi were u 3d as given in Reg. Guide 1.109, Revision 1, October 19~, . A P-32 BFi value was taken from Kahn, B. and K. S. Turgeon, "The Bioaccumulation Factor for Phosphorus-32 in Edible Fish Tissue", NUREG-CR-1336, March, I 1980. A P-32 DF value was taken from Limits for Intakes of Radionuclides by Workers, International Commission on Radiological Protection ICRP Publication 30, Supplement to Part 1, 1979.

Note 2: To develop constant P(I-131) for Section III.A, the following data were used:

P(I-131) = K' (BR) (DFA)

K' = 10E06 pCi/uCi; unit conversion factor I

I 34

I BR = 3700 m3/yr; child's inhalation rate.

I DFA I-131

= 4.39E-03 mrem /pCi; the thyroid inhalation dose factor for I-131 in the child.

I The pathway is the inhalation pathway for a child. All values are taken from Regulatory Guide 1.109, Revision 1, October 1977.

I Note 3: To develop constant R for section III.C, the following site specific data were used:

RGi (D/Q) = K'QF X (Ucp) (Fm)(r) (DFLi)a fp(1-fs g %tf A,1+b YP K' = 1E06pCi/uCi unit conversion factor QF = 6Kg/ day; goat's consumption rate Uap = 330 1/yr; yearly milk consumption by an infant M= (9.97E-07)/sec decay constant for I-131; 9.48E-06 for I-133.

lw = (5.73E-07)/sec decay constant for removal of activity in leaf and plant surfaces.

Fm = 6.0E-02 day / liter, the stable element transfer coefficient for I-131.

r = 1.0 fraction of deposited radiciodine retained in goat's feed grass.

I DFLi= 1.39E-02 mrem /pCi - the thyroid ingestion dose factor for I-131 in the infant; 3.31E-03 mrem /

pCi for I-133.

fp = 0.75; the fraction of the year the goat is on pasture (average of all farms).

fs = 0.0; the fraction of goat feed that is stored feed while the goat is on pasture (average of all farms).

Yp = 0.7 Kg/m2 - the agricultural productivity of pasture feed grass.

t = 2 days - the transport time from pasture to goat, f to milk, to receptor.

I l The pathway is the grass goat milk ingestion pathway.

These data were derived from data published in Limerick Generating Station Units 1 and 2 Environmental Report Operating Stage, Volume 3. All other data were used as lI g

I given in Reg. Guide 1.109, Revision 1, October 1977.

Similar data were used to develop the constant R for I-133.

Note 4: The methodology described herein will be implemented via computer codes. These codes have been verified as I documented in:

1. G.A. Technologies, RM-21A Computational Models, Document No. E-ll5-1241, June 1984.
2. G. A. Technologies, Meteorological Monitorino, I Display and Reporting System /RM-21A, Document No.

0375-9032, January, 1984.

Surveillance Requirement 4.11.1.2 Liquid Pathway Dose Calculations The equations for calculating the doses due to the actual release I rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I 10CFRPart 50, Appendix I", Revision 1, October 1977 and NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants". October 1978.

Surveillance Requirement 4.11.2.1.1 and 4.11.2.1.2 - Dose Noble Gases The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide I 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CPR Part 50, Appendix I", Revision 1, October 1977, I NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October 1978' and Regulatory Guide 1.111, " Methods for Estimating Atmospheric I Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977 with site specific dispersion curves and disperion methodology. The specified equations provide for determining the air doses in I areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric conditions.

I The dose due to noble gas release as calculated by the Gross Release Method is much more conservative than the dose calculated by the Isotopic Analysis Method. Assuming the release rates I given in Limerick Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3, the values calculated by the Gross Release Method for total body dose rate and skin dose rate are 4.8 times and 3.25 times, respectively, the values calculated by the Isotopic Analysis Method.

I z I

I For the Gross Release Method, Kr-87 and Kr-88 are used for the limiting skin and total body dose factors respectively, due to I half life considerations. Kr-89, the nuclide with the highest dose factors per Regulatory Guide 1.109 Table B-1 has a half-life of 3.2 minutes while the half-lives of Kr-87 and Kr-88 are 76 I minutes and 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> respectively. Therefore, by the time that gaseous effluents have been transported offsite, Kr-89 will have decayed enough so that Kr-87 and Kr-88 are effectively the most limiting nuclides.

The model Technical Specification LCO for all radionuclides and radioactive materials in particulate form and radionuclides other I than noble gases requires that the instantaneous dose rate be less than the equivalent of 1500 mrem per year. For the purpose of calculating this instantaneous dose rate, thyroid dose from I iodine-131 and iodine-133 through the inhalation pathway will be used. Since the expected annual releases presented in LGS FSAR Table 11.3-1 indicate that iodine-131 and iodine-133 releases have the major dose impact this approach is appropriate. The I value calculated is multipliad by 1.02 to account for the thyroid dose from all other nuclides. This allows for expedited analysis cnd calculation of compliance with the LCO.

Surveillance Requirement 4.11.2.2 and 4.11.2.3 - Dose Noble Gases I The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases I of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical I Specifications for Nuclear Power Plants", October 1978, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases l

I from Light-Water-Cooled Reactors", Revision 1. July 1977 with site specific dispersion curves and dispersion methodology. The specified equations provide for determining the air doses in l areas at and beyond the SITE BOUNDARY based upon the historical l

average atmospheric conditions.

l The dose due to noble gas releases as calculated by the Gross

!$ Release Method is much more conservative than the dose calculated

,5 by the Isotopic Analysis Method. Assuming the release rates given in Limerick Generating Station Units 2 and 3 Environmental

,E Report Operating License Stage, Volume 3, the values calculated

,g by the Gross Release Method for total body dose rate and skin dose rate are 4.8 times and 3.7 times, respectively, the values l calculated by the Isotopic Analysis Method.

Dose, Iodine '31, Tritium, and Radioactive Material in Particulate Form

'I l

I i 37 lI

I The equations for calculating the doses due to the actual release I rates of radiciodines, radioactive material in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days were developed using the methodology provided in I

Regulatory Guide 1.109, " Calculation of Annual Doses to Man from '

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977, NUREG-0133, " Preparation of Radiological I Effluent Technical Specifications for Nuclear Power Plants",

October 1978, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in g Routine Releases from Light-Water-Cooled Reactors", Revision 1, 3 July 1977 with site specific dispersion curves and dispersion methodology. These equations provide for determining the actual doses based upon the historical average atmospheric conditions.

Compliance with the 10 CFR 50 limits for radiciodines, radioactive materials in particulate form and radionuclides other I than noble gases with half lives greater than eight days is to be determined by calculating the thyroid dose from iodine-131 and iodine-133 releases. Since the iodine-131 and iodine-133 dose accounts for 99.97 percent of the total dose to the thyroid, the I value calculated is not increased.

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MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR ,

RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS FIGURE 'li."A5

z. 1 5 .

ATTACHMENT D3

'I I

Review and approval of Attachment D1 and ODCM, I Revision 1 by the Engineer-in-Charge, Nuclear and Environmental Section and PORC have provided the determination that these changes do not reduce the I accuracy and reliability of dose calculations or setpoints.

I I

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l I

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I ATTACHMENT D4 -

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E I

I Documentation of Engineer-in-Charge, Nuclear and Environmental Section and PORC Approval

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I I l'ECHANICAL ENGINEERING DIVISION N2 2301 Market Street l

FROM: R. A. Mulford FEB 2 21985 TO: G. M. Leitch

SUBJECT:

LGS PORC Approved Changes to the Offsite Dose Calculation Manual (ODCM)

REFERENCE:

Letter from G. M. Leitch to R. A. Mulford dated 1/28/85 FILE: GOW 1-1 (NRC)

As requested in the reference letter, we have reviewed the changes to ODCM. These changes are approved provided that the revisions noted on the attached pages are incorporated. These revisions have been agreed to in discussions with G. W. Murphy.

Please note that supporting information and docunentation should be supplied to the NRC with the ODCM changes in accordance with Tech Spec Section 6.14.2. Please address any questions to G. B. Rat >old, 841-6379.

D_-

. .. .ch ent s GBR/ nib /02208503 Copy to: A. R. Diederich/A. J. Marie w/o attach ents J. W. Ballantine/G. B. Rmt>old w/o attachnents R. W. Dublel/G. W. Murphy w/o attachments J. S. Wiley w/o attachments /

NES Chron w/o attactynents DAC 00-0 I

I

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