Text

Semiannual Effluent Releases Rept 26,Jul-Dec 1988. W/
	ML20235W060
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1988
From:	Burly W, Hunger G, Mcfadden J; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 8903100479
	Download: ML20235W060 (80)
	v • d • e

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t PEACH BOTTOM ATOMIC POWER STATION Unit Nos; 2 and 3 Docket Nos. 50-277 & 50-278 s

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SEMI-ANNUAL ~ EFFLUENT RELEASES REPORT.

n NO. 26 JULY 1, 1988 THROUGH DECEMBER 31, 1988

+

's Submitted to .

The United States Nuclear Regulatory Commission

.. Pursuant to-Facility Operating Licenses DPR-44 & DPR-56 -

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, PHILADELPHIA ELECTRIC COMPANY '

1 ;. PEACH BOTTOM ATOMIC POWER STATION Unit'Nos.-2 and 3.

Docket Nos.- 50-277-'&.50-278 c

SEMI-ANNUAL EFFLUENT. RELEASES REPORT NO.-26 JULY 1, 1988 THROUGH DECEMBER 31~, 1988.

Submitted to

' The United States. Nuclear Regulatory Commission Pursuant to

' Facility Operating Licenses DPR-44 & DPR-56 Preparation Directed By:

D. M. Smith, Vice President Peach Bottom Atomic Power Station

[, . TABLE OF' CONTENTS-Page No.

P

.I. Introduction' 11 II. . Tables IA. Gaseous Effluents - Summation 1 0 of All Releases e

i IB. Gaseous, Effluents For Release 2 Point - Main Stack IC. - Gaseous Effluents For-Release 4 Point - U/2 & U/3 Roof Vents 2A. Liquid Effluents - Summation of '6-AllLReleases 2B. Liquid Effluents. 7

3. Classes of Solid Radioactive 9" Waste-Shipments III~. Attachments A. Supplemental Information' 10 B. Calculations for Critical Organ Dose Using 12 Particulate With Half-Lives Greater than 8 Days C. Dose to the Teen Liver from Liquid Effluents Technical Concurrences: (for accuracy of information)

{ blS $~22N$

Date Dip'ctor-Radiation Protection Alas A A Al. i n4r

' Director-Radwaste' / / Date

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l SEMI-ANNUAL EFFLUENT RELEASE REPORT JULY 1, TO DECEMBER 31, 1988 I. INTRODUCTION In accordance with the Unique Reporting Requirements of Technical Specification 6.9.2.h(2) applicable during the reporting period, this report summarizes the Effluent Release Data for Peach Bottom Atomic Power Station Units 2 and 3 for the period July 1 through December 31, 1988. The notations E and E- are used to denote positive and negative exponents to the base 10.

The release of radioactive materials during the reporting period was within the Technical Specification limits. There were changes made to the Offsite Dose Calculation Manual (ODCM) during the reporting period.

A copy of the Offsite Calculation Manual is attached to this report.

There were no known unplanned releases of liquid radioactive material from the High Pressure Service Water (HPSW) system.

Iodine was not present from either the roof vents or main stack in section labeled Gaseous Effluents (Table 1A); therefore, the Critical Organ Dose for iodines in mrem was zero. In accordance with a revision of the Offsite Dose Calculation Manual (ODCM) the ,

Critical Organ Dose calculated using the particulate !

with half-lives greater than 8 days. These calculations are incorporated into the ATTACHMENT B section of the ;

report.

In section labeled Liquid Effluents (Table 2B) and section labeled Gaseous Effluents (Table 1B) the Stront ium 89 and Phosphorus 32 reported values are less than the lower limit of detection. Strontium 89 and Phosphorus 32's half-lives which are exceeded by a factor of eight half-lives; therefore, Strontium 89 and Phosphorus 32 are reported as zero for this reporting period. In general, any isotope with half-lives have been exceeded by a factor of eight half-lives shall be i reported as zero.

In order to satisfy an unresolved item from NRC Inspection Report 50-277/88 50-278/88-33, values for dose to the teen liver from liquid effluents were calculated for each semi-annual reporting period from !

1985 to date. The results, compiled in Attachment C demonstrate satisfactory compliance with Technical Specification 3.8.B.2 limits for the period reviewed.

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ya n TABLE'1A EFFLUENTANDWASTEDISPOSALSEMIANNUALREPORT;(19888) y a l y[ GASEOUS EFFLUENTS ~- SUMMATION OF ALL RELEASES Unit- Quarter -Quarter. Est. Total-2&3 3 4 Error. %

A. Fission & activation gases

1. Total release Ci 2.78 E2 3.21 E2 54.0 EO 2.. Average release rate for period uCi/sec 3.83 El 3.79 El

3. Gamma Air Dose Millirad 4.40 E-2 4.30 E-2 Percent of Tech. Spec. % 4.40 E-1 4.30 E-1

4. Beta Air Dose- Millirad 1.90 E-2 2.04 E-2 Percent of Tech. Spec. % 9.50 E-2 1.02 E .

8. Iodines

1. Total iodine-131 Ci 0 0 61.0 E0

2. Average release rate for period uCi/sec 0 0 3.tCritical Organ Dose Millirem 7.19 E-5 1.23 f-4 Percent of Tech. Spec. % 4.79 E-4 8.20 E-4 C. Particulate

1. Particulate with half-lives greater than.

8 days (includes Alpha and Strontium 89-90) C1 1.87 E-4 7.12 E,-4 61.0 E0 2.: Average release rate for period uCi/sec 2.58 E-5 8.41 E-5

3. Gross Alpha Radioactivity Ci 5.18 E-5 3.74 E-5 D. Tritium- '

1. Total release Ci 1.31 EO 4.54 E-1 94.0 EO

2. Average release rate for period uCi/sec 1.81 E-1 5.36 E-2 l

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TABLE 18 l

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (19888)

GASEOUS EFFLUEHTS FOR RELEASE POINT: MAIN STACK CONTINUOUS MODE BATCH MODE

~~

Nuclides Released Unit Quarter Quarter Quarter Quarter 2&3 3 4 3 4

~1. Fission gases Krypton-85M Ci 0 0 0 0 Krypton-87 C1 0 0 0 0 Krypton-88 Ci 0 0 0 0 Xenon-133 Ci 0 0 0 0 Xenon-135 Ci 0 0 0 0

' Xenon-135M Ci 0 0 0 0 Xenon-138 Ci 0 0 0 0 Unidentified Ci 2.07 E2 2.51 E2 0 0 Total for period Ci 2.07 E2 2.51 E2 0 0

2. Iodines Iodine-131 Ci 0 0 0 0 Iodine-133 Ci 0 0 0 0 Iodine-135 Ci 0 0 0 0 Total for period C1 0 0 0 0

3. Particulate ,

Strontium-89 Ci 0 0 0 0 Strontium-90 Ci 2.60 E-7 4.30 E-7 0 0 Strontium-91 Ci 0 o 0 0 Cesium-134 Ci 0 0 0 0 (2)

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TABLE"1B'(Continued).

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CONTINUOUS MODE BATCH MODE' .,

(',Nuclides Released Unit. Quarter Quarter- Quarter.

~

Quarter

,.t 2 &~3 3-. 4 3- ~4 i ~

Cesium-137 Ci 0 -- .4.12 E-6. 0- 0

Cesium-138 Ci .0- 0 0: 0" Barium-139-- Ci 0 .0 >0 0

-Barium-140 Ci 0 0 0 0 Lanthanum-140- Ci 0 0 0 0.

Cobalt-57. Ci 0 0 0 -0 Cobalt-58 Ci- 0 0.. 0 0-Cobalt-60 Ci 0 -5.78 E-6. O. O n Zinc-65 ,

Ci 0 0 0 0

? . Yttrium-91M Ci 0, _0 0 0 lodine-133- Ci 0 0 .- 0 0

-Copper-64 Ci 0. 0 0- 0 Rubidium-88 Ci 0 0 0 0 Manganese Ci 0 0 .0 0

. Strontium-92. Ci 0: 0 0 0 -..

Totals C1 2.60 E-7 1.03 E-5 0 0 (3)

i; n ,, TABLE ~ Ice l; EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (19888) 1

' GASEOUS EFFLUENTS-FOR RELEASE' POINT: U/2 & U/3 Roof Vents j CONTINU0US MODE BATCH MODE.

Nuclides Released- Unit Quarter Quarter -Quarter Quarter 2&3 3- 4 3 -4

1. Fission' gases Krypton-85M Ci 0 0- 0- 0 Krypton-87 Ci 0 0 0- 0-Krypton-88 C1 0 0 0 0 Xenon-133 Ci 0 0 0 .0

, : Xenon-135 Ci 0 0 0 0

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Xenon-135M Ci 0 0 0 0 Xenon-138 .Ci 0 'O O O '

Xenon-133M Ci 0- 0 0 0

. Unidentified Ci 7.05 El 7.02 El 0 0- ,

--Total for period Ci 7.05 El 7.02 El 0 0

2. Iodines Iodine-131 Ci 0 0 0 0 lodine-133 Ci 0- 0 0 0 Iodine-135- Ci 0 0 0 0 Total for period Ci 0 0 0 0

-3. Particulate Strontium-89 Ci 0 0- 0 0 Strontium-90 C1 1.11 E-5 1.00 E-5 0 0 Strontium-91 C1 0 0 0 0 Cesium-134 Ci 0 0 0 0 Cesium-137 Ci 4.86 E-5 1.07 E-4 0 0 (4)

TABLE 1C (Continued)

CONTINUOUS MODE BATCH MODE

'Nuclides Releasea Unit Quarter Quarter Quarter Quarter 2&3 3 4 3 4 Cesium-138 Ci 0 0 0 0 8arium-139. Ci 0 0 .O ~0 8arium-140 Ci 0 0 0 0 Lanthanum-140 Ci 0 0 0 0 Cobalt-57 Ci 5.18 E-5 0 0 0 Cobalt-58 Ci 0 0 0 0 Cobalt-60 Ci 0 2.72 E-4 0 0 Zinc-65 Ci 0 2.51 E-4 0 0 Yttrium-91M Ci 0 0 0 0 lodine-133 Ci 0 0 0 0 Copper-64 Ci 0 0 0 0

. Rubidium-88 Ci 0 0 0 0 Manganese-54 Ci 0 0 0 0 Strontium-92 Ci 0 0 0 0 Totals C1 1.11 E-4 6.40 E-4 0 0 (5)

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BLE 2A1 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (19888)

LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES i

Unit Quarter Quarter Est. Total 2&3 3 4 Error, %

A. Fission & activation gases

1. Total release (not including Ci 2.20 E-2 7.87 E-3 32.0 E0 tritium, gases, alpha)

2. Average diluted concentration uCi/ml 3.26 E-9 7.42 E-10 during period

3. Whole Body Dose Millirem 1.70 E-1 1.92 E-2 Percent of Technical Specification % 5.67 EO 6.40 E-1

4. Teen Liver Oose Millirem 2.45 E-1 3.53 E-2 Percent of Technical Specification % 2.45 E0 3.53 E-1 B. Tritium

1. Total release C1 1.51 E0 1.35 E0 39.0 E0

2. Average diluted concentration during period uCi/ml 2.24 E-7 1.27 E-7 C.-Dissolved and entrained gases

1. Total release Ci 0 0 42.0 E0

2. Average diluted concentration during period uCi/ml 0 0 D. Gross alpha radioactivity

1. Total release Ci 3.54 E-5 8.08 E-5 39.0 E0

2. Average diluted concentration during period uCi/ml 5.25 E-12 7.59 I 12 E. Volume of waste released (prior to dilution) liters 2.10 E6 3.47 E6 32.0 E0 F. Volume of dilution water used during period liters 6.74 E9 1.06 E10 30.0 E0 (6)

7 LTABLET2B la .

t-. : c EFFLUENT AND WASTE DISPOSAL SEMIANNUAL-REPORT (19888)'

. LIQUID' EFFLUENTS CONTINUOUS MODE- BATCH MODE

> - Nuclides. Released Unit Quarter-- Quarter- Quarter Quarter 2&3 3 .

4 3.. 4 r

p , Strontium-89 Ci 0 0 0 0 1

Strontium-90. Ci -0 0 2.76 E-5. 2.39 E-5

! Alpha Ci 0 0- -3.54 E-5 .8.0E M

.y-l Tritium Ci 0- 0 1.51 E0 1.35 E0 Phosphorus-32 Ci 0 0 0 0

, Iron-55 Ci- 0 0 3.40 E-4 1.14 E Xenon-131M Ci 0 0 0 0 Xenon-133 - Ci 0 0 0 0 --

tXenon-133M Ci- 0 0 0 0 4 Xenon-135 Ci 0 0 0 0 LXenon-138 Ci 0 0 0 0

-iKrypton-85M Ci 0 0 0 0

~ Krypton-87 Ci 0 0 0- 0 Krypton-88 Ci 0 0 0 0

Xenon-135M Ci 0 0 0 0

Manganese-54 Ci 0 0 0 0 Celsium-134 Ci 0 0 5.80 E-3 4.29 E-4 Cesium-137 Ci 0 0 1.00 E-2 1.44 E-3

. Cesium-138 Ci 0 0 0 0 Zinc-65 Ci 0 0 2.68 E-3 1.96 E-3

--Sodium-24 Ci 0 0 0 0 Cobalt-58 Ci 0 0 0 0 Cobalt-60 Ci 0 0 3.24 E-3 3.23 E-3 lodine-131 Ci 0 0 0 0 lodine-133 Ci 0 0 0 0

. Molybdenum-99 Ci 0 0 0 0

-(7) - - . - -- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 TABLE 28 (Continued) l CONTINUOUS MODE BATCH MODE Nuclides Released Unit Quarter Quarter Quarter Quarter 2&3 3 4 3 4 lodine-135 Ci 0 0 0 0

)

Barium-140 Ci 0 0 0 0 Neptunium-239 Ci 0 0 0 0

.fbromium-51 Ci 0 0 0 0 Yttrium-91M Ci 0 0 0 0 Strontium-91 Ci 0 0 0 0 Antimony-122 Ci 0 0 0 0 Tellurium-132 Ci 0 0 0 0 NTobium-95 Ci 0 0 0 0 Lanthanum-140 Ci 0 0 0 0

-Cadmium-109 Ci 0 0 0 0 Cesium-136 Ci 0 0 0 0 Silver-110M Ci 0 0 0 0 Cesium-144 Ci 0 0 0 0 Antimony-124 Ci 0 0 0 0 Iron-59 Ci 0 0 0 Tellurium-129M Ci 0 0 0 0 Tellurium-131M Ci 0 0 0 0 Zirconium-95 Ci 0 0 0 0 Cerium-141 Ci 0 0 0 0 Total for Period (above) Ci 0 0 1.532 E0 1.358 E0 (8) l

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, . ATTACHMENT.A'

, SUPPLEMENTAL.INFORMATION -

H

Facility: ~ Peach Bottom Units 2 to 3 Licenses: DPR L.- :DPR-56

-If . Regulatory L'imits (Technical' Specification Limits)-

A. Noble' Gases:

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l.- 1 500 mrem /Yr -' total body " instantaneous" limits per 5-3000 mrem /Yr - skin Tech. Spec. 3.8'.C.l.a-

2. .5.10 mrad. .

gamma air quarterly air dose limits-.per 5 20 mrad.- beta air Tech. Spec.13.8.C.2.a

3. 1-20 mrad - gamma air yearly air d.ose' limits per:

5 40; mrad.- beta air Tech.. Spec. 3.8.C.2.b B. Iodines, Tritium,' Particulate with Half Life <> 8 days:

l ~. i 1500. mrem /Yr.- any organ - " instantaneous"ilimits;per-(inhalati,on' path) . Tech. Spec.-3.8-C.1.br 2.- jl5 mrem'-'any organ

- quarterly dose limits per Tech. Spec. 3.8.C.3.a' 3.. i su mrem - any' organ - yearly dose limits per

. Tech. Spec. 3.8.C.3.b C. (Liquid" Effluents:

'1. Concentration <'10 CPR 20, " instantaneous"'limitsLper Appendix B,'TaEle II, Col. 2 Tech. Spec. 3.8.B.1'

2. . 3 3.0 mrem - total body quarterly dose limits per 5.10 mrem - any organ Tech. Spec. 3.8.B.2.a

3. $ 6.0 mrem - total body - yearly. dose limits per i_20 mrem - any. organ- Tech. Spec. 3.8.B.2.b'-

2. Maximum ~ Permissible Concentrations

MPCs1are not used to calculate permissible release rates and

' concentrations for gaseous releases.

The MPCs specified in 10 CFR 20, Appendix B, Table II, Column 2, for' identified n'uclides are used to calculate permissible release

' rates and concentrations for liquid releases per Peach Bottom Technical Specification 3.8.B.l.

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. ATTACHMENT'A'(Continued).

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'n 3.Kl Average ~ Energy

,g LNot applicable.

"4 . : . Measurements and Approximations of' Total Radioactivity A. Fission and Activation Gases-The method =used is-the Nuclear Data.6600/6700 Counting System

- Gas Marinelli B.- - Iodine The method used in'the Nuclear Data 6600/6700 Counting System

-. Charcoal. Cartridge C. Particulate:

The method used.is the Nuclear Data 6600/6700 Counting System-

- Air Particulate Sample, 47 mm filter D.' Liquid Effluents:

n The method used~is the Nuclear-Data - 6600/6700 Counting System l and the Radwaste Liquid Discharge Pre-Release Method with a

? liter bottle.

5. . Batch Releases oh. Liquid Q3 Q4

of Batch Releases: 31- 48 Total Time for batch releases, minutes 7918 11965 Maximum time period for a batch release, minutes 340 340 Average time period for batch release, minutes 255 249 Minimum time period'for a batch release, minutes 20 21 )

Dilution flow (Liters) 6.74 E9 1.12 E10 B Gaseous: N/A

6. Abnormal Releases

>A. Liquid: See Attachment B B. Gaseous: None J

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't ATTACHMENT-B ,

N' l

l Calculations'For Critical-Organ Dose Using Particulate With- j Half-Lives Greater Than 8 Days '

The'following doses were calculated using particulate identified during the third and. fourth. quarter reporting period of 1988; PARTICULATE CRITICAL ORGAN DOSE, LIVER, .(MREM)

Quarter 3 Quarter 4 I

.Sr-89 8.78 E-8 9.68 E-8 Sr-90 3.67 E-6 3.33 E-6.

i .Cs-137 3.09 E-5 6.90 E-5 Co-57 3.72 E-5 0.00 E-0 Co-60 0.00E-0 2.53E-7 Zn-65 0.00E 5.05E-5 Total 7.19 E-5 1.23.E-4 :

The percentage of Technical Specification for the Critical Organ (liver) for the third quarter is 7.19E-5 MREM, fourth quarter is 1.23E-4 MREM.

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.3 ATTACHMENT C

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Dosa to-the'Tean-Livar From Liquid = Effluents.

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' Percent T/S:3.8.B.2 Dose .(mrem)' limit (1) 1st. Half'1985 1.519'

-~2nd Half 1985 1.624

- Total Year 1985 3.143 15.72%

lst Half 1986 1.183 2nd Half'1986- 0.044

Total Year 1986 1.227 6.14%

lst Half 1987- 0.824 2nd Half 1987 .0.609 Total Year 1987 1.433 7.17%

lst Half 1988 '1.742 2nd Half 1988 0.280

. Total Year 1988 2.022 10.11%

(1)'T/S 3.8.B.2b limit of 20 mrem / year to any organ:

(most restrictive organ limit)

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i Peach Bottom Atomic Power Station Units 2 & 3 Offsite Dose Calculation Manual Revision 3

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4 Offsite Dose Calculation Manual;

' Revision.3

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Peach Bottom Atomic Power Station Units 2 and 3 Phila' delphia Electric Company Docket Nos.-' 50-277 & 50-278 PORC Approval

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97 PORC Chairman PORC Meeting i :~ ff- W Od /;2/F//F Date t

Table of Contents I. Purpose II. Instrument Setpoints III. Liquid Pathway Dose Calculations A. Liquid Radwaste Release Flow Rate Determination B. Surveillance Requirement 4.8.B.2 C. Surveillance Requirement 4.8.B.4a IV. Gaseous Pathway Dose Calculations A. Surveillance Requirement 4.8.C.1 B. Surveillance Requirement 4.8.C.2 C. Surveillance Requirement 4.8.C.3 D. Surveillance Requirement 4.8.C.5a E. Surveillance Requirement 4.8.C.6b V. Nuclear Fuel Cycle Dose Assessment - 40 CFR 190 A. Surveillance Requirement 4.8.D VI. Calendar Year Dose Calculations A. Unique Reporting Requirement 6.9.2.h VII. Radiological Environmental Monitoring Program A. Surveillance Requirement 4.8.E VIII. Bases

L Rev. ,

I.. Purpose LThe purposeLof-the.OffsiteLDose Calculation Manual'is'to

-establish methodologies and procedures for' calculating doses ~to' individuals in areas at and beyond the> SITE' BOUNDARY due-to radioactive. effluents from Peach Bottom t- Atomic; Power Station. The results ofethese calculations

'are required _to determine-compliance.with Appendix A to- A LOperating-Licenses-DPR-44 and DPR-56, " Technical.

' Specification.and. Bases:for Peach Bottom Atomic. Power- l Station Units No. 2 and 3".

-II.. Setpoint Determination for Liquid.& Gaseous Monitors'c

h. Liquid.Radwaste Activity Monitor Setpoint Each tank.of-radioactive waste'is-sampled prior to release. A small; liquid volume of this' sample is- '

analyzed for gross gamma (wellLcount) activity. This analysis is performed in a:NaI well counter. This well counter has a counting efficiency similar to the-liquid radwaste discharge; gross activity monitor. The wellicounter and liquid radwaste discharge gross activity monitor are calibrated against'the same liquid radioactivity source in the geometry'to be used by each detector. An efficiency is determined for,

each radwaste tank to be released. Exceeding the expected response would indicate that an. incorrect-sample had been obtained-for that release and the release ~is automatically stopped.

S.P. = (Net CPM /ml(well) X Eff W/RW) + Background CPS S.P. = Liquid Radwaste gross activity. monitor setpoint in' CPS Net CPM /ml(well) = gross gamma activity for the radwaste sample tank determined by the well counter..

Eff W/RW = conversion factor between well counter and liquid radwaste gross activity monitor (CPS (R/W monitor) - CPM /ml(well)).

Background CPS = Background reading of the liquid radwaste gross activity monitor (CPS).

4

Rev.. ; 3 -

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[ , IThe alarm-and trip' pot setpointsifor--the liquid P4 radwaste-activity monitor are determined from'a-

~ calibration--curve'for the alarm potTand trip pot. .The-alarm pot. setting includes:a factor;of'l'.25 to allowL 4 :for analysis error,. pot setting error, instrument E

-error =and calibration error. .The' trip 7 pot setting

'includesia: factor.of 1.'35 to allow.for' analysis' error, pot setting error, instrument L error and calibration-error. The ; flow rate detensination' includes : a margin of assurance which< includes L consideration: of these errors such'that the' instantaneous release limit ofE10 E CFR 20 is not exceeded.

B. Liquid RadwasteLRelease Flowrate Setpoint Determination.

The. trip pot setpoint for the: liquid radwaste release flowrate is determined'by multiplying the liquid radwaste flowrate. determined above'by 1.2 and using

'this value on.the appropriate calibration curve for

'the' discharge < flow meter:to be used. The Peach Bottom radwaste, system has twoLflow monitors-(high. flow (5 to 300 gpm) and low flow.(0.8 to 15-gpm)). 'The1 factor of ~

1.2' allows foripot setting error and. instrument error.

Tne flow rate determination includes a'marginfof assdrance:whichcincludes consideration of this error

'such that,the instantaneous release limit1 of.10 CFR'20' is not exceeded.

C. Setpoint Determination for Gaseous Radwaste The high and high-high alarm setpoints.for the' main stack; radiation monitor, Unit:2 roof vent radiation monitor and Unit 3 roof vent radiation monitor are determined as follows:

High' Alarm - the high alarm setpoint is. set at.

approximately 3 x the normal monitor reading.

High-High Alar _m - the high-high alarm setpoint is ,

set at a release rate from this '

vent of approximately 30% of the instantaneous release' limit of 10 CFR 20 as specified in o

Technical Specification 3.8.C.1.a for the most restrictive case (skin or total l body) on an unidentified basis. l To determine these setpoints )

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m -- , - - - .-- _

Rsv. 3 ')

~

solve the gaseous;offlusntsdose?

p rate ' equations in- section IV. A -

of the ODCM.to determine what main' stack release-rate and-

. roof vont release rate will produce a dose rate.of 150 mrem /yr to the' total body and'a fi dose rate of 900 mrem /yr to i the. skin (30%'of the limit'of l

'3000 mrem /yr) from each release ;

point. Using the smallest (most restrictive) release rate for' each release point determine q monitor response. required to produce this release rate assuming a normal vent flow

~

rate and pressure-correction.

factor. Set the high-high alarm for.approximately this monitor.~ response.

D. Setpoint Determination for Gaseous Radwaste Flow Monitors The' alarm'setpoints'for the main stack flow monitor is as follows:

Low Flow' Alarm - 10,000'CFM. - This setting insures that the main stack minimum-dilution.

flow as specified in' Technical Specification-3.8.C.4.a is maintained.

The alarm setpoints for the roof vent flow' monitors are as follows:

5 Low Flow Alarm - 1.S'x 10 cfm 5

High Flow Alarm - 5.4 x 10 cfm III. Liquid Pathway Dose Calculations

'A. Liquid Radwaste Release Flow Rate Determination Peach Bottom Atomic Power Station Units 2 and 3 have one common discharge point for liquid releases. The following calculation assures that the radwaste release limits are met.

-Rev. 3 Th2 flow rate ofiliquid radwaste rolesend fron the site to areas at and-beyond the SITE BOUNDARY shall be such that.the concentration-of radioactive-material after dilution shall be limited to the concentration specified in 10 CPR 20.106(a) for. radionuclides other than. noble gases and 2E-4 uCi/ml total activity concentration for all noble gases as specified in Technical Specification 3.8.B.l. Each tank of radioactive waste is sampled prior-to release.and is quantitatively analyzed'for identifiable gamma emitters as specified in Table 4.8.1 of-the Technical Specifications.. 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 MPCi i

uCi/ml i = the activity of_each identified gamma emitter in uCi/ml MPCi = The MPC specified in 10 CFR'20, Appendix B, Table II, Column 2 for radionuclides

-4 other-than noble gases or 2 X 10 uCi/ml for noble gases.

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

5 Release Rate (gpm) = A X l2.0 X 10 B X Dilution Factor A = The number of circulating water pumps running which will provide dilution 5

2.0 X 10 = the flow rate in gpm for each circulating water pump running B = margin of assurance which includes consideration of the maximum error in the activity setpoint, the maximum error in the flow setpoint, and possible loss of 5 out of the 6 possible circulating water pumps during a release. The value used for B is 10.0 .

l

. Rmv. 3

'B. Survaillanca Requirement 4.8.B.2 Dose' contributions from liquid effluents released'to.

areasLat and beyond'the SITE BOUNDARY shall:be calculated:using the equation below. This dose

~

calculation uses those' appropriate radionuclides-listedLin Table'III.A.l. These radionuclides account g forivirtually 100 percent ofithe total body dose and.

organ dose from liquid effluents..

The dose for each. age. group and each organ:should be calculated to determine the: maximum total ~ body. dose

.and organ' dose for'each quarter'and the year, as appropriate. Cumulative dose files for quarterly and yearly doses should be maintained separately and the

, maximum total body.and organ dose reported in each case.

L ____- - - _ - - _ - - _ .

D.

.R v. 3 h l l m h -D = A I [ 8t C F il 1-i i 1 = l- 1 q

where: 1 D ' = the cumulative' dose commitment to the total ,

body or any organ, , from liquid effluents for the total time period m~ , in mrem di t -

1}=1 _

1 l 2h t .

= the length of the lth time period over which l 1 C and F are averaged for the liquid release, il 1 in hours.

C = The average concentration of' radionuclides, i, in-11 undiluted liquid effluent during time.periode66 t-from any liquid release, (determined by the effluent sampling analysis' program, Technical-Specification Table 4.8.1), in uCi/ml.

A 1F = the site related ingestion dose commitment i factor. to the total body or organ,1F , for each radionuclides listed in Table III.A.1, in mrem-al per br-uci. See Site Specific Data.**

F = the near field average dilution factor for 1 C during any~ liquid effluent release. Defined il as the ratio of the maximum undiluted-liquid waste flow during release to the average flow from the discharge structure to Conowingo Pond.

III.C Surveillance Requirement 4.8.B.4a Projected dose contributions from liquid effluents shall be calculated using the methodology described in section III.B.

- See Note 1 in Bases ;

tD, O, D.,

- y ~ m u m m u w v w m t~y wn tw yC O O O O O O O O O O -O O C .O O O

+ +w + W+ +.U + + +W + W+ +W + W+ + + + + + + +

W 4eJ W -W W W W W W W W W W O O ID U ID N T U C M - M @ @ N O O O @ O M J @ = * * * * @ %* *

N. . v. M. . m. C. N. e C. . . . M. N. T. U. . G. . m

% N

O M - N @ *

N @ N N - M M M M M @ A V 3 O N @ T N M M M M M T T T T - - T U M N V O C D C O O O O O O O O O O O O O O O O =

= + += + + + + +W + W+ +W + W+ + + + + + + + + >

JZ W W W W W W W W W w W W W W W -

JW M L

T O O @ M 2 M N m M m N M T @ O @

W m.

N *

m. @ @

m.

.V

.I..r T. C. C.

N

. O. M. . N. . . O. O. O. O. O. C. m .c O e * *

N

T M M

O T T @ m a e @ W W WV O N m T N M M M W M T T T T -

T T T M CW r.

O O O O O O O O O OO O OO O O O O O O em

+ + + + + + + + + + + + + + + + + + + + 30 I O W W W W W W W W W W W W W W W W W W W W Uh '

O M m M m N O m O m W N O T N m N T W W m Wu 4 @ @ @ W *

mW

. . N. O. N. . M. M. M. . N. T. T. O. m. C. T. . . . C N * *

M @ N @ m W N @ m N mm a * * = OE O s M T - M T ~ ~ m T e *

"E W_

O O a O e m 5 m O 2 mO OO O O O O O O mu

+ + w +

e*e * * + * * + + + + + + + + t Om O W A 4 W e E W W 5 W W W W W W W W W VE J O C V M D D D D N D D @ N N N * @ O N N I

N

@- 0 0 0 0 0 N. N. N.

N

m. T O. E.

@ *W 0 C . .

C4 U N -

C - C C C C T C C m m a @ - N @

M se u

o N M T T M T N N m T W

s- h

.e. .e. o C C C O E 9 m E O OO O O O O O O WD

>Z + + + + * * *. * + + + + + + + + + 1 -

WW W w m W W mm 2 W W m W W W W W W W W W C4 ZW M O V e V V V V = V V T

N M @ @ N O T @W ON O = 4 v * @ --

p e a O O O O O T. O O N. O. N. . . T. . M. Mw M =

C - C C C C 4 C C @ @

m

N N N N 96 N O N M T T M T N N 2 9 m - > O J O O 2 O m a e 5 O e S O OO O O O O O O e O + + w + * * * * + * * + + + + + + + + t rc O W W e W m e m a W mm W W W W W W W W W O C M m V M V V V V N V V e m M N N N N @ m V*

C @ N (9 W e @ O C *.

. O e O O O O e O O m. . m. . . m. s M. S2 N - C

C C C C @ C C m m - 4

N N N N e a@

N

C m M M T T m T M M N

m T m M C-w m O O 2 O O mm O O O O O O O C O O O O W O + + w + + * * + + + + + + + + + + + + DV m a J m

W e W W m a W W W W W W W W W W W W - C E = = V Z W V @ C V V * @ m O @ N M T C O

N b e O U I 7 = @ * > * * *

p. O p C M. O . . O O m. . . m. O. O. . m. M. . O. N. WW U i C - M C *

C C N m @

e M T @ T T @ = m 4 2 VW l 6 I m =

N I. N M T T m W M M N - m W W N - W O e O O m a e O O O O O O O O O O O s- J W

m 2

WW 2 *m + 6 + *4 + + *

W* + + + + + + + + + + + UW W m W e e W W W W W W W W W W W W e 6 O G ZW V O C V N O V V M O = N - N C N O @ m M =

Om Da % W M @ @ w O J g c . m. O . M. O O V. M. M. N. T. M. . m. T. N. . CC ZW X C

N C m e C C N M @ w

N N M M M T m e s

.1 0N

3 mW X N m N M T T m T M M N - m T m N t-al NW w r e O O m O O e e O C O O O O O O O O O O Ww ma E J * = + * + + * * + + + + + + + + + + + + ww !

W2 3 O C e e m w W W m W W W W W W W W W W W W UW D

= UO w O V m W V N @ V V @ m N N N @ m N N O T N W l

ZU 4 J N N

m. *C

= gr O

M. O =.

O O . O. a

N. N. N. T. T. T. WW a > 0 C - N C m a C C N M N - - N N M M M T m VW J WG N B D ZU U C to 4 WW d O N O M N M N N T M N M N

O O @ O CW

= OO W O +

C, O D D C O D 0 mm O O O OO O O O O a Jw s W + + + + + + + * * + + + + + + + + + x 6 W

O

W w W W W W W e a W W W W W W W W W g@

W m 3 0 %

M M M T N @ V V N O @ @ m @ M m a sE W O V N @ O @ - - -

O e O. M. N. O. O. .' O c O. T. M. . T. C. . N. . e O N *

T @ N - M @ C C m O a T E N

m a e

CW D O .s O M N M N N T M N M N - m m m * *4 0 - 0 0 C C D C O O O e e Q OO O O O O O O *

= 4 22 wa

+ +* + + + w+ +w + W+ * * + + + + + + + + 4 m a W

W W W e e W W W W W W W W W WL

>W M C T @

N T N @ V V @ N M N @ @ @ N O VC

W

O * @ @ *w a m. O. O. O. . M. T. O 0 C. N. m. M. M. M. M. * ~

" * ~ O @ M

M m C C @ m - M @. m e @ @ uw DC O fe T M N M N N W M N M N

O O m e CD N O O C

+ + * + + + + +

D D C C C C e e o OO O O O O O O 00 J g * * * + + + + + + + t *U C W W W W W W W W W E e W W W W a W d W W VU O M M m m* N @ @ m W V V T m m m m m @ N O WS 4 @ @ ~ T 6

. m. . C. . . M. m. O 0 N. T. N. T. N. M. O. O. O N ~ @ mW N

M E C C T m

M W W * @ b Le O

& h O N W M N M N N T M m M N M N

m T T

2 O O O O O O O O O O O OO O O O O O O WV

* + + + + + + + + O, + + + + + + + + + L O W C

h m

W M

W W W W W W W W W W W W W W W d W SW J m @ @ N N @ m N @ T m @ O @ O @ N C m 5 3 O @ * - @ T MC I . M. . M. . C. . T. N. N. N. . M. N. T. M. T. a 6 U N e m * *

M @ T *

N @

N M - 5 m M Ob i

> *C '

O O N T M N M N N 9 N 2 M N M N = $ T m e Uw O O C C C O O O O O O O O OO O O O O O O e CZ + + W+ + + + + + + + + + + + + + + + + + wV W W W w w W W W W W W W w W W W W W W W W JW M O @ O O N N m m m @ N

T @ M m m* O N Wh Cr @ N

v. T * @ @ W

T W @ MN

O. . . T. . . m. . m. N. O. . . N. . OV O * * @ * *

N N M @

N T -

N M @ N M Vb N O N 9 N N M N N T N m M N M N

O T 2 - V J O O O O O O O OO O O O O O O O O O O O W W D + + + + + + + + + + + + t + + + + + + + *L O w 6 W W W 6 W W W W W W W W 6 w W W W W MW C M m M N - T @ C N M m - N O @ N T @ m C e j e.

(D.

17 *

m. T. e. (D. ED. D. O.

T ED @ N N

@ *W

. . . . . L l N = m @ ~

N N M e

N T e * * @ @ M M 90 '

i a E E E "U p-

.O @

N v @ N O e Cm 9 9 m @ m O O @ O N M M e M M M M T .w

J N r. m m m o @ @ m @ - e

M M - * -

W CJ M -

M e i e i e e e i 1 . -

I e i e +W

( O i 4 *r 2 W W D O 2 C a w w w I 1 M m m C 04 c2 1 2 k 6 6 6 W N m m > r = = = V U V O ZN l

Rev. 3

-IV. Gaseous Pathway Dose Calculations A. Surveillance Requirement 4.8.C.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 Gross Release Method, assumes that all noble gases; released are the most limiting'nuclide - Kr-88 for total body dose and q Kr-87 for skin' dose.- Method (b); the Isotopic 4 Analysis Method, utilizes the results of noble

. gas analyses required by specification 4.8.C.la.

.For normal operations, it is expected that method (a) will-be used. However, if noble gas releases-are close to the limits as calculated by method' (a), method (b) can be used to allow more operating flexibility by using data.that more accurately reflect actual releases.

a. Gross Release Method D =VQ +K (X/Q) Q TB NS V NV D = (L (X/Q) + 1.lB) h + (L + 1.1M) (X/Q) h a s NS V NV where:

The location is the site boundary, 1097m SSE from the vents. This location results in the highest calculated dose to an individual from noble gas releases.

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

TB

{

D = skin dose rate , in mrem /yr.

s

]

,W" .

Rav; ' 3

-4 V- =,4.72 X 10 mrem /yr per1uci/sec;'the1 constant

.for-Kr-88 accounting forLthe gamma radiation

, from.the elevated finite plume. This' constant was de'iloped using MARE program.with-plant specific inputs for-PBAPS.

b =,the gross release. rate of' noble gases from the.

NS stack determined by gross activity stack .

monitors averaged over one hour, in uCi/sec..

4 3-K. = 1.47 X 10 mrem /yr per.uCi/m ;.the total bddy dose factor due to gamma emissions for Kr-88 (Reg. Guide 1.109, Table B-1).

-7 3 (X/Q) .=15.33 x.10 sec/m ; the highest calculLted v annual' average relative concentrationifor any.

area'at or beyond the SITE BOUNDARY for all vent releases, b = the~ gross release rate of' noble gases in gaseous NV - effluents from vent releases determined'by gross activity vent monitors averaged over one hour, in uCi/sec.

3 3 L = 9.73 x 10 mrem /yr per uCi/m ; the skin dose factor due to beta emissions for Kr-87..(Reg.-

Guide 1.109, Table B-1).

-8 3 L (X/Q): = 9.97 x 10 sec/m'; the highest calculated s' annual-average relative concentration from the stack releases for any area at or beyond the SITE BOUNDARY.

-4 B = 1.74 x 10 mrad /yr per uCi/sec; the constant for Kr-87 accounting for the gamma radiation from

.the elevated finite plume. This constant was developed using MARE program with plant specific inputs for PBAPS.

3 3 M = 6.17 x 10 mrad /yr per uCi/m ; the air dose factor due to gamma emissions for Kr-87.

(Reg Guide 1.109, Table B-1).

/

. 9 E _ _ _ . _ _ . . _ _ _ _ _ _ _ . _ - _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ .

c /

I'

' Rav. ' 3

b. ' Isotopic Analysis' Method D =),(V' h +K (X/Q). h)

TB i i is i. v iv

=[ (L-(X/Q) + 1.lB-) h

~

D +'(L +-1.lM ) (X/Q)

(h )

s 1; i s i ' is i i V- .iv .]

where:

TThe location is the site: boundary, 1097m SSE from thel vents. This location.results in the highest calculated dose to an: individual from. noble gas

. releases.

D = total body' dose rate,-in. mrem /yr.

TB D ~= skin dose, in mrem /yr.

S V =-the constant'for each identified noble gas.

i radionuclides for the gamma radiation from the elevated finite plume. The. constants were developed using the MARE program with plant specific inputs for PBAPS. Values are. .

listed on Table IV.A.1, in mrem /yr per uC1/sec.

h = the release rate of noble gas radionuclides, is i, in gaseous effluents from.the stack determined by isotopic analysis averaged.

over one hour, in uCi/sec.

K = the. total body dose factor due to gamma i emissions for each identified noble gas radionuclides. Values are listed on Table IV.A.1, 3

in. mrem /yr per uCi/m .

-7 3 (X/Q) = 5.33 x 10 sec/m.; the highest calculated v annual average relative concentration for any area at'or beyond the SITE BOUNDARY for all vent releases.

l

y ,- - - - -

Rsv. 3 h = the release rate of noble gas' radionuclides, iv i, in' gaseous effluents from all vent

-releases determined by isotopic analysis averaged over one hour,-in uCi/sec.

L. = the skin dose factor due to beta emissions i for each identified noble gas radionuclides.

Values are listed on Table IV.A.1, in 3

mrem /yr per'uCi/m .

-8 .3 (X/Q)' = 9.97Ex 10' sec/m ;-the highest calculated-s : annual average relative. concentration from the stack releases for any area at.or'beyond the SITE BOUNDARY.

B = the constant for.each identified' noble gas i radionuclides accounting for-the gamma radiation from the elevated. finite' plume.

The constants'were developed using MARE program with plant specific' inputs for.

PBAPS.. Values are listed on' Table IV.A.1, fin' mrad /yr per uCi/sec.

M = the air dose factor due to gamma emissions.

i for each. identified noble gas radionuclides.

Values are listed on Table ~IV.A.1, in mrad /yr 3

L pert uci/m .

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

I u

\-

R v. 3 TABLE IV.A.1'- Constants for Isotopic Analysis Method (corrected for decay during transit)

' Total Plume-Body Skin Gamma Beta Air Body Plume-Air Dose Dose Air Dose Dose- -Dose Dose Factor Factor Factor. Factor Factor Factor B K L M N V i i i i i i-(mrad /yr- (mrem /yr (mrem /yr (mrad /yr (mrad /yr (mrem /yr per per per . per per per 3 3 .. 3 3 Radionuclides uC1/sec)' uCi/m ) uCi/m ) uCi/m ) (uCi/m-) uCi/sec)

Kr-85m 4.02E-05 1.17E+03 1.46E+03 1.23E+03 1.97E+03 3.76E-05 Kr-87 1.74E-04' 5.92E+03 9.73E+03 6.17E+03 1.03E+04 1.66E-04 Kr-88 4.90E-04 1.47E+04 2.37E+03 1.52E+04 2.93E+03 4.72E-04 Xa-133 1.19E-05 2.94E+02 3.06E+02 3.53E+02 1.05E+03 1.llE-05 Xo-133m 1.09E-05 2.51E+02 9.94E+02 3.27E+02 1.48E+03 1.01E Xe-135 6.37E-05 1.81E+03 1.86E+03 1.92E+03 2.46E+03 5.95E-05 Xe-135m 6.61E-05 2.53E+03 5.76E+02 2.72E+03 5.99E+02 6.17E-05 Xe-138 1.52E-04 7.33E+03 3.43E+03 7.54E+03 3.94E+03 1.46E-04 The values K , L , M , and N are taken from Reg. Guide 1.109, i i i i Tablo B-1. The values B and V were developed using the MARE i i program with plant specific inputs for PBAPS.

I R v. 3 l

l l 2. Iodine-131, iodine-133, tritium and radioactive I materials in particulate form, other than noble gases, with half-lives greater than eight days.

The dose rate shall be determined by either of two methods. Method (a), the Iodine-131 Method, uses'the iodine-131 releases and a correction factor to calculate the deae rate from all nuclides released. Method'(b),:the' Isotopic.

Analysis Method, utilizes all applicable nuclides.

For normal operations, it is expected that Method (a) will be used since iodine-131 dominates the-critical pathway thyroid. However, in the-event iodine-131 releases are minimal (e.g.,

during long term shutdown) Method (b) will be used to provide accurate calculations. In the absence of iodine-131 releases, the lung is the critical organ.

a. Iodine-131 Method D =

(CF) P W h +W h T I S IS V IV where:

The location.is the site boundary, 1097m SSE from the vents.

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

T CF = 1.09; the correction factor accounting for the use of iodine-131 in lieu of all radionuclides released in gaseous effluents including iodine-133.

7 3 P = 1.624 x 10 mrem /yr per uCi/m ; the dose I parameter for I-131 via the inhalation pathways. 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).

-7 3 W = 1.03 x 10 sec/m ; the highest calculated S annual average relative concentration for any area at or beyond the SITE BOUNDARY from

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ a

E L 1 Rav.I 3

.I stack releases. .(SSE boundary)

I Q =.the release rateLof iodine-131'in gaseous -j IS- effluents from the. stack determined by the . .

+

effluent sampling and' analysis program-(Technical Specification Table:4.8.2)-in uCi/sec.

-7 3 W. ='4.78 x 10 sec/m ;.the. highest 1 calculated v annual. average relative concentration for any-area at or beyond the SITE BOUNDARY-for all vent releasesE(SSE' boundary) l =

Q = the release rate of. iodine-131 in' gaseous IV effluents.from all vent releases, determined by the effluentcsampling and analysis program (Technical Specification Table 4;8.2)-in uCi/sec.

b. ' Isotopic Analysis' Method D

L

_=[P~i W S

. hhiS W

V h

iV where:

The' location is the site boundary, 1097m SSE from the vents.

D = dose rate to the lung, in mrem /yr.

L P = the dose parameter for radionuclides other i than noble gases for the inhalation pathway.

The dose factors are based on the critical individual organ-lung, and most restrictive age group-child. All values are from Reg.

Guide 1.109 (Tables E-5 and E-9). Values are listed on Table IV.A.2, in mrem /yr per 3

uCi/m .

-7 3 W = 1.03 x 10 sec/m ; the highest calculated S annual average relative concentration for any area at or beyond the SITE BOUNDARY from stack releases. (SSE boundary)

Rev. 3 Q = the release rate of radionuclides; i, in gaseous

} is effluents from the stack determined by the effluent sampling and analysis program (Technical i Specification Table 4.8.2) in uCi/sec.

-7 3 W = 4.78 x 10 sec/m ; the highest calculated v annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases. (SSE boundary) b = the release rate of radionuclides, i, in gaseous iv effluents from all vent releases, determined by the effluent sampling and analysis program (Technical Specification Table 4.8.2) in uCi/sec.

- Rav. '3 TABLE IV.A.2 - CONSTANTS FOR' ISOTOPIC ANALYSIS METHOD 3

(mrem /yr. per.uCi/m )

PI - Inhalation Radionuclides. Lung Dose Factor

'6

.Mr-54 1.58x10 4:

Cr-51 1.70x10 6

.Co-58 1.11x10 6

Co-60 7.07x10' 5-Zn-65 9.95x10 6

Sr-89 '2.16x10 7-

Sr-90 1.48x10-5 Ce-141 5.44x10 5

Cs-134 1.21x10 5-Cs-137 1.04x10-6

'Ba-140 1.74x10 1

Rev.-3 IViB' Surveillance' Requirement 4.8.C.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 air dose shall be determined by either:of two methods.

' Method (a),;the Gross-Release Method, assumes.that all noble 1gasesLreleased are the.most-limiting.nuclide - Kr-88' for~ gamma radiation and Kr-87 for. beta' radiation. Method (b),-the Isotopic Analysis Method, utilizes the results of Enoble gas analyses. required by specification 4.8.C.la..

For normal operations, it.is expected that Method;(a) will' be.used.. Bowever,.if noble gas releases are close to the-limits as calculated by Method.(a), Method (b) can be used.

to allow more operating flexibility by using data that more-

. accurately reflect actual releases. ,

1. for gamma radiation:

a) Gross Release Method

-8 D

g = 3.17 x 10 (M (X/Q) D' + B6) v v s where:

The location is the SITE BOUNDARY 1097m SSE from the vents. This location results in the highest calculated gamma air dose from noble gas releases. ,

Dy = gamma air dose, in mrad.

-8 3.17 x 10 = years per second.

4 3 M = 1.52 x 10 mrad /yr per uCi/m ; the air dose factor due to gamma emissions for Kr-88. (Reg Guide 1.109, Table B-1)

-7 3 (X/Q) = 5.33 x 10 sec/m ; the highest calculated V annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

i .

3 Rsv. 3 jd = the1 gross release of noble gas L v 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.

-4 B = 3.15 x'10 mrad / year per uCi/sec;-the constant.for Kr-88 accounting for the gamma radiation.from the. elevated-finite plume.

The constant was developed using the~ MARE program'with plant' specific inputs for PBAPS.

Q- = the gross' release of noble gas S radionuclides >in gaseous releases'from the stack determined by gross activity-stack monitor in uCi. Releases shall be cumulative.over the calendar quarter:or year as' appropriate.

b) Isotopic Analysis' Method D =-3.17 x 10 [.

i M

i (X/Q) v Q

iv

+B i

Q is where:

The location is the SITE BOUNDARY, 1097m SSE from the vents. This location.results in the highest calculated) gamma air dose from noble gas releases.

Dy = gamma air dose, in mrad.

-8.

3.17 x 10 = years per second.

M = the air dose factor.due to. gamma emissions i for each identified noble gas radionuclides.

Values are listed on Table IV.A.1, in mrad /yr 3

per uCi/m .

Rev. 3 ,

1

-7 3 (X/Q) = 5.33 x 10 sec/m ; the highest calculated V average relative concentration from vent l releases for any area at or beyond the SITE BOUNDARY.

6 = the release of noble gas radionuclides, i, iV 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 = the constant for each identified noble gas i radionuclides accounting for the gamma radiation for the elevated finite. plume.

The constants were developed using the MARE program with plant specific inputs for PBAPS.

Values are listed on Table IV.A.1, in mrad /yr per uCi/sec.

6 = the release of noble gas radionuclides, i, is in gaseous effluents from the stack determined by isotopic analysis, in uCi. Releases shall be cumulative over the calendar quarter or year, as appropriate.

2. for beta radiation:

a) Gross Release Method

-8 D = 3.17 x 10 N (X/Q) D + (X/Q) D h v v s s where:

The location is the SITE BOUNDARY 1097m SSE from the vents. This location results in the highest calculated gamma air dose from noble gas releases.

D = beta air dose, in mrad.

-8 3.17 x 10 = years per second.

0 Rnv. -3

h.

Q 4 -3 N:

= 1.03 x.10. mrad /yr per uci/m.; the. air dose factor'due to beta emissions for

'Kr-87. (Reg. Guide 1.109, Table B-1) s ,

3

- ( X/0 ) - ' = 5.33 x110' 'sec/m ;:the' highest: calculated-

-v' annual average relative concentration'from, vent' releases for any area at or beyond the SITE BOUNDARY.

D. = the gross release of noble gas-v 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.

-8 3

= 9.97 x 10 sec/m ; the highest: calculated

.(X/Q) ' annual: average relative concent' ration-from s

the stack releases for any area at or"beyond the SITE BOUNDARY..

D = the gross release of noble gas a

radionuclides in' gaseous. releases from the stack determined by gross activity stack monitors, in uCi. Releases shall be cumulative over the calendar quarter or year, as appropriate.

b) Isotopic Analysis Method .

= 3.17 x 10 [. N i (X/Q) Q. + (X/Q)- Q D

i v iv s is

-8 3.17 x 10- = years per second.

N = the air dose factor due to beta i emissions for each identified noble gas radionuclides. Values are listed 3 on Table IV.A.1, in mrad /yr per uCi/m .

1

[

c -- _-_

> n 3 i' g,

) 'Rev. 3'

~.

p

-7. .

3-(X/Q)' = 5.33 x 10 'sec/m ; the highest calculated' v' annual; average relative concentration'from:

- vent releases for any area-at.or beyond the SITE BOUNDARY.

p .

l 0 = the release-of noble gas' radionuclides,;i, iv in gaseous effluents'from all vents as determined by-isotopic analysis, in-uC1.

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

-8. 3

.(X/Q)' =

9.97lx.10 sec/m the highest calculated s . annual' average relative concentration from the stack releases for any area at or beyond the SITE BOU*dDARY.

D- = the releasefof noble' gas radionuclides, i, is in gaseous effluents from the stack as determined by' isotopic' analysis, in uC1.

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

IV.C' ' Surveillance iRequirement 4.8.C.3 k

The doseto'an' individual from iodine-131, iodine-133, tritium and radioactive materials in particulate form and

, radionuclides-other-than noble gases with half-lives greater thaneight days'in gaseous effluents released to

-areas.at and-beyond the SITE BOUNDARY.

The dose shall be determined by one of two methods.- Method

.(a), the Iodine-131 Method, uses the iodine-131 releases and a correction factor to calculate the dose from all nuclides' released.- Method (b), the Isotopic Analysis Method, utilizes all applicable nuclides.

For. normal operation, it is expected that Method (a) will be used since iodine-131 dominates the critic,al pathway -

thyroid. However, in the event iodine-131. releases are minimal (e.g. during-long term shutdown) Method (b) will be used to provide accurate calculations. In the absence of ,

iodine-131 releases, the liver is the critical organ. 1 1

.-___ __ ____ ______ _ w

~

1 p1 4 4- . Rev. 3 II~

-sh ' Iodine'-f131' Method-

. a

]

-8 D ='3.17:x 10 (CF)'(0.5)R W6 +WQ T-S IS v IV where:

Location is the! critical' pathway dairy l2103m SSW from vents.

D = critical. organ dose, thyroid,1from all T' pathways, in mrem. -

-8 3.17 x 10- .= years.per second.

CF = 1.09;:the correction factor accounting for.

.the use of. Iodine-131.in lieu of all radio-nuclides released in gaseous effluents: including Iodine-133..

0 ~. 5 =-fraction of iodine releases which are nonelemental.-

11 2 R = 3.08 x 10 m (mrem /yr) per uCi/sec; the dose factor for iodine-131. The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, infant. See Site Specific Data.**

~10 -2 W = 4.95 x 10 -meters ; '(D/Q).for the food s pathway for stack releases.

6 = the release of iodine-131 from the stack-IS determined by the effluent sampling and analysis program (Technical Specification Table 4.8.2),.in uCi. Releases shall be cumulative over the calendar quarter or year,.as appropriate.

-9 -2 1( . = 1.14 x 10 meters ;.(D/Q) for the food v- pathway.for vent releases.

6 = the release of iodine-131 frcm the vent

'IV determined.by the effluent sampling and

- - - ~ -

g p i .; ,

< Rsv. 3 y ' analysis. program (Technical ~ Specification.

Table 4.8.2),~in'uCi. Releases shall be-

. cumulative over'the calendar. quarter or' year, as appropriate.

- LSne NoteL2'in Bases.

b. Isotopic Analysis Method -

L N D =-3.17 x'10 -/ R WQ +W6, i i S iS 'v IV-

[l 'where:

' Location is the~ critical. pathway dairy 2103m'SSW from

. vents._

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

< -8 3.17 x 10 = years per second.-

?R = the dose factor'for each identified _

i radionuclides, 1, based on the critical' individual organ, liver and most restrictive; age group, infant.

2 Values are listed on Table'IV.C.1, in M-(mrem /yr) per uCi/sec.

-10 -2 W' = 4.95 x 10 -meters ; (D/Q) for the food s pathway for stack releases.

6 = the release of radionuclides,-i, in gaseous is effluents from the vents determined by the effluent sampling (Technical Specification Table 4.8.2), in uCi. Releases shall be camulative over the calendar quarter or year, as appropriate.

-9 -2 W = 1.14 x 10 meters ; (D/Q) for the food v for vent releases.

D' = the release of radionuclides, i, in gaseous iv effluents from the vents' determined by the effluent sampling and analysis program (Technical Specification Table 4.8.2) in uCi. Release shall be cumulative over the calendar quarter or year, as appropriate.

1 1

I

4 I

Rev. 3 TABLE ' IV.C.1 -- CONSTANTS FOR ISOTOPIC ANALYSIS METHOD ,

2 !

(m-(mrem /yr) per uCi/sec)

Rrdionuclide' _R I 7-Mr-54. 1.14x10 4

Cr-51 4.72x10.*

7-

..Co-58 2.13x10 7

Co-60 2.58x10 9

Zn-65 5.56x10

-8 Sr-89 1.06x10

9 Sr 9.06x10.*

3

'Co-141 7.73x10 10

~Cc-134 1.99x10 10

'Cs-137 1.76x10 4

Ba-140i 7.04x10

.* There.is no liver dose fa'ctor.given in R.G.1.109'for these nuclides. .Therefore, the.whole body does factor was used.

-h i

l i 1 _

_ _ __._-.______.._m__m__.._-_. .M

Rev.73 i

V.DL Surveillance' Requirement 4.8.C.5a The projected doses:from releases of' gaseous effluents.to' I areas at and beyond the SITE BOUNDARY.shall be calculated in accordance with'the-following sections-of'this manual: f l

a. ~ gamma air. dose - IV.B.1

b. beta' air. dose - IV.B.2-

c. organ. dose - IV.C The projected dose calculation shall be based on expected trelease from plant. operation. The normal release pathways result in the maximum releases from the plant. . Any alternative release pathways result in lower _ releases and therefore lower doses.

l l

i l

i

_ _ _ _ .________-____a

Revo 3 IV.E Surveillance Requirement 4.8.C.6.b l l

1. The three types of recombiner hydrogen analyzers used l at Peach Bottom are:

a. Hays Thermal Conductivity type (Analyzers 20S192L, 20S192H, 20S222, 20S223, 30S192L, 30S192H, 30S222, 30S223)

b. Scott Helium-Immume type (Analyzers 30S222 and 30S223)

c. Exosensor Helium Immume (20S192L and 20S222).

2. The calibration gases for the two types are:

a. Hays Analyzers and Exosensor Analyzers Zero Gas - Air Calibration Gas - 4% Hydrogen, Balance Nitrogen 1% Hydrogen, Balance Nitrogen

b. Scott Analyzers Zero Gas - Air Calibration Gas - 2% Hydrogen, Balance Air

Rev. 3 V.A Surveillance Requirement 4.8.D If the doses as calculated by the equations in this manual do not exceed the limits given in Technical Specifications 3.8.B.2, 3.8.C.2, or 3.8.C.3 by more than two times, the conditions of Technical Specification 3.8.D have been met.

If the doses as calculated by the equations in this manual exceed the limits given in Technical Specifications 3.8.B.2, 3.8.C.2, or 3.8.C.3 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 Reactor 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 j accordance with Technical Specification 3.8.D. l The cumulative dose contribution from direct radiation from the two reactors at the site and from radwa'ste storage shall be determ.ined by the following methods:

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 by surveillance requirement 4.8.D)

This evaluation is in accordance with ANSI /ANS 6.6.1-1979 Section 7. The error using this method is estimated to be l approximately 8%.

VI.A Unique Reporting Requirement 6.9.2.h.(3) Dose Calculations for the Radiation Dose Assessment Report The assessment of radiation doses for the radiation dose

)

assessment report shall be performed utilizing the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, 1 October 1977. Any deviations from the methodology provided

! in Regulatory Guide 1.109 shall be documented in the radiation dose assessment report.

l

Rev. 3

-l The meteorological conditions concurrent with time 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.

l The Radiation Oose Assessment Report shall be submitted l within 120 days after January 1 of each year in order to )

allow time for the calculation of radiation doses following publication of radioactive releanes in the Radioactive Effluent Release Report. There is a very short turnaround time between the determination of all radioactive releases and publication of the Radioactive Effluent Release Report.

This would not allow time for calculation of radiation doses in time for publication in the same report.

VII.A Surveillance Requirement-4.8.E The radiological environmental monitoring samples shall be collected pursuant to Table VII.A.1 from the locations shown on Figures VII.A.1, VII.A.2 and VII.A.3 and shall be analyzed pursuant to the requirements of Table VII.A.l.

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Rev. 3.

VIII. BASES-Site Specific Data Note 1: Liquid dose factors,1A , for section.III.B were' developed

.i using'the following site specific data. The liquid pathways involved are drinking water and fish. l.

A =

'(U /D +U x BF ) K x DF x RC i w w F .i O i U = liters per year; maximum age group usage of l w drinking water (Reg., Guide 1.109, Table E-5)

D = 5.4; average annual dilv* ion at Conowingo intake w

U =.kg per year; maximcm age group usage of fish I F (Reg. Guide 1.109, Table E-5)

BF = bioaccumulation factor for nuclide, i, in freshwater

i. fish. Reg.. Guide 1.109, Table A-1, except P-32 which uses a'value of 3.0E03 pC1/kg per pCi/ liter.

5 6 3 K a 1.14 x 10=(10 pCi/uCi x 10 ~ ml/l - 8760 hr/yr) l O units conversion factor.

DF = dose conversion factor for nuclide, i, for the age i . group in total body or organ, as-applicable. Reg.

Guide 1.109, Table E-ll, except P-32 bone which uses a value as indicated below. l

-5 3.0 x 10 .

RC = 1.16; reconcentration from PBAPS discharge back through PBAPS intake.

The data for Dw and RC were derived from data published in Peach Bottom Atomic Power Station Units 2 and 3 (Docket Nos. 50-277 and 50-278) Radioactive Effluent Dose Assessment, Enclosure A, September 30, 1976. All other data except P-32 BF and DF were used as given in Reg. Guide 1.109, Revision 1, October 1977. The P-32 BF and DF were used in accordance with information supplied in Branagan, E.F., Nichols, C.R., and Willis, C. A., "The In;portance of P-32 in Nuclear Reactor Liquid Effluents",

_ - _ _ _ _ _ _ ___- _ __ D

Rev. 3 NRC, 6/82. The teen and child dose factors were derived by the ratio of the adult bone dose factors in Reg. Guide 1.109 and Branagan, et al.

Note 2 To develop constant R for section IV.C, the following site specif<c data were used:

RC (D/Q) = K'O (U ) F x r x DFL )a f (1-f ) - Ds t i F ap m i p s e if Y

h+h I w p 6

K' = 10 pCi/uCi unit contersion factor Q = S0 Kg/ day; cow's consumption rate F

U = 330 1/yr; yearly milk consumption by an infant ap

-7 -l

>s = 9.97 x 10 sec decay constant for I-131 i

-7 -1

)s = 5.73 x 10 sec decay constant for removal of w activity in leaf and plant surfnces.

-3 F = 6.0 x 10 day / liter, the stable element transfer m coefficient for I-131.

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

-2 DFL = 1.39 x 10 mrem /pCi - the thyroid ingestion dose factor for I-131 in the infant.

f = 0.6; the fraction of the year the cow is on pasture p (average of all farms) f = 0.513; the fraction of cow feed that is stored feed s while the cow is on pasture (average of all farms).

Rev. 3 2

Y = 0.7 Kg/m - the agricultural productivity of pasture p feed grass.

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

The pathway is the grass-cow-milk ingestion pathway.

These data were derived from data published in Peach Bottom Atomic Power Station Units 2 and 3 (Docket Nos. 50-277 ano 50-278) Radioactive Effluent Dose Assessment, Enclosure A, September 30, 1976. All other data were used as given in Reg. Guide 1.109, Revision 1, October 1977. j Surveillance Requirement 4.8.B.2 Liquid Pathway Dose Calculations Tha equations for calculating the doses due to the actual release rctes 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 Rnactor Effluents for the Purpose of Evaluating Complinace with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and NUREG-0133 " Preparation of Radiological Effluent Technical Sp;cifications for Nuclear Power Plants", October 1978.

Surveillance Requirement 4.8.C.1 Done Noble Gases The cquations 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 of Raactor Effluents for the Purpose of Evaluating Compliance with 10 CPR Part 50, Appendix I", Revision 1, October 1977, l NUREG-0133 " Preparation of Radiological Effluent Technical Sp;cifications for Nuclear Power Plants", August 1978, and the atmospheric dispersion model presented in Information Requested in Enclosure 2 to letter from George Lear to E. G. Bauer dated I Fcbruary 17, 1976, September 30, 1976. The specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric conditions.

The dose due to noble gas release as calculated by the Gross l Release Method is much more conservative than the dose calculated l

l r

L____

1 Rav. 3 by-the Isotopic Analysis Method. Assuming the release rates giv:n in Radioactive Effluent Dose Assessment, September 30, 1976,.the values calculated by the Gross Release Method for total body dose' rate and. skin dose rate are 6.0 times and'5.7 times,

.rocpectively, the values calculated by the Isotopic Analysis M3thod.

-Tha model Technical specification LCO for all radionuclides and.

radioactive materials in particulate from and radionuclides other

'thnn noble gases requires that the instantaneous dose rate be leco than the equivalent of 1500 mrem per year. For the purpose.

'of calculating this instantaneous dose rate, thyroid dose from iodine-131 through the inhalation pathway will be used. Since the operating history to date indicates that iodine-131 releases have had the major dose impact, this approach is appropriate.

Tha value calculated is increased by nine per cent to account for the thyroid dose from all other nuclides. This allows for expedited analysis and calculation of compliance with the LCO. q In-the event that the plant is shutdown long enough so that iodine-131 is no longer present in gaseous effluents, an Isotopic Annlysis Method is available. Since no iodines'are present, the critical organ changes from the thyroid is the lung.

Surveillance Requirement 4.8.C.2 Dosn Noble Gases iThs equations for calculating the doses due to the actual release ratos of radioactive noble gases in gaseous effluents were

davoloped from the methodology provided in Regulatory Guide l 1.109,,'" Calculation of Annual Doses to Man f rom Routine Releases of R: actor Effluents for the~ Purpose of Evaluating Compliance with 10 CPR Part 50, Appendix I", Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical' Specifications for Nuclear Power Plants", August 1978, and the stsonpheric dispersion model presented in Information Requested in Enclosure 2 to letter from George Lear to E. G. Bauer dated February 17, 1976, September 30, 1976. The specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric !

conditions.

The dose due to noble gas releases as calculated by the Gross Relerse Method is much more conservative than the dose calculated by the Isotopic Analysis Method. Assuming the release rates ;

given in Radioactive Effluent Dose Assessment, September 30, 1976, the values calculated by the Gross Release Method for total body dose rate and skin dose rate are 4.3 times and 7.2 times, l

l

Rev. 3 respectively, the values calculated by the Isotopic Analysis M3thod.

Surveillance Requirement 4.8.C.3 Do97, Iodine-131, Iodine-133, itium, and Radioactive Material in Particulate Form l Th3 equations for calculating the doses due to the actual release-ratca of radiciodines, radioactive material in particulate form, cnd radionuclides other than noble gases with half-lives greater l then 8 days were developed using the methodology provided in R gulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of l

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

October 1978, and the atmospheric dispersion model presented in Information Requested in Enclosure 2 to Letter from George Lear to E. G. Bauer dated February 17, 1976, September 30, 1976.

Thesa equations provide for determining the actual doses based upon the historical average atmospheric conditions.

Compliance with the 10 CPR 50 limits for radiciodines, radioactive materials in particulate form and radionuclides other

.than noble gases with half lives greater than eight days is to be determined by calculating the thyroid dose from iodine-131 relocses. Since the iodine-131 dose accounts for 92 percent of ths total dose to the thyroid, the value calculated is increased by nine percent to. account for the dose from all other nuclides.

In the event that the plant is shutdown long enough so that iodine-131 is no longer present in gaseous effluents, an Isotopic Analysis Method is available. Since no iodines are present, the critical-organ changes from the thyroid to the liver.

RW-120 Page 1 of 12, Rev. 0..

RL:kls/ tdt PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM UNITS 2 AND 3 h/f l RW-120 SOLID RADWASTE SYSTEM PROCESS CONTROL PROGRAM 1.0 PURPOSE To provide guidance and boundary conditions for preparation of specific procedures for processing, sampling, analyzing, packaging and shipping ;

solid radioactive waste in accordance with State and Federal regulatory ;

requirements. i 2.0 SCOPE

'This program is applicable'to the Peach Bottom Atomic Power Station solid radwaste processing system. Wastes considered in this scope are filter / demineralized and bead resins, sludge, oil, and aqueous liquids. :

Dry Active Waste is only included as it applies to assurance that packaged waste is suitable for shipment dnd burial in accordance with applicable State and Federal regulations.

3.0 REFERENCES

3.1 49 CFR Parts 170 through 178 3.2 10.CFR Parts 20, 50, 61 and 71 3.3 Standard Review Plan 11.4, Rev. 2, including Branch Technical Position ETSB 11-3, Rev. 2 3.4 Low Level Waste Licensing Branch; Technir=1 Position on Radioactive Waste Classification, Rev. 0 (5/83) )

3.5 Low Level Licensing Branch; Technical Position on Waste Form, Rev. 0 (5/83) 3.6 General Criteria for high Integrity Containers (SCDHEC) dated 10/22/80 3.7 South Carolina Department of Health and Environmental Control Radioactive Material. License No. 097, Amendment No. 44. (Barnwell Facility) 3.8 Barnwell Site Disposal Criteria, (Chem. Nuc' lear Systems, Inc.)

November 1982.

3.9 State of Washington Radioactive Materials License No. WN-1019-2, Amendment No. 17 (Richland Site).

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RW-120 Page 2 of'12, Rev. 0 3.10 Transportable Modular AZTECH Plant Licensing Topical, NEDE-30878, i January 1985.

.3.11 NUPAC Dewatering Topical Report, No. TP-02-P-A, Rev. 1 3.12 Station Radwaste Procedures (RW).

3.13 ' Station Chemistry Procedures (CH).

-3.14 Station Surveillance Test Procedures (ST).

3.15 Station Routine Test Procedures (RT).

3.16 PBAPS Quality Assurance Plan, Volume III, Operational Phase (HP&C).

3.17 NQA Audit Section Instructions (ASI-01), Subsection Radioactive. Waste Material ~ Instruction for Audits of Radioactive Waste Material Activity.

3.18 Safety Evaluation for Oil Decontamination, MOD 1259A, Pev. 1.

i.

3.19 Safety Evaluation for MOD 1750A, Rev. 3, Radwaste Dewatering Facility Peach Bottom Atomic Power Station.

3.20 Chem-Nuclear Systems, Inc. Mobile Cement Solidification System Topical Report, CNSI-2 (4313-01354-01P-A) ( ,

3.21 Chem-Nuclear Systems, Inc. 10CFR61 Waste Form Certification -- Cement (WF-C-01-NP).

3.22 Upd'ated Final Safety Analysis Report Peach Bottom Atomic Power Station Units 2 and 3.

3.23 PBAPS Technical Specification 3.8.F' 4.0 DEFINITIONS 4.1 BATCH An isolated quantity of feed waste to be processed having essentially H constant physical and chemical characteristics. For the purposes of resin type waste, a batch is defined as one full volume of phase separator.

4.2 OPERABLE f~ A system, subsystem, train, component or device shall be operable or have operability when it 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 cquipment 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).

RW-120 L Page-3 of 12 Rev._0-l 4.3 PROCESSING .

l Changing, modifying, and/or packaging'the commercial nuclear power plant generated wet radioactive waste into a form that is acceptable !

to a disposal' facility. l

.4.4 QUALITY ASSURANCE / QUALITY. CONTROL i I

As used in this document, " quality assurance" comprises all those-planned and systematic actions necessary to provide adequate confidence that a structure, system, or_ component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions'related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements.

4.5 SAMPLING PLAN i

A sampling program implemented to ensure that representative samples from'the feed waste and the final waste form are obtained and tested for conformance with parameters stated in the PCP and waste form acceptance criteria.

4.6 LOW-LEVEL RADI0 ACTIVE WASTE (LLW)

Those low-level radioactive wastes containing source, special .

nuclear, or by-product material that are acceptable for disposal in a land disposal facility. For the purposes of this definition, low-level radioactive waste has the same meaning as in the Low-Level Waste Policy Act, that is radioactive waste not classified as high-level radioactive waste, transuranic waste, spent nuclear fuel, or by-product material as defined in section 11e.(2) of the Atomic Energy Act (uranium or thorium tailings and waste).

4.7 WASTE CONTAINER A vessel of any shape, size, and composition used to contain the '

final processed waste.

4.8 WASTE FORM Waste in a waste container acceptable for disposal at a licensed disposal facility.

4.9 SLUDGE Contaminated liquid which may contain water, oil, and suspended solids, including ion exchange media and other filters.

RW-120 Page 4 of 12, Rev. 0 I

4.10 LOW LEVEL WASTE STORAGE FACILITY (LLWSF)

A station facility designed to store radwaste for Peach Bottom Atomic Power Station.-

L 5.0 RESPONSIBILITIES NOTE THE STATION OPERATING ORGANIZATION IS OUTLINED IN SECTION 13.2 0F THE UFSAR.

5.1 Nuclear Quality Assurance is responsible for establishment of a Quality Assurance Program. The Nuclear Quality Assurance Audit Section performs audits to ensure compliance with the Quality Assurance Plan. '!

5.2 All PECo and PECo contract personnel are responsible for implementation of procedures and good practices so as to provide Quality Assurance and maintain exposures ALARA. .

5.3 The Engineer-Radwaste is responsible for:

5.3.1 Compliance with this Process Control Program.

5.3.2 Record keeping and document control of shipping and (

processing data.

~

5.3.3 Assuring that radwaste personnel are appropriately trained and qualified.

5.3.4 Providing radwaste personnel for training, as required.

5.4 The Superintendent-Operations is responsible for: ,

5.4.1 Providing trained personnel to operate appropriate radwaste process equipment. ,

5.4.2 Defining those Operations positions which require training.

5.5 The Superintendent-Training is responsible for:

5.5.1 Development and implementation of performance-based training for designated personnel utilizing the Training Systems Development (TSD) model in accordance with Training Division procedures.

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RW-120 Page 5 of 12, Rev. 0 6.0 PROCESSING !

6.1 WASTE TYPES 6.1.1 Condensate Filter / Demineralized Waste

a. The; contaminated waste product generated by the backwash of the condensate filter demineralizers,

b. Waste consists of contaminated powdered ion exchange resins at varying degrees of exhaustion, fibrous filter media, and small concentrations of various solids, activated and non-activated corrosion products and fission products.

6.1.2 Radwaste Demineralized Waste

a. The contaminated waste product generated by the backwash of the liquid radwaste filters and demineralizers, collected in the waste sludge tank. -l

b. Waste consists of contaminated powdered ion exchange resins and bead resins at varying degrees of exhaustion, fibrous filter media, carbon overlay material along with sman concentrations of various solids, activation procacts, fission products, and both contaminated and non-contaminated corrosion products.

6.1.3 Reactor Water Cleanup Filter / Demineralized Waste

a. The contaminated waste product generated by the backwash of the Reactor Water Cleanup filter demineralizers.

b. Waste consists of contaminated powdered ion exchange resins at varying degrees of exhaustion, fibrous filter media, and small concentrations of various solids, corrosion products, activation products, and fission products.

6.1.4 Oils / Sludges-Wet Radioactive Wastes

a. Oils consist of non-reclaimable contaminated oils and grease of various grades both synthetic and natural, in free form or containing various amounts of solid material. Sludge consists of sump wastes, filter solids, ion exchange resins, and strainer solids and other wet solids not conforming to the Process Control Program for conventional dewatering techniques.

l

i RW-120 Page 6 of 12, Rev. 0 l

6.2' PROCESS' DESCRIPTION 6.2.1 Condensate Filter / Demineralized Waste j a.- A condensate filter-demineralized backwash consists'of i approximately 9,000 gallons of slurry with an average of i 350 lbs. (dry wt.) spent resins and crud. !

b. Backwashes are collected and settled.in a condensate phase separator.. Clarified liquid is decanted until sufficient volume of settled sludge is obtained for '

processing.

c. Phase separator contents are mixed before processing.

d. The slurry is fed to the NUPAC Dewatering System for i dewatering.

e. Slurry input to the NUPAC Dewatering System is transferred to the system from the phase ~ separators via the sludge discharge pumps.

6.2.2 Radwaste Demineralized

a. 'A backwash from a radwaste filter consists of -

approximately 1,925 gallons of slurry, with an average (

of 60 lbs. (dry wt.) spent resins and crud. Filters may contain charcoal overlay for oil removal,

b. A backwash from the radwaste deep bed demineralized consists of approximately 1,500 gallons of slurry, with an average of 3,200 lbs. (dry wt.) spent resins.

c. Backwashes from radwaste filter demineralizers, radweste deep bed demineralizers, and fuel pool filter-demineralizers are collected in the Waste Sludge Tank until approximately 13,500 gallons of liquid is.

accumulated.

d. Waste Sludge Tank contents are mixed before processing. !

During normal operation the waste sludge tank contains only the radwaste and fuel pool filter demineralized backwashes. :

e. Waste sludge tank contents are sent to a condensate phase separator for processing in accordance with Section 6.2.1. From the condensate phase separators, waste is processed using the NUPAC Dewatering System.

l 1

r L

RW-120 Page 7 of 12, Rev. 0

(

6.2.3 Reactor Water Cleanup Waste

a. A RWCU filter-demineralized backwash consists of approximately 1,100 gallons of slurry with an average of 35 lb. (dry wt.) spent resins.

b. Backwashes are collected and settled in a phase separator. Clarified liquid is decanted until sufficient volume of settled sludge is obtained for processing.

c. Phase separator contents are mixed before processing.

d. The slurry is fed to the NUPAC dewatering system for dewatering.

e. Slurry input to the NUPAC dewatering system is transferred to the system from the phase separators via the sludge discharge pumps.

6.2.4 Oils / Sludges

a. Sludges and oils generated during operation and maintenance are collected in containers in appropriate approved areas throughout the plant. The filled and labeled containers are sealed and moved to available areas for temporary staging.

b. Sludges / oils may be decontaminated to below station free release limits and processed as non-radioactive' waste.

c. Sludges / oils may.be solidified using a vendor supplied mobile solidification unit, in accordance with the system's topical report.

l 6.2.5 Miscellaneous Waste

a. Certain wet wastes may be allowed to dry and be treated as dry active waste.

6.3 PROCESS CONTROL 6.3.1 Resin processing may be performed either by the NUPAC Dewatering System or through solidification.

6.3.2 The NUPAC dewatering system processes resin by using a pump to remove the bulk of free water and blowing the recirculating air through the waste container and water separator to facilitate drying of the resin (Figure 1).

a. Applicable station Radwaste Procedures shall be observed. Successful completion of applicable portions

_..-_._-__.-.______--__m

7 j ..

p RW-120 Page 8 of 12, Rev. O of the vendor PCP shall serve as an indicator of system I operability.

b. Resins are fed to and packaged in liners. When High.

Integrity Containers are used, they shall be approved by ,

the South Carolina Department of Health and .

Environmental Control. l

c. .The NUPAC Dewatering System shall be operated-in accordance with the system's Topical Report and-applicable RW procedures.

d. Station to NUPAC System interphasing is addressed in the MOD 1750A, Rev. 3. Safety Evaluation. j 6.3.3 Resins may be solidified using a vendor' supplied mobile solidification system. Processing may be performed in a location either within the protected area or in the low level waste storage facility upon completion of an appropriate safety evaluation. Processing shall be accomplished in accordance with vendor's system topical report.

a. Applicable station Radwaste Procedures shall'be observed. Successful completion of applicable portions of the vendor PCP shall serve as an indicator of system .

operability. (

b. Station to system interphasing shall be covered by the appropriate safety evaluation and procedures.

6.3.4 Sludges and oily wastes may be solidified using a mobile solidification system. Processing may be performed in a location either within the protected area or in the low level.

waste storage facility upon completion of an appropriate safety evaluation. Processing shall be accomplished in accordance with the vendor's system topical report.

6.3.5 Decontamination of oil may be performed on-site using a vendor service. The service shall be performed in accordance with the applicable Safety Evaluation.

6.4 PRODUCT CONTROL 6.4.1 A sample from each batch is analyzed quantitatively for activity and isotopic identity. If radionuclides distributions are shown to be consistent between similar batches, consideration may be given to decreasing the frequency of routine measurements. This constitutes routine sampling.

6.4.2 Scaling factors for nuclides which are hard to identify are established for waste by analysis through an off-site vendor.

Frequency of sampling is normally on an annual basis. This V _ _ _ _ . - . _ . - - - _ - _ - _ _ . _ _ _ _ _ _ - . _ _ _ _ -

kb '

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. RW-120;

.Page 9 of'12, Rev..OL '

f k)

L Lfrequency:should.be raised or lowered-based upon consideration of waste stream or radionuclides .

characteristics. Factors which would influence this consideration include.the frequency of process vessel chan'ge

.out or. waste shipment, the difficulty (e.g.' costs.

occupational exposures) in.' obtaining a; representative. sample of a particular waste ~ stream, the variability;of the radionuclides distribution within.the waste stream over time, and-the availability of analytical' capability for particular1 radionuclides. .If radionuclides distributions are shown to.be-consistent-between similar. batches, consideration may be given to decreasing the frequency-of routine measurements.

If onsite samples.show a' variation from presently'used scaling factors by more than a factor of 10, samples will be sent offsite for' analysis to establish new scaliy' factors.

6.4.3 Radionuclides concentrations and classification of future waste shipments are expected to be similar to those of-previous shipments'for each specific waste stream. Values are available from shipment manifest forms.

-6.4.4 Administrative; controls for' preventing unsatisfactory waste-forms from being released for shipment are described in applicable station procedures.

6.4.5 Solidified or absorbed radioactive wastes shall have no .;

' detectable free standing liquids. . No detectable free standing liquid shall be defined as low as practical but not

~

.more than 1% of the volume of the waste when~the wasteLis.in ,

a disposable container designed to ensure stability or not1 l more than 0.5% of the volume of the waste for waste processed '

in any other container. This is ensured using appropriate station procedures.

6.4.6' Processed resin shall be sampled in accordance with regulatory guidance and~ applicable station procedures to _

verify that.the free liquid content of the_ packaged product is within limits established by' applicable regulatory agencies. Sampling and measurement of free liquid content shall be performed whenever process changes occur that may.

significantly alter system performance,.until compliance with 1 moisture content limits under these conditions can be demonstrated.

6.4.7 Each waste shipment shall be accompanied by a shipping i manifest giving a physical description of the waste, the '

volume, the radionuclides identity and quantity, the total radioactivity, the principal chemical form and waste class, based on 10CFR61.55.

6.4.8 Radwaste Solidification Systems shall be operated within y parameters identified in prequalified testing to ensure acceptable waste form.

~

RW-120 Page 10 of 12, Rev. 0 6.4.9 Sufficient analyses shall be performed to verify that the i quality of waste forms prepared for disposal by vendor's on-site processing shall be similar to vendor's test results.

6.5 TRAINING 6.5.1 Processing of solid radioactive waste shall be performed by qualified and trained personnel. Training records for operators of mobile vendor processing units shall be maintained by the Engineer Radwaste while the vendor is active on-site. These records shall be sent to the Nuclear Records Management System after the vendor has completed work on-site.

6.6 PROCEDURE CONTROL 6.6.1 On-site processing of radioactive waste shall be performed in accordance with approved station procedures.

6.6.2 Processing of radioactive waste by on-site vendors shall be performed in accordance with applicable Process Control Programs, procedures, and applicable NRC guidance.

6.6.3 Procedures for processing, containerization, and transport of wastes shall ensure that specific DOT, 10CFR and burial site requirements are satisfied. (

6.6.4 Process Control Programs (PCP) for specific radwaste systems supplied by vendors for on-site processing shall be presented to PORC for review prior to use of the system.

7.0 RECORDS 7.1 Waste classification records, waste form records, and other records required for the preparation of the semiannual Radioactive Effluent Release Report shall be prepared and retained in accordance with the requirements of 10CFR20, 10CFR71, 49CFR170-178, and Peach Bottom Technical Specifications.

7.2 Records of processing data, test and analysis results, and results of training, inspection, and audits are retained in accordance with the PBAPS Quality Assurance Plan and applicable station Administrative procedures.

7.3 Sufficient documentation shall be maintained to demonstrate compliance of solid radwaste processing with this Process Control Program.

8.0 QUALITY ASSURANCE 8.1 Quality Assurance shall be maintained as defined by the PBAPS Quality assurance Plan, Volume III. The QA Plan shall ensure compliance with NRC and burial site criteria for waste classification and waste form.

RW-120 ;

Page 11 of 12, Rev. 0 ]

Audits are conducted in accordance with NQA Audit Section )

Instructions (ASI-01).

8.2 The Topical Reports of vendor supplied radwaste processing systems shall. undergo review either by the Radwaste Shipping Supervisor or Senior Engineer-Radwaste (or designee). The review shall ensure the vendor supplied systen will be compatible with plant operations and that the Topical Report has been submitted to the NRC for review.

The review shall be documented by a memo addressed to file.

8.3' Audits of a sampling of implementing procedures'shall be performed at least once every 24 months. Procedures should be reviewed to ensure continual compliance with the requirements and process parameters of this Process Control Program. Radioactive wastes not described within this document must be evaluated ~and approved for inclusion in this Proces's Control Program or in a vendor Process Control Program prior to processing.

9.0 REVISIONS 9.1 Any changes to the Solid Radwaste Process Control Program shall be approved by the Plant Operations Reyiew Committee (PORC) and submitted to the Nuclear Engineer-In-Charge of Licensing, Nuclear Support Division, for input to the Semi-Annual Radioactive Effluent Release Report.

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l' 110 CFR:50.36s(a)(2)

T.S. 6.9;.2.h(2);

PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET

,4, P.O. BOX 8699 PHILADELPHIA A. PA.19101 1215)8414000 February 24, 1989 Docket No. 50-277 50-278 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk

' Washington, D.C. 20555

SUBJECT:

Semi-Annual Effluent Releases Report No. 26 July 1, 1988 through December 31, 1988 Peach Bottom Atomic Power Station Unit Nos. 2 and 3

Dear Sir:

Enclosed are two copies of the Semi-Annual Effluent Releases Report'No. 26, July 1 through December 31, 1988 for Peach Bottom i Atomic Power Station Unit Nos. 2 and 3.

l This report is being submitted in compliance with 10 CPR l 50.36a(a)(2) and the-Technical Specifications of Operating Licenses

.DPR-44 and DPR-56, and to fulfill the requirements of Regulatory Guide <

-10.1.

During the report period, revisions were made to the'Offsite !

Dose Calculation Manual (ODCM) and to'the Solid Radwaste System Process' Control Program (PCP). The previous submittal of the PCP was identified as revision 4.. - However,.the PCP has been reformatted and- !'

incorporated into the Peach Bottom Radwaste (RW) Procedures.

Therefore, the revision number has been reset to zero (0) to' reflect the reformatting. Two copies of the ODCM, Revision 3, and PCP, i t . Revision 0, are enclosed with this report.

Very truly yours, o .

G. A. Hunge , Jr Director Licensing Section Nuclear Support Division Enclosures cc: W. T. Russell, Administrator, Region 1, USNRC T. P. Johnson, Senior Resident Inspector

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Text

3.0 REFERENCES

SUBJECT:

Dear Sir:

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