Text

Rev 7 to Odcm
	ML20056G011
Person / Time
Site:	Palo Verde
Issue date:	03/21/1993
From:	; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	;
Shared Package
ML17310A560	List: ML17310A559; ML20056G010; ML20056G011;
References
	PROC-930321, NUDOCS 9309010154
	Download: ML20056G011 (150)
	v • d • e

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APPENDIX D Offsite Dose Calculation Manual (ODCM), i Revision 6 and 7 1

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=*i OFFSITE DOSE CALCULATION MANUAL PALO VERDE NUCLEAR GENERATING STATION UNITS 1,2 AND 3 PSVISION 6 1

Originator 9Q V Date 3"2 0 ' f 3 >

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t/ / Date I-26-73 :

General Manager, / / f~id KA. ROF [

Site Chemistry -'. _( '

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3.2 Requirements

Secondary System Uguld Waste Discharges To Onsite Evaporstion Ponds .

Concentration i The concentration of radioactive material discharged from secondary system liquid waste to the onsite ;

evaporation ponds shall be limited to the lower limit of detectability (LLD) defined as 5.0E-07 yCi/mi i l l for the principal gamma emitters or 1.0E-06 yCi/ml for I-131. *

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Applicability: At all times. l-a Actiom :

When any semndary system liquid waste discharge pathway concentration determined in accordance with the surveillance requirements given below exceeds the specified LLD, divert that discharge i pathway to the liquid radwaste system without delay.

i 3.2.1 Surveillance Requirements: '

a Radioactive liquid wastes collected in the chemical waste neutralizer tank shall be sampled and analyzed prior to their batchwise discharge to the onsite evaporation pond in j accordance with the sampling and analysis program specified in Table 3-5.

, b. With the concentration of radioactive material in the chemical waste neutralizer tank exceeding the specified LLD, sample and analyze other secondary system discharge pathways in accordance with the sampling and analysis program specified in Table 3-5. ,

3.2.2 Implementation of the Requirements:

$ This requirement is implemented by station manual procedures. ;

For the duration required to recover from the Unit 2 SGTR occurring on March 14,1993 the following l* limits for principle gamma emitters and tritium apply to Unit 2(these limits are temporary and will be !

removed from the ODCM when recovery is deemed mmplete by the Site Chemitry General Manager):

ISOTOPE Concentration limit Cumulative activity limit for l discharges to the evaporation g (uCi/cc)

i pond (Ci) ;

j Gamma emmitter with halflife 5 3.0E-06 No limit less than or equal to 2 years

, (excluding I-131) l Co-60 5 3.0E-05 5 2.0E-01 l Cs-134 s 9.0E-06 5 6.0E-02 l Cs-137 5 2.0E-05 s 1.5E-01 l I-131 1 1.0E-06 No limit l Other gamma emitter 1 5.0E-07 No limit l H-3 s 5.0E-03 5 3.8E+01 29 ODCM Rev. 6 j l

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Evaluation Log No.: $3 O o o u (,r

, 10CFR50.59 l

l SCREENING AND EVALUATION Page 1 of

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ACTrom uhDLM EE vie w: AEvisrow pcN DV3 h3 0 b b O D0M b! e J 4 IW be d V4 OE50mPTION CF PROPO$ED CHANGE Seut s E & In -h> 4ke CbCM wijn eme &

i r~ eld Ern 4e cc ce& G G h 5 Nt elwm -I A ev3 otrS a j nnA Ms rWJ is versisru -h Acd i i a em <tr, Erm % TtL o m4 L %% 2 ~ 3-l4A3. "TL.i eka o wkcech a chw -k wr>cokrcr a d Le s c ri bo tf sk -tt c. u fSAsl a Sec t1%s Q2C ZI.3.1.' -

10CFR50.59 SCREEN (Provide References on Response Justification Page) NO YEs Does the proposed change:

1. Make changes in ibe facility as it is described in the UFSAR? /

2. Make changes in procedures as they are described in the UFSAR?

3. Involve test or experiments not described in the UFSAR? ._. -

4 Require a change to the technical specifications? I -

'- Any answer to questions 1 through 3 "YES." then a 10CFR50.59 evatuation is required. Contact Document Control at ext. 82-6633 to obtain a tracking log number and enter the number in the Evaluation Log number block above. UFSAR Change Request per procedure 93AC-OLC01 may also be required.

Answer 4 is "YES,"then Technical Specification Change Request per procedure 93AC-OLC01 and NRC approval is required prior to implementation.

All answers 1 through 4 are 'NO." no 10CFR50.59 Evaluation required or Technical Specification change required recommend action approval.

10CFR50.59 EVALUATION (Provide Response Justihcation with References)

5. May the probability of an accident previously evaluated in the UFSAR be increased? !

6. May the censequences of an accident previously evaluated in the UFSAR be increased? !

7. May the probacility of a malfunction of equipment important to safety be increased? I S. May the consequences of a malfunction of equipment important to safety be increased?

9. May the pessibikty of an accident of a cifferent type than any previously evaluated in the UFSAR be created? /

10. May the possibility of a different type of maifunction than any previously evaluated in the UFSAR be created? I -

11. Is the margin of safety as defined in the basis for any technical spec:fication reduced?

Any a 7swer to questions 5 through 11"YES,"then an unreviewed safety question is identified. Proceed to procedure 93AC-OLC03 prior to implementation.

All answers 5 through 11 are "NO." there is no unreviewed safety question and action approval is recommended.

If UFSAR Chapter S/ Chapter 15 is potentially affected. forward a copy of evaluation to Nuclear Fuels Management.

I vertfy tha the above screeninglevaluation is acequate and accurate an the undersigned has received required training.

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10CFR50.59 REVIEW AND EVALUATION l RESPONSE JUSTIFICATION Page 2 of 6 )

l ACUcN UNDER REVIEW:(NAME/HTLE) REVISION OFFSITE DOSE CALCULATION 31ANUAL 06.00 i

PROCEDURE.PCP/rEMPORARY MoDIFICADON No-51ANUAL CUESBoN RESPONSE JuSTIDCATICN As a result of a Unit 2 Steam Generator Tube Rupture a large volume of slightly contaminated secondary

side water has been created. A!! hough the water has been processed by the condensate polisher !

. demineralizers, relief to ODCM requirement 3.2 is sti!! necessary to remove the excess water in the Unit !

] 2 secondary systems.

Contamination of a retention basin was considered in the FSAR. Only the method of how the activity is I removed from the retention basin has changed. Increasing the ODCM release limits for retention basin discharges to the evaporation pond has been previously evaluated as part of an emergency Technical Specification change for Unit 1 in 1987 (references e through h). This presents the basis for the proposed FSAR change.

As part of reference g, the basis for limiting doses to LLD levels was presented to the NRC. This basis i considered routine operational discharges to the evaporation ponds at LLD levels for the entire operating ;

cycle of a!! three units (40 years). This calculation presented the anticipated offsite doses resulting from 4

evaporation pond dry up and subsequent resuspension of radioactive material in pond sediment three years after completing the operating cycle. The analysis showed that after 43 years, only those isotopes l with long ha f lives would be present in significant quantities to contribute to doses to the member of the l'

general public. Isotopes that resu!!ed in total pond sediment inventories of less than one microcurie were not included as part of the analysis.

The ODCM concentration limits were based on the assumption that 2E6 gallons of secondary side water would need to be discharged to the evaporation ponds as part of the Unit 2 SGTR recovery. The +

4 concentrations of gamma emitters (excluding 1-131) that were determined to be limiting in the calculation i discussed above, were restricted to the MPC values in 10CFR20 Appendix B, Table 11, Column 2. Since !

the release is considered a unique event and will not continue throughout the life of the plant, the total activity that would be discharged to the waste ponds was calculated and estab!ished as a limit for ;

cumulative activity. Furthermore, since there are more than 33 years remaining before the release :

scenario analyzed above would happen, gamma emitters for isotopes with half lives fess than 2 years l were assumed to be discharged at a concentration limit of 3E-6 Ci/cc (corresponding to the unrestricted i liquid release MPC for an unknown non-alpha emitting radionuclide with a ha:f life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). l Due to the decay time, no limit was placed on radioactivity from these isotopes. The concentration limits {

currently in the procedure for other gamma emitting isotopes remains the same: SE-7 uCi/cc for principle gamma emitters with half lives greater than two years, and 1E-6 pCi/cc for I-131. Since these limits are within the original basis, no limits on cumulative activity are necessary. Using these assumptions, the total 7

activity that would be added to the pond sediments after 30 years was calculated and is summarized ;

below: '

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4 10CFR50.59 REVIEW AND EVALUATION l RESPONSE JUSTIFICATION Page 3 of 6 ACTION UNDER REVIEW:(NAME/ITTI.E) REVISION I OFFSITE DOSE CALCULtTION MANUAL 06.00 PROCEDURE.T*CP/IT.MPORARY MODIFICATION NO-MANUAL CUESTION RESPONSE JUSTTDCATioN f

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!soToPE CONC. UMIT HALF UFE SEDIMENT sEo: MENT FINAL % INCREASE l

( CVec) (years) INVENTORY INVENTORY sEoiMENT IN SEDIMENT FRoM uNrr 2 FRoM INVENTORY INVENTORY RELEASE continuous (CQ ,

ANER 30 OPERATION YEARS (Cl) AT LLo ' j osCHARGE ,

UMITS FoR 40 t YEARS (C1) !

1 Co-60 3E-5 5.27 4.39E-03 2.56E-01 2.60E 01 2%

i Cs-134 SE-6 2.06 2 B2E-06 S.02E 02 5.02E-02 <1% t Cs-137 2 E-5 30.17 7.60E 02 1.25 E+00 1.33E + 00 6% !

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samma 3E-6 2 < 1.00E-06 N/A < 1.00E 06 N/A emitters wrth l half life S 2 years t

l As can be seen from the above table, the percent of activity resulting from this release is not greater than *

! 10 % for any isotope identified in the original calculation. Therefore, since dose is a function of total activity, the increase in dose would be less than 10% Based on the results contained in the calculation i from reference g, the total dose consequences to the members of the general public remain negligible j and the conclusions drawn are still valid. Specifically, the annual total body dose due to ground !

contamination of an unrestricted area, arising from transportation and deposition by wind of the {

accumulated activity discharged to the pond from the secondary system of the plant (if the pond gets ;

dried up) on the unrestricted area is within the guidelines of 10 CFR Part 20.

l The proposed change to the ODCM also address a limit to the tritium concentration discharged to the evaporation pond. The concentration was procedurally restricted to the MPC values from 10 CFR 20 ;

Appendix B Table II, Column 2. The basis for this concentration originated from Technical Specification !

Interpretation 3.11.1.1-13-02-00 and is discussed further in reference d. The evaluation discussed in i j reference d for dilution of the Condensate Waste Neutralizer Tank tritium levels for discharge was based I I on releasing 750,000 ga!!ons of tritium at a concentration of SE-3 pCi/cc to the evaporation ponds. Tne l original calculation assumed that the activity would be diluted prior to discharge to the retention basin.

l l Although the total volume being discharged to the evaporation pond has increased, the calculational j methodology used to examine the release for offs:te dose due to tntium evaporation is the same. Using '

the same assumptions for evaporation pond volume,4.63E12 cc, and a total secondary water volume of i 2ES gallons, the evaporation pond activity will be 8.17E-06 uCi/cc. Using the dose factor for evaporation l pond releases from reference i of 5.7E4 mrem /yr per uCi/cc. The resultant dose from the release would !

be 0.47 mrem /yr, well below 10CFR50 Appendix ! and 10 CFR 20 limits. The final concentration in the '

evaporation pond would also be less than 1E-5 uCi/ce, the alert level associated with the ground water monitoring permit. Therefore, there is no significant impact on offsite doses.

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!I a i 10CFR50.59 REVIEW AND EVALUATION i

RESPONSE JUSTIFICATION P3pe 4 of 6 ACDoN UNDER REVIEW:(NAME/HTI.E) REVISION '

1 OFFSITE DOSE CALCULATION hiANUAL 06.00 PROCEDURE.'PCPfrEMPORARY MODIDCAT1oN NO:

SIANUAL i

CUESDON RESPONSE JUsTIDCADON i

Finally, the effect of discharging at the elevated release concentrations to the evaporation ponds resulting from the groundwater ingestion pathway were considered. Using a total discharge t clume of 2E6 gallons and the proposed concentration limits, the dose due to offsite releases was calculated using the guidance in reference i. A 100 year transport time from the evaporation pond to the nearest offsite drinking water +

well was assumed for all isotopes. These doses are summarized below.

?

isotope Dose (mrem /yr) &

CRmCAL ORGAN n

CO-60 2.BE-6 ill ADULT GI-LU Cs-134 8.4 E-15 INFANT UVER Cs-137 6,6E 1 INFANT UVER H-3 3.0E-3 CHILD UVER ,

i From the data in the table above, the offsite dose from the ground water pathway is negligible. The conclusions obtained in the safety evaluation contained for reference i remain valid.

i in conclusion, the incremental increase in exposure is negligible and well within 10 CFR part 20; 40 CFR Part 190; 10 CFR Part 50, Appendix I; and 10 CFR Part 100 limits.

1,3,4 No. This change describes a change to the procedure described in FSAR Section 9.3.3.2.1.3 for dealing with discharges from the retention basin to the evaporation ponds that has an activity level greater than that allowed per ODCM Section 3.2. There are no Technical specification requirements, nor is this a test or experiment. There are no physical changes to the plant. 1 2 Yes. FSAR Sections 9.3.3.2.1.3.1 and 9.3.3.3.8 describes the procedure for dealing with secondary side water when racioactivity levels in excess of TS 3/4.11.1 LLD limits (currently referred to as ODCM 3.2 LLD limits) are detected in the Turbine building sumps, retention basin, Chemical Waste Neutralizer Tank, and the Oily waste /Non radioactive waste systems. The FSAR states that in the event that activity is detected the Oily waste /Non radioactive waste systems, the waste will be directed to the LRS for processing. If radioactive waste above ODCM LLD limits is detected in the retention basin, a portable ion exchanger will be used to transfer the water from the dirty basin to the clean basin. The clean water will then be released. !

The proposed change to the ODCM will now allow the waste to be discharged above CDCM 3.2 LLD limits to the evaporation ponds via the retention basins. This change represents a change to the procedures as described in the FSAR.

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10CFR50.59 REVIEW AND EVALUATION l RESPONSE JUSTIFICATION Pa;;e 5 of 6 ACRON UNDER REVIEW:(NAMETHTLE) REVISION OFFSITE DOSE CALCULATION MANUAL 06.00 ,

1 PROCEDURE?CP/IEMPORARY MODIRCATION NO:

MANUAL QUESTION RESPONSE JUSTIFICA' DON I

i 5 Discharging radioactivity to the evaporation ponds in excess of ODCM 3.2 LLD limits results in ,

activity levels in excess of the CDCM LLD limits. This situation has been previously addressed in reference c. From this reference, FSAR Chapter 15.7.2 describes the analysis for liquid radwaste release to the environment. Contamination of the retention basins in not the limiting accident scenario, and it was considered explicitly in FSAR Chapter 9.3.3.2.1.3.1. This change does not change the probability of an inadvertent release of liquid radwaste.

6. As discussed in reference c, the limiting accident analyzed in the FSAR is the rupture of the RVH ,

in section 2.4.13.3. The proposed concentration limits contained in this ODCM change are several orders of magnitude less than what was used in the RWT Rupture analysis. Although the volume of both retention basins (approximately 1,000,000 gallons) is larger than the RVH (approximately 700,000 gallons), the total activity released by a RWT rupture will be several orders of magnitude greater than a retention basin release due to the difference in assumed concentrations.

7,8 None of the equipment to be used in this evolution is considered 'important to safety *. The ;

retention basin is located outside of the protected area. <

9. Contamination of a retention basin was considered in the FSAR. Only the method of how the i activity is removed from the retention basin has changed. increasing the ODCM release limits for retention basin discharges to the evaporation pond has been previously evaluated as part of an ,

emergency Technical Specification change for Unit 1 in 1987 (references e through h). This presents the basis for the proposed FSAR change. The proposed limits are evaluated in the general discussion section of this evaluation. As concluded from this evaluation, the increased limits does not after the original conclusions discussed in references e through h.

10. The limiting failure mechanism for this change would be the inadvertent release of the entire contents of the retention basin to the evaporation ponds. Sinca the proposed change was based I on discharging a maximum of 7.5E9 cc of secondary volume xd the totalvolume of the retention basins is 1E6 cc, the evaluation is bounded by this event. The postulated leakage from the retention basin to the ground water has been already discussed in the FSAR and is bounded by the RWT rupture scenario. There are no other credible failure mechanisms.

11. There are no Technical Specification Limits governing this equipment or operation. The limits are specified by the ODCM, and as described in the general discussion section, the incremental increase in exposure is negligible and well within 10 CFR part 20; 40 CFR Part 190; 10 CFR Pan 50, Appendix 1; and 10 CFR Part 100 limits.

REFERENCES:

a). Operating License: Unit 1, Amendment 69; Unit 2 Amendment 55; Unit 3, Amendment 43 Sections and BASES: 1.18, 6.8.1, 6.8.4.g, 6.8.4.h, 6.9.1.8, and 6.14 gg

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i 10CFR50.59 REVIEW AND EVALUATION RESPONSE JUSTIFICATION Page 6 of 6 ACDON UNDER REVIEW:(NAME/ TITLE) REVISION l OFFSITE DOSE CALCULATION MANUAL 06.00 PROCEDURE /PCP/EMPORARY MODIFICADON NO:

MANUAL !

CUESDON RESPONSE JUSDFICADON i

b). UFSAR, Revision 4:

f Sections: [

t c). 10CFR 50.59 Safety Evaluation 93-00041 concerning Intentional Dilution to Release of i Retention Basins.

i d). Letter 214-00667-JAS, from J.A. Scott to R.K. Flood, dated March 16,1993, concerning

' Disposition of Unit 2 Secondary Water *.

+

e). Letter 161-00204-JGH/DAL, from J.G. Haynes to USNRC, dated March 10, 1987, !

concerning

Proposed Emergency Technical Specification Change - Secondary System i i

Liquid Waste Discharges to Onsite Evaporation Pond'. I f). Letter 161-00212-JGH/DAL, from J.G. Haynes to USNRC, dated May 14,1987, conceming

' Response to NRC Information Request Regarding Proposed Emergency Technical ,

l Specification Change - Secondary System Uquid Waste Discharges to Onsite Evaporation

! Pond'. l g). Letter 161-00413-JGH/JBK, from J.G. Haynes to USNRC, dated July 31,1987, concerning

'NRC Request for Additional Information Regarding Onsite Evaporation Ponds for Palo ;

Verde'. I h). Letter from USNRC to E.E. Van Brunt, dated June 6,1987, concerning ' Issuance of j Admendment No.18 to Facility Operating License No. - NPF-41, for Palo Verde Unit 1 (TAC No. 65290)*.

i). EER 89-OW-004, Evaporation Pond #1 h?'

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Page 6 cf 7 REVIEW AND CONTROL CF THE 74AC-9CY12 01.00 CFFSITE DOSE CALCUuTiCN MANUAL -

Appendix A Page 1 of 1 SAMPLE

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i PEV!SION REQUEST FORM l DATE: 3-20-93 OR!GINATCR: MAR 1!EV LE5AtJ EXT: 649O '

PAGE I CF_/

Description and Just;fication of Revision:

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i fi Approved By: [M" - 7 ate: 3- 24 D EMS /Eificents Superv:scr (Site Chemistry)

Use additional pages as required.

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WORKING COPY Page 7 cf 7 i RE'EN AND CONTROL OF THE 74AC-9CY12 01.00 CFFSITE DOSE CALCUl.ATION MANUAL .

Appendix B Page 1 of 1 I l

SAMPLE TECHNICAL SPEC;FICATION REFERENCE A. Periodic Review and/or Revision Requiremen:s:

Technical Specification, Section 6.8.4.g and Section 6.8.4.h have been reviewed. The program elements required to be contained in the ODCM are present in this review / revision of the CDCM.

CDCM Revision No. b initiator Name (p inted) MAft/EY_LESAI/

Signatur M ') ate E-20-93 Technical Re e er f 6 GM-c - Date J-20-f3 i B. Acditional Revision Requirements:

This CDCM revision submi::al contains:

1.

Sufficient information to support the change together with the appropria:e anatyses or '

evaluations justifying the change (s) (RCTS 011072-01) and; 2.

A determination that the change wi1 maintain the level of radioactive effluent control required

by 10 CFR 20.106,40 CFR Part 190,10 CFR 50.3Sa, and Appendix 1 to 10 CFR Part 50 and not adversely impact the accuracy er reliability of effluent, dose, or setpoint calculations.

(RCTS 011050-01). t t

3.

Eacn change shall be identified by markings in the margin of the affected pages, clearly incicating the area if the page that was enanged, and shall indicate the date(e.g., month / year) '

the change was impleme ted.

Ini:ia: A. Date 5-20-f]

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Technical Reviewe -_MW Date 2-20-73 i

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l l OFFSITE DOSE CALCULATION MAhTAL

! PALO VERDE NUCLEAR GENERATINO STATION UNITS 1,2 AhT 3 i

j REVISION 7 l

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! Tech. Reviewer [2/ W/M ZW N 4 Date 5,M- 93 5 7 "V General Manager, Site Chemistry dg 74 L.rr Date f-zG D

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i Table of Contents Title Page 1.0 Intnxtuction 1 t 1.1 Liquid Effluent Pathways i 1.2 Gaseous Effluent Pathways 2 13 Nuisance Pathwap 2 1.4 Meteorology 4 i

2.0 Gaseous Emtsent Monitor Setpoints 5 ,

2.1 Requirements

Gaseous Monitors 5 2.1.1 Surveillanx Requirements 5 2.1.2 Implementation of the Requirements 12 2.1.2.1 Equivalent Dose Factor Determination 13 2.1.2.2 Site Release Rate limit 14 2.1.23 Unit Release Rate Limits 15 2.1.2.4 Setpoint Determination 16 2.1.2.5 Monitor Calibration 17 .

3.0 Gaseous and Liquid Emuer.t - Dose Rate 18

3.1 Requirements

Gaseous Efiluents 18 3.1.1 Surveillance Requirements 18 3.1.2 Implementation of the Requirements 19

3.2 Requirements

Secondary System Liquid Waste Discharges to Onsite Evaporation Ponds - Concentration 27 3.2.1 Surveillance Requirements 27 3.2.2 Irnplementation of the Requirements 27 !

\

4.0 Gaseous and Liquid Emuents - Dose 30 >

4.1 Requirements

Noble Gases 30 '

4.1.1 Surveillance Requirements 30 i 4.1.2 Implementation of the Requirements 31

4.2 Requirements

Iodine - 131, Iodine-133, Tritium, and All Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 32 4.2.1 Surveillance Requirement 32 '

4.2.2 Implementation of the Requirements 33 43 Requirements: Gascous Radwaste Treatment 35 43.1 Surveillance Requirement 35 43.2 Implementation of the Requiremer: 35

4.4 Requirements

Liquid Emuents 56 4.4.1 Surveillance Requirement 56 4.4.2 Implementation of the Requirements 56 i I

i ODCM Rev. 7 l

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Table of Contents ( Continued )

Title Page 5.0 Total Dose and Dose to Public Onsite 57

5.1 Requirements

Total Dose 57 l 5.1.1 Surveillance Requirement 57 5.1.2 Implementation of the Requirement 57 6.0 Radiological Environmental Monitoring Pmgram - REMP 61

6.1 Requirement

REMP 61 6.1.1 Surveillance Requirement 62 6.1.2 Implementation of the Requirement 62 ft2 Requirement: Land Use Census 70 6.2.1 Surveillance Requirement 70 6.2.2 Implementation of the Requirement 70 63 Requirement: Interlaboratory Comparison Program 71 63.1 Surveillance Requirement 71 63.2 Implementation of the Requirement 71 7.0 Radioksical Reports 82

7.1 Requirement

Semiannual Radioactive Effluent Release Report 82

7.2 Requirement

Annual Radiological Environmental Operating Report 84 APPENDIX A DETERMINATION OF CONTROLLING LOCATION 85 APPENDIX B BASES FOR REQUIREMENTS 86 I

l 2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING I INSTRUMENTATION 86 3.1 GASEOUS EFFLUENT - DOSE RATE 86 ,

3.2 SECONDARY SYSTEM LIQUID WASTE DISCHARGE TO ONSITE EVAPORATION PONDS - CONCENTRATION 87 4.1 GASEOUS EFFLUENT - DOSE, Noble Gases 87 4.2 GASEOUS EFFLUENT - DOSE - hxline - 131, lodine-133, Tritium, ar.d All Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 88 43 GASEOUS RADWASTE TREATMENT 88 4.4 SECONDARY SYSTEM LIQUID WASTE DISCHARGE l TO ONSITE EVAPORATION PONDS - DOSE 89 l 5.1 TOTAL DOSE AND DOSE TO PUBLIC ONSITE 89 6.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 90 6.2 LAND USE CENSUS 90 1 63 INTERLABORATORY COMPARISON PROGRAM 90 APPENDIX C DEFINITIONS 91 APPENDIX D Disposition of NRC Generic letter 89-01 Items from PVNGS Technical Specifications to the ODCM %

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ii ODCM Rev. ,

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l List of Tables TABLE TITLE PAGE ,

4 1-1 NUISANCE PATHWAYS 3 i 2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING

! INSTRUMENTATION 6 s 2-2 RADIOACTIVE GASEOUS EITLUENT MONITORING .

I INSTRUMENTATION SURVEILLANCE REQUIREMEN"13 10 3-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 21 3-2 DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE SITE BOUNDARY 24 3-3 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS 25 3-4 P VALUES 26 3-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 28 4-1 R VALUES FOR THE GROUND PATHWAY 38 4-2 R VALUES FOR THE ADULT VEGETATION PATHWAY 39 l 4-3 R VALUES FOR THE TEEN VEGETATION PATHWAY 40 4-4 R VALUES FOR THE CHILD VEGETATION PATHWAY 41 4-5 R VALUES FOR THE ADULT MEAT PATHWAY 42 4-6 R VALUES FOR THE TEEN MEAT PATHWAY 43 4-7 R VALUES FOR THE CHILD MEAT PATHWAY 44 -

4-8 R VALUES FOR THE ADULT COW MILK PATHWAY 45 ;

i 4-9 R VALUES FOR THE TEEN COW MILK PATHWAY 46 ;

4-10 R VALUES FOR THE CHILD COW MILK PATHWAY 47 i 4-11 R VALUES FOR THE INFANT COW MILK PATHWAY 48 l

4-12 R VALUES FOR THE ADULT INHALATION PATHWAY 49 !

iii ODCM Rev. 7

F b

List of Tables !

l TABLE TITLE PAGE .

4-13 R VALUES FOR THE TEEN INHALATION PATHWAY 50 4-14 R VALUES FOR THE CHILD INHALATION PATHWAY 51 l 4-15 R VALUES FOR THE INFANT INHALATION PATHWAY 52

4-16 DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST RESIDENT LOCATIONS 53 l

6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 63 I

! t l 6-2 Reporting Ixvels for Radioactivity Concentrations in Environmental Samples 67 6-3 Detection Capabilitics for Environmental Analysis 68 !

6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS 72 C-1 FREQUENCY NOTATION 95

C-2 OPERATIONAL MODES 95 i

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6-1 Radiological Emironmental Monitoring Program !

Sample Sites, O to 10 miles 76 !

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l 6-2 Radiological Emironmental Monitoring Program Sampic Sites, O to 35 miles 77 1

6-3 Radiological Emironmental Monitoring Program Sample Sites,35 to 75 miles 78 ,

6-4 Site Exclusion Area Boundary 79

! 6-5 Gaseous Effluent Release Points 80 ;

i 6-6 Low Population Zone 81 l

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1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) implements the program elements which are required by the Administrative Controls section of the Technical Specifications, Section 6.8.4.g. Radioac'ive Effluent Controls Program, and Section 6.8.4.h. Radiological Environmental Monitoring Ptogram at the Palo Verde Nuclear Generating Station (PVNGS) for Unit 1 Unit 2 and Unit 3. The ODCM is defined in Technical Specifications, Section 1.18 and in the Definitions in Appendix C of this manual. The ODCM contains the operational requirements, the surveillance requirements, and actions required if the operational requirements are not met for the Radioactive Effluent Controls Program and the Radiological Emironmental Monitoring Program to assure compliance with 10 CFR 20.106,40 CFR Part 190,10 CFR j 50.36a, and Appendix 1 to 10 CFR Part 50. The Technical Specifications, section 3/4.0, also app!y to the g

ODCM. Substitute the word

Requirements

for

Limiting Condition for Operation". It should be noted

, that the hot and cold shutdown and operability requirements in Technical Specification 3.0.3 and 4.0.3 do not apply to any of the requirements contained in this ODCM. The ODCM also contains descriptions of the information that should be included in the Annual Radiological Emironmental Operating Report ;

and the Semiannual Radioactive Efuuent Release Report required by Technical Specifications Section 6.9.1.7 and 6.9.1.8.

The ODCM provides the parameters and methodology to be used in calculating offsite doses resulting from radioactive effluents, in the calculation of gaseous effluent monitor Alarm / Trip Setpoints, and in the conduct of the Radiological Emironmental Monitoring Program. Included are methods for determining air, whole body, and organ dose at the controlling location due to plant effluents to assure compliance with the regulatory requirements detailed in the ODCM. Methods are included for i performing dose projections to assure mmpliance with the gaseous treatment system operability sections of the ODCM. He ODCM utilizes information from NRC Regulatory Guide 1.109, ' Calculation of Annual Doses to Man from Routine Releases of Reactor EfDuents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,* October 1977, and NRC NUREG 0133,

Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978. NUREG 0133 utilizes some of the key information in Regulatory Guide 1.109 to provide methods which were used in the preparation of the radiological effluent Technical Specifications and which have now been transferred to the ODCM in accordance with NRC Generic Letter 89-01,' Implementation of Programmatic Controls for Radiological Effluent Techrucal Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program," January 31, 1989, and NUREG 1301,'Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors", Generic letter 89-01, Supplement No.1 April 1991. I 1.1 Liouid Efnuent Pathways Dose calculation methodology for liquid effluents is not included in this manual due to the desert location of the plant, the hydrology of the area, and the fact that there are no liquid releases to areas at or beyond the SITE BOUNDARY during normal operation. All liquid discharges to the onsite evaporation ponds are controlled by Section 3.2. The impact of postulated accidental scepages on the groundwater system, and in particular on the existing wells located in the 5-mile zone around the site area has been calculated and analyzed in Section 2.4.13.3 of the PVNGS FSAR.

If plant operating conditions become such that the likelihood of a liquid effluent pathway is created, then dose calculation methodology for this pathway will be added to this manual.

1 ODCM Rev. 7

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j 1.2 Gaseous Effluent Pathways l

All gaseous effluents are treated as ground level reicases and are considered to be 'long-term

as discussed in NUREG-0133,

Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants". This includes the containment purge and Waste Gas Decay Tank releases as well as the normal ventilation system and condenser vacuum exhaust releases. All releases are either greater than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in duration or are made at random, not depending upon atmospheric conditions or ,

time of day. The releases are lumped together and calculated as an entity. Historical annual average X/Q values are used throughout this manual for all gaseous effluent setpoint and dose calculations.

Airborne releases are further subdivided into two subclasses:

1.2.1 Todine - 131. lodine - 133. Tritium and Radionuclides in Particulate Form with Half-lives Greater than Eicht Days In this model, a controlling location is identified for assessing the maximum exposure to a MEMBER OF THE PUBLIC for the various pathways and to critical organs. Infant exposure occurs through inhalation and any actual milk pathway. Child, teenager and adult exposure l derives from inhalation, consumed vegetation pathways, and any actual milk and meat l pathways. Dose to each of the seven organs listed in Regulatory Guide 1.109 (bone, liver, total body, thyroid, kidney, lung and GI-LLI) are computed from individual nuclide contributions in each sector. The largest of the organ doses in any sector is compared to 10 CFR 50, Appendix 1 design objectives. The release rates of these nuclides will be converted to instantaneous dose rates for comparison to the limits of 10 CFR 20.

1.2.2 Noble Gases The air dose from both the beta and gamma radiation component of the noble gases will be I

assessed and compared to the 10 CFR 50, Appendix 1 design objectives. The noble gas release i rate will be converted to instantaneous dose rates for comparison to the limits of 10 CFR 20.

Section 2.0 of this manual discusses the methodology to be used in determining effluent i monitor alarm / trip setpoints to assure compliance with the 10 CFR Part 20 limits as implemented in Section 3.0. Section 4.0 discusses the methods to assure releases are As 1.aw As Reasonably Achievable (ALARA) in accordance with Appendix 1 to 10 CFR Part 50.

Methods are described in Section 5.0 for determining the annual cumulative dose to a MEMBER OFTHE PUBLIC from gaseous effluents and direct radiation to assure compliance with 40 CFR Part 190.

The requirements for the Semiannual Radiological Effluent Release Report and the Radiological Environmental Monitoring Program, including the Annualland Use Census and the Interlaboratory Comparison Program, and the Annual Environmental Report are described in Sections 6.0 and 7.0 of this manual.

1.3 Neisance Pathwws This section addresses the potential release pathways which should not contribute more than 10% of the doses evaluated in this manual. Table 1-1 lists examples of potential release pathways. The ODCM methodology for calculation of doses will be applied to an applicable release pathway if a likely potential arises for contributing more than 10% of the doses evaluated in this manual.

2 ODCM Rev. 7

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i TABLE 1-1 l i

NUISANCE PATHWAYS i (EXAMPLES) {

i Evaporation Pond '

l Cooling Towers laundry /Demn Building Exhaust Unmonitored Semndary System Steam Vents / Reliefs i Turbine Building Ventilation Exhaust Unmonitored Tank Atmospheric Vents Dry Active Waste Processing and Storage (DAWPS) Building Respirator Cleaning Facility Semndary Side Decontamination Equipment i

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_ _ _ ,_ _, , _ . . . , . - . . ..m...-- -

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1.4 Metenrolocv ?

?

i Historical annual average atmospheric dispersion (X/Q) and deposition (D/Q) data, based on mne ;

years of meteorological data, and given in Table 3-2 for each of the three nuclear generating units are !

used to demonstrate compliance with the ODCM Requirements. These Requirements include:

i Section 2.0 Gaseous Effluent Monitor Setpoints:

Section 3.0 Gaseous and Liquid Efnuent - Dose Rate Section 4.0 Gaseous and Liquid Effluent - Dose Section 5.0 Total Dose and Dose to Public Onsite Sections 2.0 and 3.0 specify utilizing the highest X/Q or D/O meteorological dispersion parameter at l the Site Boundary for any of the three units as applicable. Using the highest dispersion parameter for any of the units provides a conservative assumption to assure compliance with the higher 10 CFR ;

l Part 20 limits. l l

l Section 4.0 specifies utilizing the highest X/O at the Site Boundary for the particular unit, from Table 3-2 for noble gases. The highest X/O and D/Q are utilized for the particular unit's releases as applicable for gases other than noble gases (iodines, particulates, and tritium) for the controlling i pathway's location (site boundary using Tabic 3-2 or other controlling locations using Table 416). ,

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Section 5.0 specifies utilizing the highest X/O for the particular unit's releases at the controlling location from Table 4-16 for noble gases. The highest X/Q and D/O are utilized for the particular

, unit's releases as applicable for gases other than noble gases at the controlling pathway's location l using Table 4-16.

i Section 7.0 requires that the meteorological conditions concurrent with the time of release of ,

radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. {

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1 i 2.0 GASEOUS EFFLUENT MONITOR SETPOINTS l

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2.1 Requirements

Gaseous Monitors The radioactive gascous effluent monitoring instrumentation channels shown in Tabic 2-1 shall be ,

OPERABLE with their alarm / trip setpoints set to ensure that the dose requirements in Section 3.0 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Section 2.1.2.

l Applicability: As showie in Tabic 2-1. This requirement does not apply to RU-141 or RU-142 if i DCP-13-PJ-SQM5 has been implemented. :

Action:

a. With the low range radioactive gaseous effluent monitoring instrumentation channel alarm / trip ;

setpoint less conservative than required by the above Requirement, immediately suspend the i release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-1. Restore the moperable instrumentation to OPERABLE status within 30 days or, if unsuccessful, explain in the next :

Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within 1 the time specified.

2.1.1 Surveillance Requirements:

a. Each radioactive gaseous effluent monitoringinstrumentation channel shall be demonstrated .

OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL i CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2. ,

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TA11I.E 2-1 RADIOACTIVE GASEOUS EFFI.UENT MONITOltING INSTRUMENTATION MINIMUM CIIANNEIS INSTRUMENT OPERAlli,E APPI,1CAllllIIT ACrlON

1. GASEOUS HADWASTE SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic Termination of Release #RU-12 1 # 35

b. How Rate Monitor 1 # 36

2. NOT USED

3. CONDENSER ETACUATION SYSTEM A. IA>w Range Monitors

a. Nobic Gas Activity Monitor #RU-141 1 1, 2, 3 " *, 4 * " 37

b. lodine S mpler 1 1, 2, 3 ' ", 4 * *

40

c. Particulate Sampler 1 1, 2, 3 " *, 4 " ' 40

d. Flow Rate Monitor 1 1, 2, 3 " *, 4 " ' 36

c. Sampler Row Rate Measuring Device 1 1, 2, 3 " *, 4 * " 36 II. Iligh Range Monitors

a. Noble Gas Activity Monitor #RU-142 1 1, 2, 3 * ", 4 "

42

b. lodine Sampler 1 1, 2, 3 " ', 4" ' 42

c. Particulate Samplcr 1 1, 2, 3" *, 4 * " 42

d. Sampler Flow Rate Measuring Device 1 1, 2, 3" *, 4"' 42

4. PIANT VENT SYSTEM A. I;>w Range Monitors

a. Nobic Gas Activity Monitor #RU-143 1 37

b. h> dine Sampler 1 40

c. Particulate Sampler 1 40

d. How Rate Monitor

1 36

c. Sampler How Rate Measuring Device 1 36 6 ODCM Rev. 7

TAllt.E 21 (Continucd)

RAI)IOACTIVE GASEOUS EFFI,UENT MONITORING INSTRUMENTATION MINIMUM CilANNEl,S INSTRUMENT OPERAllLE APPI.lCAllllIIT ACTION

4. PLANT VENT SYSTEM (Contimied)

II. liigh Range Monitors

a. Noble Gas Activity Momtor #RU-144 1 42

b. lodine Sampler

1 42

c. Particulate Sampler

1 42

d. Sampler Flow Rate Measuring Device 1 42

5. FUEL, llUll.l>ING VENTILATION SYSTEM =

A. Inw Range Monitors

a. Noble Gas Activity Monitor #RU-145 1 ## 37,41

b. Iodine Sampler 1 ## 40

c. Particulate Sampic 1 ## 40

d. Flow Rate Monitor 1 ## 36

c. Sampler Flow Rate Measuring Device 1 ## 36

11. Iligh Range Monitors

a. Nobic Gas Activity Monitor #RU-146 1 ## 41,42

b. Iodine Sampler 1 ## 42

c. Particulate Sampic 1 ## 42

d. Sampler Flow Rate Measuring Device 1 ## 42 7 ODCM Rev. 7

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Table 2-1 (Continued) .

l TABLE NOTATION At all times.

During GASEOUS RADWASTE SYSTEM operation Whenever the condenser air removal system is in operation, or whenever turbine glands are being i supplied with steam from sources other than the auxiliary boiler (s). '

During waste gas release.

1. In MODES 1,2,3, and 4 or when irradiated fuel is in the fuel storage pool.

ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment i provided that prior to initiating the release:

1 I 1

a. At least two independent samples of the tanks mntents are analyzed, and

b. At least two technically qualified members of the facility staff indeper.dently verify I the release rate calculations and discharge valve lineup:

Otherwise, suspend release of radioactive effluents via this pathway, i ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may mntinue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, cffluent releases via this pathway may continue provided the e actions of (a) or (b) or (c) are performed I

a. Initiate the Preplanned Alternate Sampling Program to monitor the appropriate parameter (s). [

b. Place moveable air monitors in line.

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c. Take grab samples at Icast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement,immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 39 - NOT USED 1

l 8 ODCM Rev. 7

Table 2-1 (Continued)

TABLE NOTATION l

ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases sia the effected pathway may continue prosided ;

samples are continuously collected with auxiliary sampling equipment as required in Table ;

3-1 within one hour after the channel has been declared inoperable.

ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, comply with the ACTION b of Technical Specification 3.9.12 or operate the fuel building essential ventilation system while moving irradiated fuel.

ACTION 42 - With the number of channels OPERABLE less than required by the Minimum Channels l OPERABLE requirement restore the channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or:

l

a. Initiate the Preplanned Alternate Sampling Program to monitor the appropriate l parameter (s) when it is needed.

b. Prepare and submit a Special Report to the Commission pursuant to Technical l Specification 6.9.2 within 30 days following the event outlining the action (s) taken, l the cause of the inoperability, and the plans and schedule for restoring the system i to OPERABLE status. j Note: l Action item numbering and instrument numbering are the same as in the Technical Specifications from which this section was taken to avoid potential confusion. Thus not all action item numbers will be found in this ODCM.

9 ODCM Rev. 7

i TAllLE 2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILIANCE REQUIREMENTS CIIANNEL MODE IN WIIICII i CllANNEL SOURCE CIIANNEL FUNCTIONAL SURVEll,IANCE !

INSTRUMENT CIIECK CIIECK CALIHRATION TEST IS REQUIRED

1. GASEOUS RADWASTE SYSTEM

a. Noble Gas Activity Monitor - !

< Providing Alarm and Automatic Termination of Release RU-12 P P(7) R(3) Q(1),(2).P# # # # !

, b. Flow Rate Monitor P N.A. R Q,P### #

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2. del.ETED

3. CONDENSER EVACUATION SYSTEM (RU-141 and RU 142) j

a. Noble Gas Activity Monitor D(5) M(7) R(3) Q(2) 1,2,3* ", 4 * "

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b. lodine Sampict N.A. N.A. N.A. N.A. 1,2,3 * ", 4 " * ,

c. Particulate Sampler N.A. N.A. N.A. N.A. 1,2,3 * ", 4 " '

d. Flow Rate Monitor D(6) N.A. R Q 1,2,3 " *, 4" *

c. Sampler Flow Rate Measuring Device D(6) N.A. R O 1,2,3 * ", 4 * " t

4. PI ANT VENT SYSTEM l (RU 143 and RU 144) j a. Noble Gas Activity Monitor D(5) M(7) R(3) . Q(2) 1 b. lodine Sampler N.A. N.A. N.A. N.A. *

Particulate Sampler N.A. N.A. N.A. *

} c. N.A.

2

d. Flow Rate Monitor D(6) N.A. R Q *

c. Sampler Flow Rate Measuring Device D(6) N.A. R Q *

! 5. FUEL llUILDING VENTIIATION SYSTEM l 3 (RU-145 and RU 146) !

' a. Noble Gas Activity Monitor ##

D(5) M(7) R(3) Q(2) l

b. Iodine Sampler N.A. N.A. N.A. N.A. ##

3

c. Particulate Sampler N.A. N.A. N.A. N.A. ##

! d. Flow Rate Monitor D(6) N.A. R Q ## !

c. Sampler Flow Rate Measuring Device D(6) N.A. R Q ##

. 10 ODCM Rev. 7 i

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-. - -. . _ ._-..w.

. ...,.mwe,.,...,,om-...=-, ...-.,,e ...,,1 .,-, ,._. .w...,w,,,-,.-e-. .w .,.,, . .. ..y_3 , m_, ,. - ,m.

Table 2-2 (Continued) ,

i TAflLE NOTATION At all times.

During GASEOUS RADWASTE SYSTEM operation Whenever the condeser air removal system is in operation, or whenever turbine glands are being l supplied with steam from sources other than the auxiliary boiler (s).

During waste gas release.

1. In MODES 1,2. 3, and 4 or when irradiated fuel is in the fuel storage pool. ,

1. 1. Functional test should consist of, but act be limited to, a veri'ication of system isolation capability !

by the insertion of a simulated alarm condition.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if the instrument indicates racasured levels above the alarm / trip setpoint.

(2) The CHANNEL FUNCTIONAL TESTshall also demonstrate that control room alarm annunciation ;

occurs if any of the following mnditions exists:

1. Instrument indicates measured levels above nie alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and ,

measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial caliDration shW be used. ,

(4) NOT USED (5) The channel check for channels in standby status shall consist of verification that the channel is l on.line and reachable . '

(6) Daily channel check not required for flow monitors in standby status.

(7) LLD may be utilized as the check source in lieu of a source of increased activity.

Note: Action item numbering and instrument numbering are the same as in the Technical Specifications from which this section was taken to avoid potential confusion. Thus not all action item numbers will be found in this ODCM.

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i 2.1.2 Implementation of the Requirements:

he general methodology for establishing low range gaseous effluent monitor setpoints is based upon a site release rate limit in Ci/sec derived from site specific meteorological dispersion conditions, radioisotopic distribution, and whole body and skin dose factors. He high alarm of the low range monitors will alarm / trip when the release rate from an indhidual vent will result in exceeding the limits in Section 3.1. 80% of Section 3.1 limits is considered to be the site release rate limit. He site release rate limit will be allocated among the i

licensed units' release points. He unit release rate limit will then be utilizal for the determination of gaseous effluent monitor setpoints. A fraction of the unit release rate limit '

is then allotted to each release point and its monitor alert setpoint (yCi/cc) is derived using actual or fan design flow rates.

Administrative values are used to reduce each setpoint to account for the potential acti ity la

' other releases. Rcse administrative values shall be reviewed based on actual release data. He RU-141 alert alarm setpoint rnay be further reduced to provide early indication of steam

! , generator tube leakage, l

For the purpose of implementation of Section 2.1, the alarm setpoint levels for low range :

effluent noble gas monitors are established to ensure that personnel are alerted when the noble gas releases are at a r:te such that if the releases would continue for the year they would [

approach the total body dose rate of 500 mrem /yr and 3000 mrem 6T skin dose in Section 3.1.

The equations in Section 3.1 of this manual provide the methodology for calculating the gaseous effluent dose rate.

He evaluation of doses due to releases of radioactive material can be simplified by the use of equivalent dose factors as defined in Section 2.2.1.

l l De equivalent dose factors will be evaluated periodically to assure that the best information on isotopic dictribution is being used for the dose equivalent value.

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l 2.1.2.1 Ecuivalent Dose Factor Determination ,

The equivalent whole body dose factor is calculated as follows:

f K, = L [(K,)(f,)] (2-1)

Whete: '

s K, = the equivalent whole body dose factor weighted by historical j radionuclide distribution in releases in mrem /yr per yC /m'.

K, = the whole body dose factor due to gamma emissions for each identified noble gas radionuclide i, in mrem /yr per pCi/m' from Table 3-3. ,

r f, = the fraction of noble gas radionuclide i in the total noble gas ,

radionuclide mix. :

The equivalent skin dose factor is calculated as follows:

(L+1.1M), = E[(14 + 1.1M,)(fj] (2-2)

, Where:

l (L+1.1M), = the equivalent skin dose factor due to beta and gamma emissions from all noble gases released, weighted by the historical radionuclide distribution in releases in mrembT per yCi/m'.

t I4 = the skin dose factor due to the beta emissions for each identified l noble gas radionuclide i, in mrem /yr per pCi/m' from Table 3-3.

M, = the air dose factor due to gamma emissions for each identified '

l noble gas radionuclide i, in mradlyr per uCi/m' from Table 3-3.

f, = the fraction of nobic gas radionuclide i in the total noble gas radionuclide mix.

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

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2.1.2.2 Site Release Rate Limit (Qsrm)

The release rates cx>rresponding to 80% of the whole body (Qwa) and skin (Osg) dose rate limits are calculated using the equivalent dose factors defined in Section 2.1.2.1.

The site release rate limit (Ostm) is the lower of 0.3 c,r Osg, thus assuring that the ;

more restrictive dose rate limit will not be exceeded. !

The Osrm is established as follows: f (Dws) (0.8)

Omw, =

(2-3)

(() (X/Q)stm Where: ,

O m wa = the site release rate,in pCi/sec, that would icliver a dose rate 80%

of the whole body dose rate limit, Dw3 Dw, = whole body dose rate limit of 500 mren@T. ,

( = equivalent whole body dose factor, in mren4T per weighted by the historical radionuclide distribution.

yCi/m' (X/Q)srm = 8.91E-06, the highest calculated annual average dispersion parameter, in sec/m', at the Site Boundary for any of the 3 units, from Table 3-2, 0.8 = administrative factor to compensate for any unexpected variability ,

in the radionuclide mix and to ensure that Site Boundary dose rate !

limits will not be exceeded.

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14 ODCM Rev. 7 .

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(D 3g) (0.8) l

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Osm,sg =

(2-4)

(L+1.1M), (X/Q)smi l

Where:

Osnr sg = the site release rate limit,in Ci/sec, that would deliver a dose rate 80% of the skin dose rate limit, Dst.

Dsg = skin dose rate limit of 3000 mrem /yr.

(L+1.lM), = equivalent skin dose factor,in mrem /yt per pCi/m', weighted by the radionuclide distribution.

(X/Q)srm = 8.91E-06, the highest calculated annual average dispersion parameter, in sec/m', at the Site Boundary for any of the three units, from Table 3-2.

0.8 = administrative factor to compensate for any unexpected variability in the radionuclide mix and to ensure that Site Boundary dose rate limits will not be exacded.

After determination of the Ostmwhole body and skin dose rates (equations 2-3 and 2-4, respectively), the most conservative result will be used as Ostm, the site release :

rate limit.

2.1.2.3 Unit Release Rate Limits (Quwrrl Typically Ostrewill be divided equally among operating units. If operational history dictates a larger fraction of the Ostm be assigned to a specific unit then a weighted average of each unit's contribution to the Osrmwill be utilized to determine the Quxrn

=

Qusrr (fusrr) (Ostm) (2-5) where:

Qusrr = unit release rate limit,in yCi/sec.

= the fraction (s 1) of noble gas historically released from a specific fusrr operating unit to the total of all noble gas released from the site.

Ostm = the site release rate limit,in uCi/sec determined in section 2.1.2.2.

15 ODCM Rev. 7

i 2.1.2.4 Setpoint Determination ,

i To comply with the requirements in Section 2.1, the alarm / trip setpoints can now bc ,

established using the unit release rate limit (Quwrr) to ensure that the noble gas ,

releases do not exceed the dose rate limits.

To allow for multiple sources of releases from different or common release points, the effluent monitor setpoint includes an administrative factor which allocates a ;

percentage of the unit release rate limit to each of the release sources. Monitor !

setpoints will also be adjusted in accordance with Nuclear Administrative and 6 l Technical Manual procedures to account for monitor-specific characteristics. ;

i Monitors RU-141. RU-143. and RU-145 i The alarm / trip setpoint for Monitors RU-141, RU-143, and RU-145 is calculated as '

follows:

P Monitor (Quwn) (a) f Setpoint s (2-6) ;

(472) (Flow Rate) l Where:

Monitor Setpoint = the setpoint for the effluent monitor,in Ci/ce, which provides a safe margin of assurance that the allowable dose rate limits will not be exceeded. ,

l Quurr = unit release rate limit, in vCi/sec, as determined in Section 2.1.23. ,

Flow Rate = the flow rate,in cfm, from flow rate monitors or the fan design flow '

rate for the release source under consideration.

1 l

472 = conversion factor, cubic centimeter /second per cubic feet / minute.

l a = fraction of Qusrr allocated for a specific release point. The sum of ,

these administrative values shall be less than or equal to one.

l Note - The RU-141 setpoints may be further reduced to provide earlyindication of steam generator tube leakage.'

l 16 ODCM Rev. 7

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Monitor RU-12 The alarm! trip setpoint for Monitor RU-12, the Waste Gas Decay Tank Monitor, is calculated as follows:

Monitor [(Quxrr)(a)(0.9)-(H)(PF)(472)}

setpoint s (2-7)

(Flow Rate)(472)

Where:

l Monitor l Setpoint = the setpoint for the monitor, in uCi/cc at STP, which provides a safe margin of assurance that the allowable dose rate limits will not be exceeded. ,

Quxrr = unit release rate limit,in Ci/sec, as determined in Section 2.1.23.

l l

l Flow Rate = flow rare, in cfm at STP at which the tank will be released.

PF = the current process flow of the plant vent in CFM. ,

i H = the current plant vent monitor concentration in pCi/cc.

a = fraction of Qusrr allocated for a specific release point. This ,

administrative value should be equal to or less than the administrative value used for the Plant Vent.

0.9 = an administrative value to account for potential increases in activity from other contributors to the same release point. i 472 = conversion factor, cubic centimeter /second per cubic feet / minute.

If there is no release associated with this monitor, the monitor setpoint should be established as close as practical to background to prevent spurious alarms, and yet assure an alarm should an inadvertent release occur.

2.1.2.5 Monitor Calibration I

The Radiation Level Conversion Factor (RLF) for each monitor is entered into the Radiation Monitoring Splem Database and may change whenever the monitor is calibrated. Calibration is performed in accordance with Nuclear Administrative and i Technical Manual procedures.

I l

1

)

i 17 ODCM Rev. 7 1

1 1

3.0 Gaseous and Uquid Effluent Dose Rates !

3.1 Requirements

Gaseous Eftluents l The dose rate due to radioactive materials released in gaseous effluents from the site (see Figures 6-4 and 6-5) shall be limited to the following:

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

3000 mrems$r to the skin, and

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

Applicability: At all times. This requirement does_ not apply to the Condenser Vacuum Pump l Exhaust if DCP 13-PJ SQ-065 has been implemented.

i I

Action: :

With the dose rate (s) creceding the above limits, immediately decrease the release rate to within the above limits (s).

3.1.1 Surveillance Requirements:

i

a. The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods contained in Section 3.1.2.

[

b. The dose rate due to 1-131,1-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods contained in Section 3.1.2 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3-1.

I l

l l

18 ODCM Rev. 7

I i t

]

i 3.1.2 Implementation of the Requirements: l Noble Gases !

l Noble gas activity monitor setpoints are established at release rates which permit corrective j action to be taken before exceeding offsite dose rates mrresponding to the 10 CFR 20 annual dose limits as described in Section 2.0. The requirements for sampling and analysis of l continuous and batch effluent releases are given in Table 3-1. The methods for sampling and analysis of continuous and batch effluent releases are given in the Nuclear Administrative and j

'lechnical Manual procedures. The dose rate in unrestricted areas shall be determined using the following equations. ;

l For whole body dose rate: ;

Dwn =

L [(R) (X/0)stm (O)) (3-1) !

For skin dose rate ,

Dg 3

= L [(1 4+ 1.1M,) (X/Q)srm (O.)] (3-2) l Where: i i

l

& = the whole body dose factor due to gamma emissions for each identified ;

noble gas radionuclide i, in mrem /yr per Ci/m' from Table 3-3. {

Q, = the release rate of radionuclide i, in pCi/sec. ~ !

l (X/Q)srm = 8.91E-06, the highest calculated annual average dispersion parameter, j in sec/m', for any of the three units, from Table 3-2. i l

Dws = the annual whole body dose rate (mrem /yr.). j 14 = the skin dose factor due to the beta emissions for each identified noble I gas radionuclide i, in mrem /yr per yCi/m' from Tabta. 3-3. .

Mi = the air dose factor due to gamma emissions for each identified noble gas I radionuclide i, in mradlyr per Ci/m' from Table 3-3. !

Dsg = the annual skin dose rate (mrem,6yr). I 1.1 = unit conversion constant of 1.1 mrem / mrad converts air dose to skin dose.

19 ODCM Rev. 7

_,.7

?

I-131.1-133. tritium and radionuclides in particulate form with half-lives creater than 8 days The methods for sampling and analysis of continuous and batch releases for I-131,1-133, tritium and radionuclides in particulate form with half-lives greater than 8 days, are given in j the applicable Nuclear Administrative and Technical Manual procedures. Additional monthly ;

and quarterly analyses shall be performed in accordance with Table 3-1. The total organ dose rate in unrestricted areas shall be determined by the following equation:

i D. - L I(P,)(X/Q)stm (Q)] (3-3) :

Where:

P, = the dose factor, in mrem /yr per Ci/m', for radionuclide i, for the inhalation pathway, from Table 3-4.

X/Q)srm = 8.91E-06, the h, ~ :st calculated annaal average dispersion parameter, in sec/m', at the Site Boundary, for any of the three units, O, = the release rate of radionuclide i, in Ci/see D. = the total organ dose rate (mrem /yr).

0 1

l i

20 ODCM Rev. 7 1

I i

i 1

i

f i

TABLE 3-1 ;

i RADIOACTIVE GASEOUS WASTE SAMP11NG AND ANAIXSIS PROGRAli l MINIMUM LOWER llMIT :

SAMPLING ANALYSIS TYPE OF OFDEFFLTION l GASEOUS RELEASE TYPE FREOUENCY FREOUENCY ACTITTTY ANALYSIS (Ill)) (uCUml)'_ [

P P i A. Waste Gas Storage Each Tank Each Tank Principal Gamma l Grab Sample Emitters

1.0E-04 {

P P .

H. Containment Purge Each Purge

Each Purge

Principal Gamma l Emitters 5 1.0E-04 l

Grab Sample !

H-3 1.0E-06 .

C. 1. Condenser Vacuum M* M* Principal Gamma f Pump Exhaust Emitters

1.0E-04 i

2. Plant Vent Grab Sample l'

3. Fuel Eldg. Exhaust H-3 1.0E-06 I

Continuous' 4/M d 1-131 1.0E-12 i Charcoal i, Sample I-133 1.0E-10 ;

t Continuous' 4/M' Principal Gamma !

Particulate Emitters 8 1.0E-11 :

Sampic (I-131, Others) j Continuous' M Gross Alpha 1.0E-11 ,

Composite l Particulate l l Sample i Continuous' O Sr-89, Sr-90 1.0E-11 f' Composite Par:iculate ;

Sample D. All Radwaste Types Continuous' Noble Gas Noble Gases 1.0E-06 j as listed in A., B, Monitor Gross Beta j and C. above. or Gamma l

21 ODCM Rev. 7

Table 3-1 (Continued) i TABLE NOTATION ,

a. The LLD is the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely ,

concluding that a blank observation represents a real signal.

For a particular measurement system (which may include radiochemical separation):

4.66 s, LLD = !

E

V

2.22E6

Y

cxp ( -lat )

Where:

LLD is the a priori lower limit of detection as defined above (as Ci per unit mass or volume).

Current literature defines the LLD as the detection capability for the instrumentation only and the MDC minim um detectable concentration, as the detection capability for a given instrument, procedure and type of sample.

s,is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per transformation), !

V is the sample size (in units of mass or volume),

1 2.22E6 is the number c4 trandormations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and At is the clapsed time between the midpoint of sample collection and time of counting (for plant effluents, not emironmental samples).

r The value of s, used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry the background should include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y, and ;

At should be used in the calculation. t j 11 should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measuiement system and not as an a posteriori (atter the fact) limit for a particular measurement *.

For a more complete discussion of the LLD, and other dctection limits, see the following:

l (1) HASL Procedures Manual, HASI 300 (revised annually).

l l

(2) Currie, L A., ' Limits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry" Anal. Chem. 40. 586-93 (1968).

(3) Hartwell, J. K.," Detection Limits for Radioisotopic Counting Techniques", Atlantic Richfield Hanford Company Reports ARH-2537 (June 22,1972).

22 ODCM Rev. 7 i

l I

l

Table 3-1 (Continued)

TABLE NOTATION b~ Analyses shall also be performed following SHUTDOWN, STARTUP, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period if 1) analysis shows that the DOSE EQUIVALENT I-131 mncentration in the primary coolant has increased more than a factor of 3; and 2) the noble gas activity monitor on the plant vent shows that effluent activity has increased by more than a factor of 3. If the associated noble gas vent monitor is inoperable, samples must be obtained as soon as possible. Analyses shall be performed within a four-hour period.

This requirement does not apply to the Fuel Building Exhaust.

l c Sampling and analyses shall also be performed at least once per 31 days when purging time exceeds 30 days continuous.

l d Samples shall be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l after changing (or after removal from sampler). When samples mllected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

e Tritium grab samples shall be taken at ist monthly from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuci p(x>l.

f The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time Mod covered by each dose or dose rate calculation made in accordance with Requirements 3.1,4.1 a ad 4.2 of the ODCM.

g The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xc-133m, Xc-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134 Cs-137, Cc-141 and Cc-144 for particulate emissions.

'Diis list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides shall also be identified and reported in the Semiannual Radioactive Effluent Release Report.

23 ODCM Rev. 7 i

TABLE 3-2 DISPERSION AND DEPOSITION PARAMETERS FOR 1ANG TERM RELEASES AT THE SITE BOUNDARY UNIT 1 UNIT 2 UNIT 3 DISTANCE X/Q D/Q DISTANCE X/Q D/Q DISTANCE X/Q D/Q DIRECTION (METERS) 3 (SEC/m 1 Ing(21 (METERS) S (SEC/m 1 .(g21 (METERS) (SEC/m'l .(_qf2 1 N 1037 4.93E-06 9.24E-09 1318 3.85E 06 6.17E-09 1661 3.54E-06 4.86E-09 NNE 1057 4.14E-06 1.19E-08 1342 3.18E-06 7.93E-09 1693 2.86E-06 6.23E-09 NE 2206 2.84E-06 6.84E-09 2545 2.42E-06 5.34E-09 2756 2.21E-06 4.65E-09 ENE 1967 2.51E-06 4.43E-09 2206 2.22E-06 3.64E-09 2337 2.08E-06 3.30E-09 E 1927 2.56E-06 3.24E-09 2163 2.27E-06 2.66E-09 2290 2.14E-06 2.41E-09 ESE 1967 2.61E-06 2.46E-09 2067 2.32E-06 2.11E-09 2023 2.37E-06 2.10E-09 SE 2049 3.56E-06 2.36E-09 2101 3.47E-06 2.26E-09 2256 3.24E-06 2.00E-09 SSE 2730 3.80E-06 1.58E-09 3026 3.43E-06 1.32E-09 2786 3.72E-06 1.52E-09 5 3006 5.07E-06 1.78E-09 2699 5.16E-06 1.97E-09 2346 5.90E-06 2.51E-09 SSW 2258 6.52E-06 3.20E-09 1836 7.90E-06 4.56E-09 1607 8.91E-06 5.73E-09 SW 1487 7.47E-06 5.65E-09 1208 7.72E-06 6.88E-09 1057 8.68E-06 8.61E-09 WSW 1251 4.52E-06 5.93E-09 1014 5.55E-06 8.44E-09 889 5.34E-06 8.83E-09 W 1225 4.73E-06 9.49E-09 993 5.86E-06 1.34E-08 871 6.72E-06 1.67E-08 WNW 1244' 3.76E-06 6.76E-09 1010 4.67E-06 9.60E-09 885 5.37E-06 1.19E-08 NW 1254 3.43E-06 5.87E-09 1191 3.62E-06 6.40E-09 1045 4.17E-06 7.98E-09 NNW 1069 3.70E-06 7.26E-09 1342 2.85E-06 4.87E-09 1561 2.93E-06 4.58E-09

Reference:

Distances are from the PVNGS ER-OL, Table 2.3-33. Dispersion and Deposition parameters are from a September,1985, calculation by NUS Corporation based on 9 years cf meteorological data; NUS Corporation letter NUS-ANPP-1386, dated October 4, 1985.

24 ODCM Rev. 7

i TABLE 3-3 i i

DOSE FACIORS FOR NOBLE GASES AND DAUGHTERS l

Whole Body Skin Gamma Air Beta Air Dose Factor Dose Factor Dose Factor Dose Factor ;

K 4 % N 3 Radionuclide mrem-m' mrem-m 5 mrad-m' mrad-m*

l yr-pCi yr-pCi yr-pCi yr-pCi l

l Kr-83m 7.56E-02 1.93E+01 2.88E+02 l Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 l

Kr-88 1.47E+D4 2.37E+03 1.52E+ 04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 l Xc-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xc-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xc-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xc-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03

Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 l

l i

Reference:

Regulatory Guide 1.109, Table B.I.

l 25 ODCM Rev. 7 l

l

Table 3-4 1

P t Values for the Inhalation Pathway (arem/yr/pCi/m 3) l l

t NUCLIDE Ace Group Orcan P. !

I l H-3 TEEN LIVER 1.27E+03 l CR-51 TEEN UUNG 2.10E+04 MN-54 TEEN LUNG 1.98E+06 !

FE-59 TEEN LUNG 1.53E+06 l CO-58 TEEN LUNG 1.34E+06 l CO-60 TEEN LUNG 8.72E+06 3 ZN-65 TEEN LUNG 1.24E+06 SR-89 TEEN LUNG 2.42E+06 I

! SR-90 TEEN BONE 1.08E+08 ZR-95 TEEN LUNG 2.69E+06 SB-124 TEEN LUNG 3.85E+06 l

I-131 CHILD THYROID 1.62E+07 ;

I-133 CHILD THYROID 3.85E+06 CS-134 TEEN LIVER 1.13E+06 i CS-137 CHILD BONE 9.07E405 BA-140 TEEN LUNG 2.03E+06 i CE-141 TEEN LUNG 6.14E+05 CE-144 TEEN LUNG 1.34E+07 r I

i l

26 ODCM Rev. 7 i

i l

3.2 Requirements

Secondary System IJguid Waste Discharges To Onsite Evaporation Ponds -

Concentration l The concentration of radioactive material discharged rsom secondary system liquid waste to the onsite ;

l evaporation ponds shall be limited to the lower limb of detectability (LLD) defined as 5.0E-07 Ci/ml l l for the principal gamma emitters or 1.0E-06 yCt/ml for 1-131.

l Applicability: At all times.

l Action:

When any secondary system lia aid waste discharge pathway concentration determined in acmrdance with the surveillance requircments given below exceeds the specified LLD, divert that discharge pathway to the liquid radwaste system without delay.

3.2.1 Surveillance P.equirements:

a. Radioactive liquid wastes collected in the chemical waste neutralizer tank shall be sampled !

and analyzed prior to their batchwise discharge to the onsi'e evaporation pond in a'.cordance with the sampling and analysis program specified in Table 3-5.

n. With the concentration of radioactive material in the chemical waste neutralizer tank exceeding the specified LLD, sample and analyze other secondary system discharge pathways in accordance with the sampling and analysis program specified in Table 3-5.

3.2.2 Implementation of the Requirements:

This requirement is implemented by Nuclear Administrative and Technical Mr nual procedures.

j* For the duration required to recover from the Unit 2 SGTR occurring on March 14,1993 the following limits for principle gamma emitters and tritium apply to Unit 2(these limits are temporary and will te g

g removed from the ODCM when recovery is deemed complete by the Site Chemistry General Manager):

I ISOTOPE Concentration limit Cumulative activity limit for (uCi/cc) discharges to the evaporation a pond (Ci)

Gamma emmitter with half life 1 3.0E-06 No limit l less than or equal to 2 years

, (excluding 1-131) l Co-60 5 3.0E-05 5 2.0E-01 l Cs-134 5 9.0E-06 5 6.0E-02 l Cs-137 5 2.0E-05 s 1.5E-01 l I-131 s 1.0E-06 No limit l Other gamma emitter s 5.0E-07 No limit l H-3 < 5.0E-03 < 3.8E+ 01 27 ODCM Rev. 7 l i

l I

1

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TABLE 3-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANAIXSIS PROGRAM .

t LOWER LIMTT SECONDARY SYSTEM MINIMUM TYPE OF OFDEITfrION LIQUID RELEASE SAMPLING ANALYSIS ACTIVrIY (LLD)* ,

PATIIWAY FREOUENCY FREOUENCY ANAIXSIS (.filmfl l

?

A. Batch dischartes*

1. Chemical Waste P P Principal Gamma 5.0E-07 i Neutralizer Tank Each Each Emitters

l Batch Batch l I-131 1.0E-06 l i

2. Steam Generator P P Principal Gamma 5.0E-07 i Blowdown low Each Each Emitters * ,

TDS Sump

Batch Batch '

I-131 *0E-06 i

3. Condensate P P Principal Gamma 5.0E-07 l Polishing low Each Each Emitters * '

TDS Sump

Batch Batch 1-131 1.0E46 B. Continuous Releases

7

1. Turbine Building D D Principal Gamma 5.0E-07 l Grab Grab Emitters

Sump

Sample Sample l I-131 1.0E-06 ;

2. Condenser Area D D . Principal Gamma 5.0E-07 Sumps

Grab Grab Emitters * .,

Sample Sample l I-131 1.0E46 i Sampling and analysis for pathways 2 and 3 under htch discharges and 1 and 2 under continuous releases are required only when concentration for chemical waste neutralizer tank pathway exceeds the LLD.

28 ODCM Rev. 7 l

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l i

Table 3-5 (Continued)

TABLE NOTATION i

a. The LLD is defined as the smallest concentration of radioactie material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5%

probability of falsely concluding that a blank observation represents a "real' signal.

For a particular measurement system which may include radiochemical separation:

4.66 s.

LLD =

E

V

2.22E6

Y

cxp ( -AAt )

Where:

LLD is the 'a priori" lower limit of detection as defined above as microcuries per unit mass or volume, s,is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate as counts per minute, E is the counting efficiency as counts per disintegration, V is the sample size in units of mass or volume, 2.22E6 is the number of disintegrations per minute per microcurie Y is the fractional radiochemical yield when applicable, A is the radioactive decay constant for the particular radionuclide, and l t At is the clapsed time between midpoint of sample collection and time of counting. 6 l

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

It should be recognized that the LLD is defined as an a priori (before the fact) hmit representing the capability of a measurement system and not as an a nosteriori (after the fact) limit for a particular measurement.

b. A batch release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

c. The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58 Co-60, Zn-65, Mo-99, Cs-134 Cs-137, and Cc-141. Cc-144, shall also be measured, but with an LLD of 5.0E-06. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8.

l

[ d. A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

29 OLCM Rev. 7 I

l

i l 4.0 Caseous & IJguid Effluents - Dose

4.1 Requirements

Noble Gases f

The air dose due to noble gases released in gaseous effluents, from each reactor unit to areas at and beyond the SITE BOUNDARY (see Figure 6-4 and 6-5) shall be limited to the following:

a. During any calendar quarter: Ins than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation and,

b. During any calendar year: Ins than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

Applicability: At all times.

I j Action:

With the calculated air dose from radioactive noble gases in gaseous effluents excx:cding any of the alec limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifics the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions l to be taken to assure that subsequent releases will be in mmpliance with the above limits.

l 4.1.1 Surveillance Requirements:

(

1

a. Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology contained in Section 4.1.2 at least once per 31 days.

30 ODCM Rev. 7

1 I

4.1.2 Implementation of the Requirement: Noble Gas l l The air dose in unrestricted areas beyond the site boundary due to noble gases released in gaseous ellluents from each unit during any specified time period shall be determined by the I following equations: !

For gamma radiation:

1 D y, = (3.17E-08) L [(M,) (X/0)uurr(0,)] (4-1) ]

i For beta radiation. .

l D B,, = (3.17E-08) L [(N,) (X/0)uun(0,)] (4-2)

Where:

M, = the air dose factor due to gamma emissicns for each identified noble gas l l radionuclide i, in mrad /yr per uCi/m' from Table 3-3. ,

N, = the air dose factor due to beta emissions for each identified noble gas l radionuclide i, in mrad $T per Ci/m' from Table 3-3.

l (X/0)uurr = the highest calculated annual average dispersion parameter, in sec/m',

at the site boundary for the particular unit, from Table 3 2.

= 7.47E-06 from Unit i

= 7.90E-06 from Unit 2

= 8.91E-06 from Unit 3 D y,, = the total gamma air dose, for the particular unit, in mrad, due to noble gases released in gaseous effluents for a specified time period at the SITE BOUNDARY.

D p, = the total beta air dose, for the particular unit, in mrad, due to noble gases released in gaseous effluents for a specified time period at the SITE BOUNDARY. I O, = the integrated release, from the particular unit, in Cl, of each identified noble gas radionuclide i, in gaseous effluents for a specified time period.

3.17E-08 = the inverse of seconds in a year (>T/sec).

The cumulative gamma air dose and beta air dose for a quarterly or annual evaluation shall be based on the calculated dose contribution from each specified time period occurring during the reporting time period.

31 ODCM Rev. 7

4.2 Requirement

Iodine - 131, Iodine-133, Tritium, and All Radionuclides in Particulate Form With Italf-IJves Greater Than 8 Days The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figures 6-4 and 6-5) shall be ;

limited to the following:

a. During any calendar quarter: less than or equal to 7.5 mrems to any organ and,

b. During any calendar year: less than or equal to 15 mrems to any organ.

Applicability: At all times.

Action:

With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the atKwe limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identi'ics the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

4.2.1 Surveillance Requirements:

a. Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters contained in Section 4.2.2 at least once per 31 days.

i l

l 32 ODCM Rev. 7

4.2.2 Implementation of the Requirement The organ dose to an individual from 1-131,1-133, tritium, and all radionuclides in particulate form,with half 4ives greater than eight days,in gaseous cinuents released to unrestricted areas from each reactor unit is calculated using the following expressions:

D. =

(3.17E-08) L [L (R W ) (Q)] (4-3) i Where:

l D. = the total accumulated organ dose from gaseous cinuents for a particular

unit. to a MEMBER OF THE PUBLIC, in mrem, at the SITE BOUNDARY or at the controlling location.

Q, = the quantity of radionuclide i,in yCi, released in gaseous effluents from a particular unit.

R, = the dose factor for each identified radionuclide i, for pathway k (for the

! inhalation pathway in mrem /yr per pCi/m' and for the food and ground [

l plane pathways in mr - mrem /yr per pCi/sce, except H-3, which has units of mrem /yr per pCi/m') at the controlling location. The R.'s for each ;

age group are given in Tables 4-1 through 4-15. !

l 3.17E418 = the inverse of seconds per year (yr/sec). ,

l W. = the highest annual average dispersion or deposition parameter for the l l particular unit, used for estimating the dose at the site boundary or to l a MEMBER OF THE PUBLIC at the controlling location for the ,

l particular unit.

= (X/Q) uun, in sec/m' for the inhalation pathway and for all tritium calculations, for organ dose at the site boundary, from Table 3-2.

1

= 7.47E46 from Unit 1

= 7.90E46 from Unit 2

= 8.91E46 from Unit 3 i

= (X/Q)usn, in sec/m' for the inhalation pathway and for all tritmm I calculations, for organ dose at the controlling location, from Table 4-16.

= 2.92E46 from Unit 1 l

= 2.19E46 from Unit 2

= 2.31E4M from Unit 3

= (D/Q)uun, in m 2, for the food and ground plane pathways, for organ dose at the site boundary, from Table 3-2.

= 1.19E-08 from Unit 1

= 1.34E-08 from Unit 2

= 1.67E-08 from Unit 3 1

33 ODCM Rev. 7

=

(D/Q) in m-2, for the food and ground plane pathways, for organ i dose at the controlling location, from Table 4-16.

= 3.25E-09 from Unit 1

= 3.88E-10 from Unit 2 ,

= 4.21E-10 from Unit 3 Residences, vegetable gardens and milk animals located within 5 miles of the site will be identified during the annual land use census. The controlling pathway and location will be identified and will be used for all MEMBER OF THE PUBLIC dose evaluations.

The R; values were calculated in amordance with the methodologies in NUREG-0133. The following site specific information was used to calculate R; Value The length of the grazing season for milk animals (f).

Ref. ER-OL, Section 2.13.43 0.75 The length of the grazing season for meat animals (f,).

Ref. ER-OL, Section 2.13.4.4 0.25 The fraction of daily feed derived from pasture while on pasture for milk animals (f,).

Ref. ER-OL, Section 2.13.43 035 l The fraction of daily feed derived from pasture l while on pasture for meat animals (f,).

! Ref. ER-OL, Section 2.13.43 0.05 l The fraction of year vegetables are grown, (f) i approximation.

! Ref. ER-OL, Section 2.13.4, Table 2.1-8. 0.667 l The annual absolute humidity (g/m'), H, Ref. UFSAR, Table 23-16 6 l

l l

l 1

34 ODCM Rev. 7

4.3 Requirements

Gaseous Radwaste Treatment t

De GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, from each reactor unit, from the site (see Figures 6-4 and 6-5) when averaged over 31 days, would exceed 0.2 mrad for gamma .

radiation and 0.4 mrad for beta radiation. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, to areas at and beyond l

the SITE BOUNDARY (see Figures 6-4 and 6-5) when averaged over 31 days would exceed 0.3 mrem i i to any organ of a MEMBER OF THE PUBLIC.

l Applicability: At all times: ,

Action:

With radioactive gaseous waste Ning discharged without treatment and in excess of the above limits, I

prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

a. Identification of the inoperable equipment or subsystems and the reason for inoperability,

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

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

l

! 4.3.1 Surweillance Requirements:

I

a. Doses due to gaseous releases from the site shall be projected at least once per 31 da3s,in accordance with the methodology and parameters in Section 4.3.2.

4.3.2 Implementation of the Requirement Where possible, consideration for expected operational evolutions (i.e., outages, etc.) should be taken in the dose projections.

Dose Proicction. Noble Gases The air dose,in mrads for the current quarter is determined using the methodology described in Section 4.1.2. His information is used to determine an air dose projection for the next 31 l days using the following equations:

l l

l 35 ODCM Rev. 7

l I

For gamma radiation: l 31 day y = (Dy qtr/rgtr) 31 + CDy (4-4)

For beta radia'. ion: ,

I 31 day B = (DB qtr/Tqtr) 31 + CDB (4-5) l Where: r Dy qtr = the total gamma air dose due to noble gases released in gaseous effluents for the current quarter,in mrads, at the site boundary.

DB qtr = the total beta hir dose due to noble gases released in gaseous effluents for the current quarter, in mrads, at the site boundary. t Tqtr = the time period,in days, over which Dy qtr and DB qtr were integrated.

31 = the number of days over which the dose projections are made.

{

31 day y = the 31 day projected gamma air dose due to noble gases released in gaseous effluents, in mrads, at the site boundary. l 31 day B = the 31 day projected beta air dose due to noble gases released in gaseous effluents, in mrads, at the site boundary.

CDy = any current or projected gamma air dose,in mrads, due to noble gases i released in gaseous effluents, which could have a significant impact on 31 day y.

l CDB = any current or projected beta air dose, in mrads, due to noble gases .

released in gaseous effluents, which could have a significant impact on 31 day S.

i l

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l l l

36 ODCM Rev. 7 I

i

, I l

l

i t

Dose Projection 131.1-133. tritium. and all radionuclides in particulate form with half-lives creater than cicht dass The organ dose, in mrem, for a particular unit, for the current quarter is determined using the methodology described in Section 4.2.2 of this manual. This information is used to determine an organ dose projection for the next 31 days using the following equation:

31 day, = (D.qtr/Tqtr)31 + CD, (4-6) where: 1 D, qtr = the total organ dose from a particular unit due to I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days, released in gaseous effluents for the current quarter, in mrem.

Tqtr = the time period, in days, over which D,qtr was integrated.

l 31 = the number of days over which the dose projections are made. l l

! r 31 day, = the 31 day projected organ dose, in mrem, from a particular unit. l CD, = any current or projected organ dose for a particular unit, in mrem, !

which could have a significant impact on 31 day,. ;

i l

i !

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f l

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l 37 ODCM Rev. 7 i

= _ - - . . . _ . - ___

TABLE 4-1 ;

Ri DOSE CONVERSION FACTORS FOR THE GROUND PLANE PATHWAY >

NUCLIDE T. BODY SKIN H-3 0.00E+00 0.00E+00 CR-51 4.66E+06 5.51E+06 l MN-54 1.39E+09 1.63E+09 FE-59 2.73E+08 3.21E+08 CO-58 3.79E+08 4.44E+08 CO-60 2.15E+10 2.53E+10 ,

ZN-65 7.47E+08 8.59E+08 SR-89 2.16E+04 2.51E+04 SR-90 0.00E+00 0.00E+00 ZR-95 2.45E+08 2.84E+08 SB-124 5.98E+08 6.90E+08 '

I-131 1.72E+07 2.09E+07 I-133 2.45E+06 2.98E+06 i CS-134 6.86E+09 8.00E+09 CS-137 1.03E+10 1.20E+10 BA-140 2.05E+07 2.35E+07 CE-141 1.37E+07 1.54E+07 i CE-144 6.95E+07 8.04E+07 i

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38 ODCM Rev. 7 i

TABLE 4-2 :

i Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - ADULT RECEPTOR t r

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI l H-3 0.00E+00 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 .

. CR-51 0.00E+00 0.00E+00 4.00E+04 2.39E+04 8.82E+03 5.31E+04 1.01E+07 l MN-54 0.00E+00 2.97E+08 5.66E+07 0.00E400 8.83E+07 0.00E+00 9.09E+08 r l FE-59 1.14E+08 2.68E+08 1.03E+08 0.00E+00 0.00E+00 7.49E+07 8.93E+08 CO-58 0.00E+00 2.84E+07 6.38E+07 0.00E+00 0.00E+00 0.00E+00 5.76E+08

.. Co-60 0.00E+00 1.59E+08 3.51E+08 0.00E+00 0.00E+00 0.00E+00 2.99E+09 >

l ZN-65 3.00E+08 9.56E+08 4.32E+08 0.00E+00 6.39E+08 0.00E+00 6.02E+08 ,

SR-89 9.08E+09 0.00E+00 2.61E+08 0.00E+00 0.00E+00 0.00E+00 1.46E+09 SR-90 5.76E+11 0.00E+00 1.41E+11 0.00E+00 0.00E+00 0.00E+00 1.67E+10

( ZR-95 1.08E+06 3.47E+05 2.35E+05 0.00E+00 5.45E+05 0.00E+00 3.10E+09 l SB-124 9.53E+07 1.80E+06 3.78E+07 2.31E+05 0.00E+00 7.42E+07 2.71E+09 I-131 5.49E+07 7.85E+07 4.50E+07 2.57E+10 1.35E+08 0.00E+00 2.07E+07 I-133 1.39E+06 2.42E+06 7.38E+05 3.56E+08 4.22E+06 0.00E+00 2.17E+06 l

CS-134 4.44E+09 1.06E+10 8.64E+09 0.00E+00 3.42E+09 1.13E+09 1.85E+08 6.06E+09 8.29E+09 5.43E+09 0.00E+00 2.81E+09 9.36E+08 1.60E+08 CS-137 j EA-140 9.43E+07 1.19E+05 6.18E+06 0.00E+00 4.03E+04 6.78E+04 1.94E+08 {

l f.E-141 1.73E+05 1.17E+05 1.33E+04 0.00E+00 5.44E+04 0.00E+00 4.48E+08 l CE-144 3.12E+07 1.30E+07 1.67E+06 0.00E+00 7.73E+06 0.00E+00 1.05E+10 .

l t

i

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5 i

l 39 ODCM Rev. 7 l

l

l TMLE 4-3 l I

Ri DOSE CONVERSION FACTORS FOR THE VMGETATION PATHWAY - TEEN RECEPTOR I

NUCLIDE BONE LIVER T . BODi THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 3.36E+03 3.36E+03 3.36E+03 3.36E+03 3.36E+03 3.36E+03 CR-51 0.00E+00 0.00E+00 5.60E+04 3.11E+04 1.23E+04 7.99E+04 9.41E+06 MN-54 0.00E+00 4.41E+08 8.74E+07 0.00E+00 1.31E+08 0.00E+00 9.04E+08 FE-59 1.69E+08 3.94E+08 1.52E+08 0.00E+00 0.00E+00 1.24E+08 9.31E+08 CO-58 0.00E+00 4.16E+07 9.59E+07 0.00E+00 0.00E+00 0.00E+00 5.74E+08 CO-60 0.00E+00 2.42E+08 5.45E+08 0.00) - D.00E+0C 0.00E+00 3.15E+09 3 ZN-65 4.11E+08 1.43E+09 6.65E+08 0.00LuF. 9.12E+08 0.00E+00 6.04E+08 SR-89 1.43E+10 0.00E+00 4.10E+08 0.00E+00 0.00E+00 0.00E+00 1.70E+09 ,

SR-90 7.30E+11 0.00E+00 1.80E+11 0.00E+00 0.00E+00 0.00E+00 2.05E+10 ZR-95 1.64E+06 5.17E+05 3.56E+05 0.00E+00 7.60E+05 0.00E+00 1.19E+09 l SB-124 1.47E+08 2.70E+06 5.73E+07 3.33E+05 0.00E+00 1.28E+08 2.96E+09 l I-131 5.29E+07 7.41E+07 3.98E+07 2.16E+10 1.28E+08 0.00E+00 1.47E+07 ;

1-133 1.29E+06 2.19E+06 6.68E+05 3.06E+08 3.84E+06 0.00E+00 1.66E+06 ,

CS-134 6.90E+09 1.62E+10 7.53E+09 0.00E+00 5.16E409 1.97E+09 2.02E+08 I CS-137 9.86E+09 1.31E+10 4.57E+09 0.00E+00 4.46E+09 1.73E+09 1.87E+08 [

BA-140 1.07E+08 1. 31E+ 05 6.88E+06 0.00E+00 4.44E+04 8.80E+04 1.65E+08 -

CE-141 2.61E+05 1.74E405 2.00E+04 0.00E+00 8.19E+04 0.00E+00 4.98E+08 7 CE-144 5.11E+07 2.12E+07 2.75E+06 0.00E+00 1.26E407 0.00E+00 1.29E+10 4

[

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40 ODCM Rev. 7

i TABLE 4-4 Ri DOSE CONVERSION FACTORS FOR THE VEGETATION PATHWAY - CHILD RECEPTOR l l

NUCLIDES BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 5.23E+03 5.23E+03 5.23E+03 5.23E+03 5.23E+03 5.23E+03 CR-51 0.00E+00 0.00E+00 1.08E+05 6.02E+04 1.64E+04 1.10E+05 5.75E+06 !

MN-54 0.00E+00 6.49E+08 1.73E+08 0.00E+00 1.82E+08 0.00E+00 5.45E+08 FE-59 3.79E+08 6.13E+08 3.05E+08 0.00E+00 0.00E+00 1.78E+08 6.38E+08 CO-58 0.00E+00 6.21E+07 1.90E+08 0.00E+00 0.00E+00 0.00E+00 3.62E+08 CO-60 0.00E+00 3.70E+08 1.09E+09 0.00E+00 0.00E+00 0.00E400 2.05E+09 !

ZN-65 7.93E+08 2.11E+09 'l.31E+09 0.00E+00 1.33E+09 0.00E+00 3.71E+08 I SR-89 3.44E+10 0.00E+00 9.83E+08 0.00E+00 0.00E+00 0.00E+00 1.33E+09 SR-90 1.22E+12 0.00E+00 3.09E+11 0.00E+00 0.00E+00 0.00E+00 1.64E+10 '

ZR-95 3.72E+06 8.17E+05 7.27E+05 0.00E+00 1.17E+06 0.00E+00 8.52E+08 ,

SB-124 3.38E+08 4.39E+06 1.19E+08 7.47E+05 0.00E+00 1.88E+08 2.12E+09 I-131 9.95E+07 1.00E+08 5.68E+07 3.31E+10 1.64E+08 0.00E+00 8.90E+06 I-133 2.36E+06 2.91E+06 1.10E+06 5.41E+08 4.85E+06 0.00E+00 1.17E+06 .

CS-134 1.57E+10 2.57E+10 5.43E+09 0.00E+00 7.98E+09 2.86E+09 1.39E+08 !

CS-137 2.34E+10 2.24E+10 3.31E+09 0.00E+00 7.31E+09 2.63E+09 1.40E+08 BA-140 2.20E+08 1.93E+05 1.28E+07 0.00E+00 6.27E+04 1.15E+05 1.11E+08 CE-141 6.15E+05 3.07E+05 4.55E+04 0.00E+00 1.34E+05 0.00E+00 3.83E+08 CE-144 1.24E+08 3.89E+07 6.62E+06 0.00E+00 2.15E+07 0.00E+00 1.01E+10 i l

l

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41 ODCM Rev. 7 l

TABLE 4-5 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - ADULT RECEPTOR NUCLIDE BONE LTVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 4.33E+02 4.33E+02 4.33E+02 4. 33E+ 02 4.33E+02 4.33E+02 CR-51 0.00E+00 0. 00E+ 00 3.44E+02 2.06E+02 7.58E+01 4.57E+02 8.65E+04 MN-54 0.00E+00 2.71E+06 5.18E+05 0.00E+00 8.08E+05 0.00E+00 8.31E+06 FE-59 2.60E+07 6.11E+07 2.34E+07 0.00E+00 0.00E+00 1.71E+07 2.04E+08 l CO-58 0.00E+00 2.84E+06 6.36E+06 0.00E+00 0.00E+00 0.00E+00 5.75E+07 l CO-60 0.00E+00 2.61E+07 5.76E+07 0.00E+00 0.00E+00 0.00E+00 4.90E+08 l ZN-65 9.97E+07 3.17E+08 1.43E+08 0.00E+00 2.12E+08 0.00E+00 2.00E+08 [

SR-89 3.41E+07 0.00E+00 9.79E+05 0.00E+00 0.00E+00 0.00E+00 5.47E+06 SR-90 4.43E+09 0.00E+00 1.09E+09 0.00E+00 0.00E+00 0.00E+00 1.28E+08 ZR-95 2.68E+05 8.58E+04 5.81E+04 0.00E+00 1.35E+05 0.00E+00 2.72E+08 SB-124 2.67E+06 5.05E+04 1.06E+06 6.48E+03 0.00E+00 2.08E+06 7.59E+07 '

1-131 1.36E+05 1.94E+05 1.11E+05 6.37E+07 3.33E+05 0.00E+00 5.13E+04 I-133 4.56E-03 7.94E-03 2.42E-03 1.17E+00 1.39E-02 0.00E+00 7.14E-03 CS-134 2.17E+08 5.17E+08 4.23E+08 0.00E+00 1.67E+08 5.56E+07 9.05E+06 CS-137 3.11E+08 4.25E+08 2.78E+08 0.00E+00 1.44E+08 4. 79E+ 07 8.22E+06 I

BA-140 4.35E+05 5.46E+02 2.85E+04 0.00E+00 1.86E+02 3.13E+02 8.95E+05 CE-141 8.87E+02 6.00E+02 6.80E+01 0.00E+00 2.79E+02 0.00E+00 2.29E+06 CE-144 4.23E+05 1.77E+05 2.27E+04 0.00E+00 1.05E+05 0.00E+00 1.43E+08 ,

l

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l t

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42 ODCM Rev. 7

l TABLE 4-6 h l

Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - TEEN RECEPTOR ;

i i

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 2.58E+02 CR-51 0.00E+00 0.00E+00 2.75E+02 1.53E+02 6.03E+01 3.93E+02 4.62E+04 MN-54 0.00E+00 2.07E+06 4.11E+05 0.00E+00 6.18E+05 0.00E+00 4.25E+06 i l FE-59 2.08E+07 4.85E+07 1.87E+07 0.00E+00 0.00E+00 1.53E+07 1.15E+08 '

l CO-58 0.00E+00 2.19E+06 5.04E+06 0.00E+00 0.00E+00 0.00E+00 3.02E+07 CO-60 0.00E+00 2.03E+07 4.56E+07 0.00E+00 0.00E+00 0.00E+00 2.64E+08 ZN-65 7. 01E+ 07 2.43E+08 1.14E+08 0.00E+00 1.56E+08 0.00E+00 1.03E+08 SR-89 2.88E+07 0.00E+00 8.24E+05 0.00E+00 0.00E+00 0.00E+00 3.43E+06 ;

SR-90 2.87E+09 0.00E+00 7.08E+08 0.00E+00 0.00E+00 0.00E+00 8.05E+07 ZR-95 2.14E+05 6.76E+04 4.65E+04 0.00E+00 9.93E+04 0.00E+00 1.56E+08 SB-124 2.18E+06 4.02E+04 8.52E+05 4.95E+03 0.00E+00 1.91E+06 4.40E+07 I-131 1.13E+05 1.58E+05 8.49E+04 4.61E+07 2.72E+05 0.00E+00 3.13E+04 I-133 3.82E-03 6.48E-03 1.98E-03 9.04E-01 1.14E-02 0.00E+00 4.90E-03 CS-134 1.73E+08 4.07E+08 1.89E+08 0.00E+00 1.29E+08 4.94E+07 5.06E+06 CS-137 2.58E+08 3.43E+08 1.20E+08 0.00E+00 1.17E+08 4.54E+07 4.88E+06 BA-140 3.59E+05 4.40E+02 2.31E404 0.00E+00 1.49E+02 2.96E+02 5.54E+05 CE-141 7.45E+02 4.97E+02 5.71E+01 0.00E+00 2.34E+02 0.00E+00 1.42E+06 CE-144 3.56E+05 1.47E+05 1.91E+04 0.00E+00 8.80E+04 0.00E+00 8.96E+07 i

L k

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i 43 ODCM Rev. 7

l TABLE 4-7 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MEAT PATHWAY - CHILD RECEPTOR ,

l 1

NUCLIDES BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI l I

H-3 0.00E+00 3.12E+02 3.12E+02 3.12E+02 3.12E+02 3.12E+02 3.12E+02 CR-51 0.00E+00 0.00E+00 4.29E+02 2.38E+02 6.51E+01 4.35E+02 2.28E404 MN-54 0.00E+00 2.37E+06 6.31E+05 0.00E+00 6.64E+05 0.00E400 1.99E+06 i FE-59 3.68E+07 5.96E+07 2.97E+07 0.00E+00 0.00E+00 1.73E+07 6.20E+07 CO-58 0.00E+00 2.55E+06 7. 82E+06 0.00E+00 0.00E+00 0.00E+00 1.49E+07 ;

CO-60 0.00E+00 2.40E+07 7.09E+07 0.00E+00 0.00E+00 0.00E+00 1.33E+08 ZN-65 1.05E+08 2.80E+08 1.74E+08 0.00E+00 1.77E+08 0.00E+00 4.92E+07 SR-89 5.45E+07 0.00E+00 1.56E+06 0.00E+00 0.00E+00 0.00E+00 2.11E+06 ;

SR-90 3.70E+09 0.00E+00 9.39E+08 0.00E+00 0.00E+00 0.00E+00 4.99E+07 l ZR-95 3.81E+05 8.36E+04 7.45E+04 0.00E+00 1.20E+05 0.00E+00 8.73E+07 5 SB-124 3.95E+06 5.12E+04 1.38E+06 8.72E+03 0.00E+00 2.19E+06 2.47E+07 -i' I-131 2.09E+05 2.11E+% 1.iOE+05 6.96E+07 3.46E+05 0.00E+00 1.87E+04 I

I-133 7.09E-03 8.77E-03 3.32E-03 1.63E+00 1.46E-02 0.00E+00 3.53E ,

CS-134 3.05E+08 5.00E+08 1.06E+08 0.00E+00 1.55E408 5.56E+07 2.70E+06 ,

i CS-137 4.75E+08 4.55E+08 6.71E+07 0.00E+00 1.48E+08 5.33E+07 2.85E+06 i BA-140 6.63E+05 5.81E+02 3.87E+04 0.00E+00 1.89E+02 3.46E+02 3.36E+05 CE-141 1.40E+03 6.99E+02 1.04E+02 0.00E+00 3.07E+02 0.00E+00 8.72E+05 CE-144 6.72E+05 2.11E+05 3.58E+04 0.00E400 1.17E+05 0.00E+00 5.49E+07 l

l ,

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f 44 ODCM Rev. 7 i

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6 l

l TABLE 4-8 ,

F Ri DOSE CONVERSION FACTOR 3 FOR THE GRASS-COW-MILK PATHUAY - ADULT RECEPTOR I

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 1.02E+03 1.02E+03 1. 02E+ 03 1.02E+03 1.02E+03 1.02E+03 CR-51 0.00E+00 0.00E+00 8.28E+03 4. 95E+03 1.82E+03 1.10E+04 2.08E+06 [

l MN-54 0.00E+00 3.99E+06 7.61E+05 0.00E+00 1.19E+06 0.00E+00 1.22E+07 l FE-59 9.69E+06 2.28E+07 8.73E+06 0. 00E+00 0.00E+00 6.36E+06 7.59E+07 CO-58 0.00E+00 1.74E+06 3.90E+06 0.00E+00 0.00E+00 0.00E+00 3.53E+07 CO-60 0.00E+00 8.41E+06 1.85E+07 0.00E+00 0.00E+00 0.00E+00 1.58E+08 ;

ZN-65 6.34E+08 2.02E+09 9.12E+08 0.00E+00 1.35E+09 0.00E+00 1.27E+09 SR-89 4.90E+08 0.00E+00 1.41E+07 0.00E+00 0.00E+00 0.00E+00 7.86E+07 SR-90 2.43E+10 0.00E+00 5.96E+09 0.00E+00 0.00E+00 0.00E+00 7.02E+08 :

ZR-95 3.39E+02 1.09E+02 7.37E+01 0.00E+00 1.71E+02 0.00E+00 3.45E+05 SB-124 9 11E+06 1.72E+05 3.61E+06 2.21E+04 0.00E+00 7.09E+06 2.59E+08 I-131 7.77E+07 1.11E+ 08 6.37E+07 3.64E+10 1.91E+08 0.00E+00 2.93E+07 1-133 1.02E+06 1.77E+06 5.39E+05 2.60E+08 3.08E+06 0.00E+00 1.59E+06 CS-134 2.83E+09 6.73E+09 5.50E+09 0.00E+00 2.18E+09 7.23E+08 1.18E+08

! CS-137 3.83E+09 5.24E+09 3.43E+09 0.00E+00 1.78E+09 5.91E+08 1.01E+08 BA-140 7.11E+06 8.93E+03 4.66E+05 0.00E+00 3.04E+03 5.11E+03 1.l6E+07 ;

CE-141 8.73E+03 5.90E+03 6.70E+02 0.00E+00 2.74E+03 0.00E+00 2.26E+07 {

CE-144 1.01E+06 4.21E+05 5.41E+04 0.00E+00 2.50E405 0.00E+00 3.41E+08 ;

I i

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45 ODCM Rev. 7 l 1 I

4 l

TABLE 4-9 Ri DOSE CONVERSION FACTORS FOR THE CRASS-COW-MILK PATHWAY - TEEN RECEPTOR NUCLIDE BONE LIVER T, BODY THYROID KIDNEY LUNG GI-LLI j H-3 0.00E+00 1.33E+03 1.33E+03 1.33E+03 1.33E+03 1.33E+03 1.33E+03 CR-51 0.00E+00 0.00E+00 1.45E+04 8.03E+03 3.17E+03 2.06E+04 2.43E+06 MN-54 0.00E+00 6.64E+06 1.32E+06 0.00E+00 1.98E406 0.00E+00 1.36E+07 '

FE-59 1.69E+07 3.95E+07 1.52E+07 0.00E+00 0.00E+00 1.24E+07 9.33E+07 CO-58 0.00E+00 2.93E+06 6.76E+06 0.00E+00 0.00E+00 0.00E+00 4.04Et07 l

CO-60 0.00E+00 1.42E+07 3.21E+07 0.00E+ 00 0.00E+00 0.00E+00 1.86E+08 l

ZN-65 9.74E+08 3.38E+09 1.58E+09 0.00E+00 2.17E+09 0.00E+00 1.43E+09 SR-89 9.03E+08 0.00E+00 2.59E+07 0.00E+00 0.00E+00 0.00E+00 1.08E+08 SR-90 3.43E+10 0.00E+00 8.48E+09 0. 00E+ 00 0.00E400 0.00E+09 9.64E+08 ;

ZR-95 5.94E+02 1.87E+02 1.29E+02 0.00E+00 2.75E+02 0.00E+00 4.32E+05 SB-124 1.62E+07 2.99E+05 6.34E+06 3.69E+04 0.00E+00 1.42E+07 3.27E+08 '

I-131 1.41E+08 1.98E+08 1.06E+08 5.76E+10 3.40E+08 0.00E+00 3.91E+07 ,

I-133 1.86E+06 3.15E+06 9.60E+05 4.39E408 5.52E406 0.00E+00 2.38E+06 CS-134 4.91E+09 1.16E+10 5.36E+09 0.00E+00 3.67E+09 1.40E+09 1.44E+08 CS-137 6.95E+09 9.24E+09 3.22E+09 0.00E+00 3.15E+09 1.22E+09 1.32E+08 BA-140 1.28E+07 1.57E+04 8.27E+05 0.00E+00 5.33E+03 1.06E&O4 1.98E+07 :

CE-141 1.60E+04 1.07E+04 1.23E+03 0.00E+00 5.03E+03 0.00E+00 3.06E+07 [

j CE-144 1.86E+06 7.68E+05 9.97E+04 0.00E400 4.59E+05 0.00E+00 4.67E+08 l !

f i

s t

i 46 ODCM Rev. 7

{

l

i l

TABLE 4-10 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK PATHWAY - CHILD RECEPTOR ,

I

?

NUCLIDES BONE LIVER T. BODY TlWROID KIDNEY LUNG GI-LLI ;

H-3 0.00E+00 2.09E+03 2.09E+03 2.09E+03 2.09E+03 2.09E+03 2.09E+03 CR-51 0.00E+00 0.00E+00 2.95E+04 1.64E+04 4.47E+03 2.99E+04 1.56E+06 ;

l MN-54 0.00E+00 9.94E+06 2.65E+06 0.00E+00 2.79E+06 0.00E+00 8.34E+06 FE-59 3.92E+07 6.35E+07 3.16E+07 0.00E+00 0.00E+00 1.84E+07 6.61E+07 !

CO-58 0.00E+00 4.48E+06 1.37E+07 0.00E+00 0.00E+00 0.00E+00 2.61E+07 CO-60 0.00E+00 2.21E+07 6.52E+07 0.00E+00 0.00E+00 0.00E+00 1.23E+08 ZN 65 1.91E+09 5.09E+09 3.17E+09 0.00E+00 3.21E+09 0.00E+00 8.95E+08 l SR-89 2.23E+09 0.00E+00 ' 6.38E+07 0.00E+00 0.00E+00 0.00E+00 8.65E+07 >

SR-90 5.80E+10 0.00E+00 1.47E+10 0.00E+00 0.00E+00 0.00E+00 7.81E+08 ZR-95 1.38E+03 3.03E+02 2.70E+02 0.00E+00 4.34E+02 0.00E+00 3.16E+05 SB-124 3.84E+07 4.99E+05 1.35E+07 8.49E+04 0.00E+00 2.13E+07 2.41E+08 l 3.42E+08 3.44E+08 1.96E+08 1.14E+11 5.65E+08 0.00E+00 3.06E+07 I-131 I-133 4.51E+06 5.57E+06 2.11E+06 1.04E+09 9.29E+06 0.00E+00 2.25E+06 i CS-134 '1.13E+10 1.86E+10 3. 92 E+ 09 0.00E+00 5.76E+09 2.07E+09 1.00E+08 CS-137 1.67E+10 1.60E+10 2.36E+09 0.00E+00 5.22E+09 1.88E+09 1.00E+08 BA-140 3.10E+07 2.71E+04 1.81E+06 0.00E+00 8.83E+03 1.62E+04 1.57E+07 i CE-141 3.94E+04 1.97E+04 2.92E+03 0.00E+00 8.62E+03 0.00E+00 2.45E+07 .

l CE-144 4.57E+06 1.43E+06 2.44E+05 0.00E+00 7.94E+05 0.00E+00 3. 74E+08

( l l

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l 47 ODCM Rev. 7 l

l l

1 l

{

l TABLE 4-11 Ri DOSE CONVERSION FACTORS FOR THE GRASS-COW-MILK' PATHWAY - INFANT RECEPTOR !

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI l H-3 0.00E+00 3.18E+03 3.18E+03 3.18E+03 3.18E+03 3.18E+03 3.18E+03 CR-51 0.00E+00 0.00E+00 4.67E+04 3.05E+04 6.66E+03 5.93E+04 1.36E+06 MN-54 0.00E+00 1.85E+07 4.19E+06 0. 00E+ 00 4.10E+06 0.00E+00 6.79E+06 ;

FE-59 7.32E+07 1.28E+08 5.04E+07 0.00E400 0.00E+00 3.78E+07 6.11E+07 !

l CO-58 0.00E,00 8.96E+06 2.23E+07 0.00E+00 0.00E+00 0.00E+00 2.23E+07 !

CO-60 0.00E+00 4.52E+07 1.07E+08 0.00E+00 0.00E+00 0.00E+00 1.07E+08 e ZN-65 2.57E+09 8.81E+09 4.06E+09 0.00E+00 4.27E+09 0.00E+00 7.44E+09 .

.SR-89 4.25E+09 0.00E+ 00 1.22E+08 0.00E+00 0.00E+00 0.00E+00 8.74E+07 SR-90 6.31E+10 0.00E+00 1.61E+10 0.00E+00 0.00E+00 0.00E+00 7.88E+08 ;

ZR-95 2.45E+03 5.97E+02 4.23E+02 0.00E+00 6.43E+02 0.00E+00 2.97E+05 '

SB-124 7.41E+07 1.09E+06 2.30E+07 1.97E405 0.00E+00 4.64E+07 2.29E+08 I-131 7.14E+08 8.42E+08 3.70E+08 2.77E+11 9.83E+98 0.00E+00 3.00E+07 l I-133 9.52E+06 1.39E+07 4.06E+06 2.52E+09 1.63E+07 0.00E+00 2.35E+06 l l CS-134 1.82E+10 3.40E+10 3.44E+09 0.00E+00 8.76E+09 3.59E+09 9.24E+07 CS-137 2.67E+10 3.13E+10 2.22E+09 0.00E+00 8.39E+09 3.40E+09 9.78E+07 BA-140 6.37E+07 6.37E+04 3.28E+06 0.00E+00 1.51E+04 3.91E+04 1.57E+07 CE-141 7.81E+04 4.77E+04 5.61E+03 0.00E400 1.47E+04 0.00E+00 2.46E+07 CE-144 6.55E+06 2.68E+06 3.67E+05 0.00E+00 1.08E+06 0.00E+00 3.76E+08 t

i 48 ODCM Rev. 7 l

l ., , . - . ..

t TABLE 4-12 i

Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATINAY - ADULT RECEPTOR NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 CR-51 0.00E+00 0.00E+00 1.00E+02 5.95E+01 2.28E+01' 1.44E+04 3.32E+03 MN-54 0.00E+00 3.96E+04 6.30E+03 0.00E+00 9.84E+03 1.40E+06 7.74E+04 FE-59 1.18E+04 2.78E+04 1.06E+04 0.00E+00 0.00E+00 1.02E+06 1.88E+05 CO-5S 0.00E+00 1.58E+03 2.07E+03 0.00E+00 0.00E+00 9.28E+05 1.06E+05 CO-60 0.00E+00 1.15E+04 1.48E+04 0.00E+00 0.00E+00 5.97E+06 2.85E+05 ZN-65 3.24E+04 1.03E+05 4.66E+04 0.00E+00 6.90E+04 8.64E+05 5.34E+04 '

SR-89 3.04E+05 0.00E+00 8.72E+03 0.00E+00 0.00E+00 1.40E406 3.50E+05

SR-90 9.92E+07 0.00E+00 6.10E+06 0. 00E+ 00 0.00E+00 9.60E+06 7.22E+05 I ZR-95 1.07E+05 3.44E+04 2.33E+04 0.00E+00 5.42E+04 1.77E+06 1.50E+05 -

l SB-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 0.00E+00 2.48E+06 4.06E+05

1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 0.00E+00 6.28E+03 l I-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 0.00E+00 8.88E+03

! CS-134 3.73E+05 8.48E+05 7.28E+05 0.00E+00 2.87E+05 9.76E+04 1.04E+04 CS-137 4.78E+05 6.21E+05 4.28E+05 0.00E+00 2.22E+05 7.52E+04 8.40E+03 ;

BA-140 3.90E+04 4.90E+01 2.57E+03 0.00E+00 1.67E+01 1.27E+06 2.18E+05 CE-141 1.99E+04 1.35E+04 1.53E+03 0.00E+00 6.26E+03 3.62E+05 1.20E+05 -

CE-144 3.43E+06 1.43E+06 1.84E+05 0.00E+00 8.48E+05 7.78E+06 8.16E+05 l

1 i

)

49 ODCM Rev. 7 I

l I

.i TABLE 4-13

{ Ri DOSE CONVERSION FACTORS FOR THE INHA1ATION PATHWAY - TEEN RECEPTOR i

BJCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI ;

H-3 0.00E+00 1.27E+03 1.27E+03 1.27E+03 1.27E+03 1.27E+03 1.27E+03 CR-51 0.00E+00 0.00E+00 1.35E+02 7.50E+01 3.07E+01 2.10E404 3.00E+03 MN-54 0.00E+00 5.11E+04 8.40E+03 0.00E+00 1.27E+04 1.98E+06 6.68E+04 FE-59 1.59E+04 3.70E+04 1.43E+04 0.00E+00 0.00E+00 1.53E+06 1.78E+05 CO-58 0.00E+00 2.07E+03 2.78E+03 0.00E+00 0.00E+00 1.34E+06 9.52E+04 CO-60 0.00E+00 1.51E+04 1.98E+04 0.00E+ 00 0.00E+00 8.72E+06 2.59E+05 l ZN-65 3.86E+04 1.34E+05 6.24E+04 0.00E+00 8.64E+04 1.24E+06 4.66E+04 l

SR-89 4.34E+05 0.00E+00 1.25E+04 0.00E+00 0.00E+00 2.42E+06 3.71E+05 !

SR-90 1.08E+08 0.00E+00 6.68E+06 0.00E+00 0.00E+00 1.65E+07 7.65E+05 ZR-95 1.46E+05 4.58E+04 3.15E404 0.00E+0C 6.74E+04 2.69E+06 1.49E+05 SB-124 4.30E+04 7.94E+02 1.68E+04 9.76E+01 0.00E+00 3.85E+06 3.98E+05

I-131 3.54E+04 4.91E+04 2.64E+04 1.46E+07 8.40E+04 0.00E+00 6.49E+03 i 1-133 1.22E+04 2.05E+04 6.22E+03 2.92E+06 3.59E+04 0.00E+00 1.03E+04 CS-134 5.02E+05 1.13E+06 5.49E+05 0.00E+00 3.75E+05 1.46E+05 9.76E+03 i CS-137 6.70E+05 8.48E+05 3.11E+05 0.00E+00 3.04E+05 1.21E+05 8.48E+03 BA-140 5.47E+04 6.70E+01 3.52E403 0.00E+00 2.28E+01 2.03E+06 2.29E+05 CE-141 2.84E+04 1.90E+04 2.17E+03 0.00E+00 8.88E+03 6.14E+05 1.26E+05 !

CE-144 4.89E+06 2.02E+06 2.62E+05 0.00E400 1.21E+06 1.34E+07 8.64E+05 t

l l

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4 50 ODCM Rev. 7

I l

TABLE 4-14 Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - CHILD RECEPTOR i

5 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 CR-51 0.00E+00 0.00E+00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 MN-54 0.00E+00 4.29E+04 9.51E+03 0.00E+00 1. 00E+ 04 1.58E+06 2.29E+04 FE-59 2.07E+04 3.34E+04 1.67E+04 0.00E+00 0.00E+00 1.27E+06 7.07E+04 '

CO-58 0.00E+00 1.77E+03 3.16E+03 0.00E+00 0.00E+00 1.11E+06 3.44E+04 ,

CO-60 0.00E+00 1.31E+04 2.26E+04 0.00E+00 0.00E+00 7.07E+06 9.62E+04 ZN-65 4.26E+04 1.13E+05 7.03E+04 0.00E+00 7.14E+04 9.95E+05 1.63E+04 SR-89 5.99E+05 0.00E+00 1.72E+04 0.00E+00 0.00E+00 2.16E+06 1.67E+05 SR-90 1.01E+08 0.00E+00 6.44E+06 0.00E+00 0.00E+00 1.48E+07 3.43E+05 ZR-95 1.90E+05 4.18E+04 3.70E+04 0.00E+00 .5.96E+04 2.23E+06 6.11E+04 SB-124 5.74E+04 7.40E+02 2.00E+04 1.26E+02 0.00E+00 3.24E+06 1.64E+05 I-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00E+00 2.84E+03 !

I-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.00E+00 5.48E+03 CS-134 6.51E+05 1.01E+06 2.25E^05 0.00E+00 3.30E+05 1.21E+05 3.85E+03 CS-137 9.07E+05 8.25E+05 1.28E+05 0.00E+00 2.82E+05 1.04E+05 3.62E+03 ;

BA-140 7.40E+04 6.48E+01 4.33E+03 0.00E+00 2.11E+01 1.74E+06 1.02E+05 CE-141 3.92E+04 1.95E+04 2.90E+03 0.00E+00 8.55E+03 5.44E+05 5.66E+04 CE-144 6.77E+06 2.12E+06 3.61E+05 0.00E+00 1.17E+06 1.20E+07 3.89E+05 '

a 1

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l 51 ODCM Rev. 7 l l

j <

P l

l TABLE 4-15 l

Ri DOSE CONVERSION FACTORS FOR THE INHALATION PATHWAY - INFANT RECEPTOR NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 6.47E+02 6.47E+02 6.47E+02 6.47E+02 6.47E+02 6.47E+02 CR-51 0.00E+00 0.00E400 8.95E+01 5.75E+01 1.32E+01 1.28E+04 3.57E+02 MN-54 0.00E+00 2.53E404 4.98E+03 0.00E+00 4.98E+03 1.00E+06 7.06E+03 FE-59 1.36E+04 2.35E+04 9.48E+03 0. 00E+ 00 0,00E+00 1.02E+06 2.48E+04 j

~

CO-58 0.00E+00 1.22E+03 1.82E+03 0.00E400 0.00E+00 7.77E+05 1.11E+04 CO-60 0.00E+00 8.02E+03 1.18E+04 0.00E+00 0.00E+00 4.51E+06 3.19E+04 IN-65 1.93E+04 6.26E+04 3.11E+04 0.00E+00 3.25E404 6.47E+05 5.14E+04 SR-89 3.98E+05 0.00E+00 1.14E+04 0.00E+00 0.00E+00 2.03E+06 6.40E+04 l SR-90 4.09E+07 0.00E+00 2.59E+06 0.00E+00 0.00E+00 1.12E+07 1.31E+05 ZR-95 1.15E+05 2.79E+04 2.03E+04 0.00E+00 3.11E+04 1.75E+06 2.17E+04 SB-124 3.79E+04 5.56E+02 1.20E+04 1.01E+02 0.00E+00 2.65E+06 5.91E+04

I-131 3.79E+04 4.44E+04 1.96E+04 1.48E+07 5.18E+04 0.00E+00 1.06E+03 I-133 1.32E+04 1.92E+04 5.60E+03 3.56E+06 2.24E+04 0.00E+00 2.16E+03 ,

, CS-134 3.96E+05 7.03E+05 7.45E+04 0.00E+00 1.90E+05 7.97E+04 1.33E403 1 l

CS-137 5.49E+05 6.12E+05 4.55E+04 0.00E+00 1.72E+05 7.13E+04 1.33E+03 BA-140 5.60E+04 5.60E+01 2.90E+03 0.00E+00 1.34E+01 1.60E+06 3.84E+04 CE-141 2.77E+04 1.67E+04 1.99E+03 0.00E+00 5.25E+03 5.17E+05 2.16E+04 :

CE-144 3.19E+06 1.21E+06 1.76E+05 0.00E+00 5.38E+05 9.84E+06 1.48E+05 i

t i

52 ODCM Rev. 7

TABLE 4-16 l

PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR IDNG TERM RELEASES AT Tile NEAREST PATilWAY LOCATIONS CENTERED ON UNIT 1 RESIDENCE (b) CARDEN(b) MILK (b)

DIRECTION X/Q Dist. D/Q X/Q Dist. D/Q X/Q Dist. D/Q (Sec/m ) 3 Miles (m-2) 3 (Sec/m ) Miles (m-2) 3 (Sec/m ) Miles (m-2)

N 2.92E-06 1.4 3.25E-09 2.92E-06 1.4 3.25E 09 7.03E-07 (a) 3.48E-10 NNE 1.81E-06 1.8 2.88E-09 4.70E-07 (a) 4.04E-10 4.70E-07 (a) 4.04E-10 NE 1.95E-06 1.9 3.85E-09 1.76E-06 2.1 3.29E-09 5.77E-07 (a) 6.51E-10 ENE 1.03E-06 2.7 1.08E-09 1.03E-06 2.7 1.08E-09 3.86E-07 (a) 2.86E-10 E 9.39E-07 2.8 6.68E-10 3.71E-07 (a) 1.87E-10 3.71E-07 (a) 1.87E-10 ESE 6.37E-07 3.7 2.84E-10 4.12E-07 4.6 1.60E-10 4.12E-07 4.6 1.60E-10 goat SE 8.83E-07 4.1 2.61E-10 8.83E-07 4.1 2.61E-10 5.84E-07 (a) 1.52E-10 SSE 1.27E-06 4.7 2.61E-10 1.09E-06 (a) 2.15E-10 1.09E-06 (a) 2.15E-10 S 2.58E-06 4.6 4.85E-10 2.09E-06 5.2 3.59E-10 2.13E-06 5.1 3.71E-10 cow SSW 3.26E-06 3.5 8.26E-10 2.28E-06 (a) 4.53E-10 2.28E-06 (a) 4.53E-10 SW 2.80E-06 2.9 9.10E-10 1.58E-06 (a) 3.56E-10 1.58E-06 (a) 3.56E-10 WSW 1.95E-06 2.6 1.09E-09 8.55E-07 (a) 3.18E-10 8.55E-07 (a) 3.18E-10 W 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 VNW 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 NW 8.24E-07 3.8 5.25E-10 7.55E-07 4.1 4.61E-10 6.02E-07 (a) 3.27E-10 NNW 1.46E-06 2.0 1.47E-09 5.20E-07 (a) 3.04E-10 5.20E-07 (a) 3.04E-10 (a) 5-mile value used since there is no pathway located within the sector up to five miles.

(b) Controlling locations are discussed in Appendix A.

References:

1984 Land Use Census (letter ANPM-21221-JRM/LEB). NUS Corporation letters NUS ANPP-1385 and NUS-ANPP-1386.

53 ODCM Rev. 7

TABLE 4-16 (Continued)

PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT Tile NEAREST PATHWAY LOCATIONS CENTERED ON UNIT 2 RESIDENCE (b) GARDEN (b) MIIX(b)

DIRECTION X/Q Dist. D/Q X/Q Dist. D/Q X/Q Dist. D/Q 3

(Sec/m ) Miles (m-2) (Sec/m )

8 Miles (m-2) (Sec/m') Miles (m-2)

N 2.73E-06 1.5 2.92E-09 2.39E-06 1.7 2.35E-09 7.03E-07 (a) 3.48E-10 NNE 2.20E-06 1.5 3.87E-09 2.20E-06 1.5 3.87E-09 4.70E-07 (a) 4.04E-10 NE 1.85E-06 2.0 3.55E-09 1.57E-06 2.3 2.78E-09 5.77E-07 (a) 6.51E-10 ENE 1.03E-06 2.7 1.08E-09 1.03E-06 2.7 1.08E-09 3.86E-07 (a) 2.86E-10 E 8.80E-07 3.0 6.06E-10 3.71E-07 (a) 1.87E-10 3.71E-07 (a) 1.87E-10 ESE 6.25E 07 3.7 2.76E-10 3.96E-07 4.7 1.51E-10 3.96E-07 4.7 1.51E-10 goat SE 9.06E-07 4.0 2.72E-10 9.06E-07 4.0 2.72E-10 5.84E-07 (a) 1.52E-10 SSE 1.34E-06 4.5 2.81E-10 1.09E-06 (a) 2.15E-10 1.09E-06 (a) 2.15E-10 i S 2.63E-06 4.5 5.01E-10 2.19E-06 5.0 3.88E-10 2.19E-06 5.0 3.88E-10 cow SSW 3.48E-06 3.2 9.19E-10 2.28E-06 (a) 4.53E-10 2.28E-06 (a) 4.53E-10 SW 2.93E-06 2.7 9.75E-10 1.58E-06 (a) 3.56E-10 1.58E-06 (a) 3.56E-10 WSW 2.01E-06 2.5 1.16E-09 8.55E-07 (a) 3.18E-10 8.55E-07 (a) 3.18E-10 W 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 WNW 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 NW 7.84E-07 4.0 4.88E-10 7.84E-07 4.0 4.88E-10 6.02E-07 (a) 3.27E-10 NNW 1.46E-06 2.0 1.47E-09 5.20E-07 5.0 3.04E-10 5.20E-07 (a) 3.04E-10 (a) 5-mile value used since there is no pathway located within the sector up to five miles.

(b) Controlling locations are discussed in Appendix A.

References:

1984 Land Use Census (letter ANPM-21221-JRM/LEB). NUS Corporation letters NUS-ANPP-1385 and NUS-ANPP-1386.

L 54 ODCM Rev. 7

TABLE 4-16 (Continued)

PALO VERDE NUCLEAR GENERATING STATION DISPERSION AND DEPOSITION PARAMETERS FOR LONG TERM RELEASES AT THE NEAREST PATINAY LOCATIONS CENTERED ON UNIT 3 RESIDENCE (b) CARDEN(b) MIIX(b)

DIRECTION X/Q Dist. D/Q X/Q Dist. D/Q X/Q Dist. D/Q (Sec/m3 ) Miles (m-z) (Sec/m3 ) Miles (m-2) (Sec/m') Miles (m-2)

N 2.58E-06 1.8 2.47E-09 2.42E-06 1.9 2.22E-09 7.03E-07 (a) 3.48E-10 NNE 1.85E-06 1.7 2.97E-09 1.85E-06 1.7 2.97E-09 4.70E-07 (a) 4.04E-10 NE 1.66E-06 2.2 3.00E-09 1.48E-06 2.4 2.54E-09 5.77E-07 (a) 6.51E-10 ENE 8.75E-07 2.9 8.86E-10 8.75E-07 2.9 8.86E-10 3.86E-07 (a) 2.86E-10 E 8.90E-07 3.0 6.17E-10 4.06E 07 4.6 2.15E-10 4.25E-07 4.5 2.31E-10 goat ESE 6.37E-07 3.7 2.84E-10 5.80E-07 4.0 2.46E-10 3.73E-07 (a) 1.37E-10 SE 5.84E-07 (a) 1.52E-10 5.84E-07 (a) 1.52E-10 5.84E-07 (a) 1.52E-10 SSE 1.36E-06 4.4 2.88E-10 1.09E-06 (a) 2.15E-10 1.09E-06 (a) 2.15E-10 S 2.65E-06 4.2 5.25E-10 2.25E-06 4.9 4.06E-10 2.31E-06 4.8 4.21E-10 cow SSW 3.64E-06 3.1 9.82E-10 2.28E-06 (a) 4.53E-10 2.28E-06 (a) 4.53E-10 SW 3.19E-06 2.5 1.11E-09 1.58E-06 (a) 3.56E-10 1.58E-06 (a) 3.56E-10 WSW 2.12E-06 2.4 1.26E-09 8.55E-07 (a) 3.18E-10 8.55E-07 (a) 3.18E-10 W 7.54E-07 (a) 4.44E-10 7.54E-07 (a) 4.44E-10 7.54E-10 (a) 4.44E 10 WNW 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 6.03E-07 (a) 3.25E-10 IN 6.83E-07 4.3 4.05E-10 6.82E-07 4.3 4.05E-10 6.02E-07 (a) 3.27E-10 NNW 1.34E-06 2.2 1.26E-09 5.16E-07 5.0 3.01E-10 5.20E-07 (a) 3.04E-10 (a) 5-mile value used since there is no pathway located within the sector up to five miles.

(b) Controlling locations are discussed in Appendix A.

References:

1984 Land Use Census (letter ANPM-21221-JRM/LEB). NUS Corporation letters NUS-ANPP-1385 and NUS-ANPP-1386.

55 ODCM Rev. 7

i

4.4 Requirements

IJguld Emuents

Ite dose or dose commitment to a MEMBER OFTHE PUBLIC from radioactive materials in liquid emuents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY ( See Figure 6-4 ) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and

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

l Applicability: At all times.

! Actiom With the calculated dose from the release of radioactive materials in liquid emuents exceeding any

of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical l

Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

l 4.4.1 Surveillance Requirements:

Cumulative dose contributions from liquid emuents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

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i 4.4.2 Implementation of the Requirements: l This Requirement does not require implementation guidance. There are no offsite liquid j emuent releases.

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56 ODCM Rev. 7 t

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5.0 TOTAL DOSE AND DOSE TO PUBLIC ONSITE

5.1 Requirement

Total Dose The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to i releases of radioactisity and to radiation from uranium fuel cycle sources shall be limited to less than ;

i or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

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j Applicability: At all times.

Action:

With the calculated doses from the release of radioactive materials in liquid and gaseous effluents exceeding twice the limits of Section 4.4a,4.4b,4.la,4.lb,4.2a or 4.2b calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Section 5.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pu suant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be takert to reduce subsequent releases to prevent recurrence of execeding the above limits and includes the schedule for achieving conformance with j the above limits. This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that !

estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes l releases (s) covered by this report. It shall also describe levels of radiation and concentrations of l radioactive materialimolved, and the cause of the exposure levels or mncentrations. If the estimated i dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part l

190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

I 5.1.1 Surveillance Requirements: ;

a. Cumulative dose contributions from the gaseous effluents shall be determined in accordance with the surveillance requirements of Section 4.4.1,4.1.1 and 4.2.1 and in accordance with the methodology and parameters contained in Section 5.1.2.

b. Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in Section 5.1.2. This requirement is applicable only under conditions set forth l

in Section 5.1, Action.

5.1.2 Implementation of the Requirement l Since all other uranium fuel cycle sources are greater than 20 miles away, only the PVNGS l site need be considered.

l The total dose to any MEMBER OFTHE PUBLIC will be determined based on a sum of the doses from all three units' releases and doses from direct radiation from PVNGS.

57 ODCM Rev. 7

This dose evaluation is performed annually and submitted with the Semiannual Radioactive Effluent Release Report for July through December to assure compliance with 40CFR Part 190, Emironmental Radiation Protection Standards for Nucicar Pcrver Operation.

NUREG-0543, Methods for Demonstrating LWR Compliance With the EPA Unmium Fuel Cycle Standard (40 CFR Psrt 190), February,1980, provides a discussion on mmpliance with 40 CFR Part 190 in relation to the Radiological Environmental Technical Specifications for sites of up to four nucicar power reactors. The NUREG concludes that as long as a nuclear plant site operates at a level below the 10 CFR Part 50, Appendix I reporting requirements, and there is no significant source of direct radiation from the site, no extra analysis is required to demonstrate compliance with 40 CFR Part 19G. As a result, this dose evaluation will also be performed whenever calculated doses associated with ellinent releases exceed twice the limits of Section 4.4a,4.4b,4.la,4.1b,4.2a or 4.2b.

Dose Contribution from Liauid and Gaseous Effluents The annual whole body dose accumulated by a MEMBER OF THE PUBLIC for the noble gases released in gaseous efiluents is determined by using the following equation:

l l Dwa = (3.17E-08) L [(K,) (X/Q)o,m. (Q,)] (5-1)

Where:

K. = the whole body dose factor due to gamma emissions for cach identified noble gas radionuclide i, in mrem,7r per pCi/m' from Table 3-3.

l Q, = the integrated release of radionuclide i,in yCi for the previous calendar

! year.

(X/Q)mrr = the highest calculated annual average dispersion parameter, in sec/m8 ,

for a particular unit, at the controlling location, from Table 4-16, or concurrent meteorological data if available.

= 2.92E-06 from Unit 1

= 2.19E-06 from Unit 2

= 2.31E-06 from Unit 3 Dws = the annual whole body dose in mrem to a MEMBER OF THE PUBLIC at the mntrolling k) cation due to noble gases released in gaseous l cffluents.

3.17E-08 = the inverse of seconds in a year (yr/sec).

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58 ODCM Rev. 7

l De annual dose to any organ accumulated by a MEMBER OF THE PUBLIC for iodine-131, J

iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 da3s :

released in gaseous effluents is determined by using the following equation: ),

D, = (3.17E-08) L [L(R.W. )(0,)] (5-2)

Where:

D, = the total annual organ dose from gaseous effluents to a MEMBER OF THE PUBLIC, in mrem, at the controlling location. ;

Q, = the integrated release of radionuclide i, in Ci, for the presions !

calendar year.

R. = the dose factor for each identified radionuclide i, for pathway k (for the inhalation pathway in mrem /yr per pCi/m' and for the food and ground plane pathways in m2 -mremfyr per Ci/sec) at the controlling location.

The R.'s for each age group are given in Tables 4-1 through 4-15.

%'g = the highest annual average dispersion or deposition parameter for the particular unit, used for estimating the total annual organ dose to a MEMBER OF THE PUBLIC at the controlling location for the particular unit.

= (X/0)uum in sec/m' for the inhalation pathway and for all tritium calculations, for organ dose at the mntrolling location, from Table 4-16 or concurrent meteorological data if availabic.

! = 2.92E-06 from Unit I !

! = 2.19E-06 from Unit 2

= 2.31E-06 from Unit 3 l = (D/0)usm in m.2, for the food and ground plane pathways, for organ dose at the controlling location, from Table 4-16 or concurrent meteorological data if available. ,

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= 3.25E-09 from Unit 1 I

= 3.88E-10 from Unit 2 l

= 4.21E-10 from Unit 3 !

3.17E-08 = the inverse of seconds in a year (31/sec).

Dose Due to Direct Radiation he component of dose to a MEMBER OF THE PUBLIC due to direct radiation will be l cvaluated by first determining the direct radiation dose at the site boundary in each sector, and l then extrapolating the site boundary dose to the controlling location by the inverse square law of distance.

l 59 ODCM Rev. 7

i k

i Dose from Radioactive Liquid and Gaseous Efiluents to MEMBERS OF THE PUBLIC due ,

to their activities within the SITE BOUNDARY ;

These activities have been determined to be limited to the vicinity of the Visitor Center located inside the SITE BOUNDARY west of Unit 1. An assumption was made that no MEMBER OF THE PUBLIC would spend more than eight hours per year at this location. ;

However this calculation has been historically performed assuming an occupancy factor of one, !

(implying continuous occupancy over the entire year). :

A X/Q, determined for the Visitor Center, will be used for this assessment. ;

Equations 5-1 and 5-2 in Section 5.1.2 should be used for this assessment. I t

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60 ODCM Rev. 7

6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

6.1 Requirements

REMP The radiological emironmental monitoring program shall be conducted as specified in Tabic 6-1.

Applicability: At all times. l Action:

a. With the radiological environmental monitoring program not being mnducted as specified in ,

Table 6-1, prepare and submit to the Commission, in the Annual Radiological Environmental l Operating Report, as required by Section 7.2, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. ;

h. With the level of radioactivity as the result of plant efuuents in an emironmental sampling medium at a specified location exceeding the reporting levels of Table 6-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose

to A MEMBER OF THE PUBLIC is less than the calendar year limits of Section 4.4,4.1 and 4.2 When more than one of the radionuclides in Table 6-2 are detected in the sampling medium, this report shall be submitted if:

l j concentration (1) concentration (2) i + + . . 2 1.0

( reporting level (1) reporting level (2)

When radionuclides other than those in Table 6-2 are detected and are the result of plant i effluents, this report shall be submitted if the potential annual dose

to a MEMBER OF THE ;

PUBLIC is equal to or greater than the calendar year limits of Section 4.4,4.1 and 4.2. This i report is not required if the measured level of radioactivity was not the result of plant ef0uents; however, in such an event, the condition shall be reported and described in the Annual Radiological Emironmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations j required by Table 6-1, identify locations for obtaining replacement sampics and add them to the Radiological Emironmental Monitoring Program within 30 days. The specific locations from ;

which samples were unavailable may then be deleted from the monitoring program. Pursuant to Section 7.1, Semiannual Radioactive Effluent Release Report, identify the cause of the !

unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF ;

THE PUBLIC shall be indicated in this report.

61 ODCM Rev. 7

l 6.1.1 Surveillance Requirements:

l

a. The radiological emironmental monitoring samples shall be collected pursuant to Table 6-1 from the specific k> cations given in Table 6-4 and Figures 6-1,6 2, and 6-3, and shall be analyzed pursuant to the requirements of Table 6-1, and the descetion capabilitics required by Table 6-3.

6.1.2 Implementation of the Requirements: REMP The results of the radiological emironmental monitoring program are intended to supplement the results of the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected based on the effluent measurements and modeling of the emironmental exposure pathways. l Thus the specified emironmental monitoring program provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to [

the highest potential radiation exposures to individuals resulting from station ope ation.

l This requirement is implemented by Nuclear Administrative and Technical Manual procedures.

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62 ODCM Rev. 7 i

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TABLE 6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Type and Exposure Pathway Number of Representative Sampling and Frequency of ,

and/or Sample Samples and Sample locations

Collection Frequency

Analysis

i Airimrne Samples from 5 kotions: 3 Continuous sampling Gross beta ;

j samples at or near the SITE collected weekly, or weekly *, I-131 l Radiciodine and BOUNDARIES (#14A.15,21) more frequently if weekly; gamma l isotopic analysis l particulates in different sectors of the required by dust <

i highest calculated annual loading. of composite (by !

average ground level D/Q.

location) l quarterly.

I sample (#40) from areas of +

special interest, which is from the vicinity of a community having the highest calculated annual average D/Q.

I sample (#6) from a mntrol k> cation 15-30 km (10-20 mi) distant and in the least prevalent wind direction.

6 l Direct radiation 41 stations (#6-45, #50) with Quarterly Gamma dose two or more dosimeters for quarterly.

measuring dose rate mntinuously, placed as follows:

an inner ring of stations at the site boundary and an oute-r ring in the 4-to-5 mi range from the site with a station in each sector of each ring (16 sectors X 2 l rings = 32 stations). 7 ,

additional stations are at local !

! schools and/or population centers; 2 other stations are used as controls.

D/O refers to average annual relative ground deposition rate.

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63 ODCM Rev. 7 l

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1 TABLE 6-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i Exposure Pathway Number of Representative Sampling and Type and and/or Sample Samples and Sample locations

Collection Frequency

Frequency of Analysis d Waterhorne Surface Water storage reservoir (#60) Monthly composite of Gamma isotopic Evaporation pond #1 (#59) weekly grab sample. analysis monthly; l l Evaporation pond #2 (#63) tritium quarterly. l

! Ground 2 onsite wells ' (#57, #58) Quarterly grab sample Tritium and [

t gamma isotopic analysis quarterly.

Drinking (well) 3 wells from surrounding Composite sample of I-131 analysis on residenas (#46, #48, #49) that weekly grab samples each composite would be affected by its over 2-week period when the dose l discharge. when I.131 analysis is calculated for the performed, monthly consumption of composite of weekly the water is greater grab samples otherwise than I mrem per year.8 Composite for gross beta and l gamma isotopic l l analyses monthly.

Composite for tritium analysis quarterly.

Incestion Samples from milking animals Semimonthly for Gamma isotopic :

in 3 locations within 5 km animals on pasture; and I-131 analysis Milk distance having the highest dose otherwise, monthly. semimonthly when :

potential. If there are none, I animals are on sample from milking animals in pasture or monthly each of three areas (#50, #51, at other times.

53) between 5 and 8 km distant where doses are calculated to be greater than 1 mrem per year. 8 i

i 64 ODCM Rev. 7 i

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l TABLE 6-1 (Continued)

^

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM '

Exposure Pathway Namber of Sampling and Type and Frequency of and/or Sample Representative Samples Collection Frequency

Analysis d and Sample locations

Food Products

Samples (#47, #52) of Monthly during Gamma isotopic l 3 different kinds of growing season. analysis.

i l broad leaf vegetation l grown nearest each of two offsite locations of highest predicted annual average ground-level D/Q if milk r

sampling id not performed.

1 sampic (#62) of each Monthly during Gamma isotopic l of the similar broad growing season. analysis.

I leaf vegetation grown 15=30 km distant in the least prevalent wind direction if milk sampling is not performed.

When broad leaf vegetation samples are not available, reports from 4 cristing supplemental airborne radioiodine sample locations will be substituted.

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65 ODCM Rev. 7

1

\

TABLE 6-1 (Ccmtinued)

TABLE NOTATIONS a The number, media, frequency, and location of sampling may vary from site to site. It is recognized that, at times,it may not be possible or practical to obtain samples of the media of cholm at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the l particular pathway in question. Actuallocations (distance and direction) from the site shall be provided in Table 6-4 and Figures 6-1,6-2, or 6-3 in the ODCM. Refer to Regulatory Guide 4.1,' Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants.'

b Regulatory Guide 4.13 provides guidance for thermoluminescence dosimetry (TLD) systems used for l emironmental monitoring. One or more instruments, such as a pressurized ion chamber, for measuring l and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters.

l For the purposes of this table, a thermoluminescent dosimeter may be considered to be one phosphor, ;

and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges ;

I ' should not be used for measuring direct radiation.

e Particulate sample filters shall be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the indisidual samples.

d Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

l e The purpose of this sample is to obtain background information. Ifit is not practical to establish c(mtrol locations in accordance with the wind direction criteria, other sites that provide vaha l'ackground data may be substituted.

f Groundwater samples should be taken when this source is tapped for drinking or irrigation purposes in !

areas where the hydraulic gradient or recharge properties are suitable for contamination. l g The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

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i TAH116-2 l Reporting Irrels for Radioactivity Concentrations in Envimamental Samples Airborne Particulate or Fresh Milk Food Products i Analysis Water (pCill) Gas (pCi/m') (pCill) (pCi/kg, wet)

H-3 20,000

Mn-54 1.000 Fe-59 400 i Co-58 1,000 J

Co-60 300 i Zn-65 300 Zr-Nb-95 400 1-131 2" 0.9 3 100 Cs-134 30 10 60 1,000 Cs-137 50 20 70 2,000 I Ba-La-140 200 300 For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a va!ue of 30,000 pCi!! may be used.

If no drinking water pathway exists. a reporting level of 20 pCill may be used.

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1 67 ODCM Rev. 7

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TABl.E 6-3 l

Detection Capabilities for Envirtmmental Analysis

i imer Ilmit of Detection (Ill))

Airborne Particulate or Fresh Milk Food Prtxtucts Analysis Water (pCi/l) Gas (pCi/m') (pCi/1) (pCi/kg, wet)

Gross Beta 4 0.01 H-3 2000

Mn-54 15 Fe-59 30 ,

Co-58, -60 15 Zn-65 30 Zr-95 30 l Nb-95 15 I-131 1" 0.07 1 60 Cs-134 15 0.05 15 60 Cs-137 18 0.06 18 80 Ba-140 60 60 La-140 15 15 NOTE: "Diis list does not mean that only these nuclides are te te detected and reported. Other peaks that are measurat?e and identifiable, together with the above nuclides, shall also be identified and reported.

l If no drinking water pathway exists, a value of 3(XX) pCill may be used.

l If no drinking water pathway exists, a vahe of 15 pCill may be used.

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68 ODCM Rev. 7 I

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TABLE 6-3 (Continued)

TABLE NOTATION a Guidance for detection capabilities for thermoluminescent dosimeters used for emironmental measurements is given in Regulatory Guide 4.13.

b Table 6-3 indicates acreptable detection capabilities for radioactive materials in environmental samples. These detection capabilities are tabulated in terms of the lower limits of detection (LLDs).

'lhe LLD is defined, for purposes of this guide, as the smallest mncentration of radioactive material in a sample that will yield a net count (abcwe system background) that will be detected with 95% ,

probability with only 5% probability of falsely concluding that a blank observation represents a "real

signal.

For a particular measurement system (which may include radiochemical separation):

4.66 s,,

LLD =

E

V ' 2.22

Y ' exp ( -1 At )

Where:

LLD is the a priori lower limit of detection as defined above (as pCi per unit mass or volume),

s, is the standard deviation of the background counting rate or of the munting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield (when applicabic),

A is the radioactive decay mnstant for the particular radionuclide, and at for emironmental samples is the clapsed time between sample collection (or end of the sample collection period) and time of munting.

In calculating the LLD for a radionuclide determined by gamma-ray spectrometry the background should include the typical contributions of other radionuclides normally present in the samples (e.g., ,

potassium-40 in milk samples). Typical values of E, V, Y, and At should be used in the calculation. l It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual l Radiological Emironmental Operating Report.

l 69 ODCM Rev. 7

6.2 Requirement

land Use Census ,

A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the kication in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and 2

the nearest garden

of greater than 50 m (500 ft2 ) pmducing broad leaf vegetation.

Applicability: At all tiraes.

i Actiom

a. With a land use :ensus identifying a location (s) that yields a calculated dose or dose mmmitment greater than the values currently being calculated in Section 4.2.1, identify the new location (s) in the next Semian.aual Radioactive Effluent Release Report, pursuant to Section 7.1.

b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are cuni 'y '

being obtained in accordance with Section 6.1, add the new k) cation (s) to the radiologico.

emironmental monitoring program within 30 days. The sampling kication(s), excluding the l control station location, having the lowest calculated dose or dose commitment (s), sia the same exposure pathway, may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted. Pursuant to Section 7.1, identify the new location (s) '

in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

6.2.1 Surveillance Requirements:

i

a. The land use census shall be mnducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door.to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land ,

use census shall be included in the Annual Radiological Emironmental Operating Report !

pursuant to Section 7.2. i Broad Iraf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampl!ng in Table 6-1 shall be followed, including analysis of cxmtrol samples.

6.2.2 Implementation of the Requirements:

l The above Requiremont is implemented by Nuclear Administrative and Technical Manual procedures.

I I i l

l 70 ODCM Rev. 7 l

l

i

6.3 Requirements

Intertaboratory Comparison Program Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission that correspond to samples required by Table 6-1. ,

Applicability: At all times. l Actiom

a. With analpes not being performed as required above, report the corrective actions taken to i

prevent a recurrence to the Commission in the Annual Radiological Emironmental Operating Report pursuant to Section 7.2.

6.3.1 Surveillance Requirements:

l a. A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the methodology and parameters in this !

j manual shall be included in the Annual Radiological Environmental Operating Report pursuant to Section 7.2.

I l

6.3.2 Implementation of the Requirements:

PVNGS laboratories or contract laboratories which perform analyses for the Radiological ,

t i Emironmental Monitoring Program (REMP) participate in the Emironmental Protection l Agency (EPA) Emironmental Radioactivity Laboratory Intercomparison Studies (crosscheck) l Program. The participation includes all of the determinations (sample medium-radionuclide l combinations) that are offered by the EPA and that are also included in the monitoring l program. ;

1 The sample handling preparation and analysis procedures approved for use on routine REMP '

samples, at the time the crosscheck samples are received from the EPA, are used to implement the program. The results of the crosscheck sample analyses are reviewed, at minimum on an l annual basis, to ensure that the control limits established by the EPA are not exceeded. ,

i If deviation from these specified limits is identified an investigation is made to determine the reason for the deviation and corrective actions are taken as necessary. The results of all ,

analyses made under this program are included in the Annual Radiological Emironmental Operating Report.

l l

[ l i

l 71 ODCM Rev. 7 l l

t l

i TABLE 6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS i l

l LOCATION ,

SAMPLE SAMPLE NOTE (d) DESIGNATION LOCATION DESCRIPTION (c) l SITE TYPE (a) )

l 1 TLD E30 APS Western Division Office, Goodyear ;

1 Air E30 Same as TLD (E of RR tracks) j 1

2 TLD ENE24 Smit.Libby Sc:nool, Perryville and Thomas Rds.

3 TLD E21 Liberty School,19800 W. Hwy 85 4 TLD E16 APS Buckeye Office,615 N. 4th St., Buckeye 4 Air E16 Same as TLD 5 TLD ESE11 Palo Verde School, Palo Verde Rd. (291st Ave.) and l l Old US 80 l 6 TLD (b) SP SSE31 APS Gila Bend substation, frontage road W of town 6 Air (b) Control SSE31 Same as TLD

! 7 TLD (b) SP SE7 Old US 80 and Arlington School Rd.

l 7A Air SE8 Arlington School,16351 S. Arlington School Rd.

8 TLD (b) OR SSE5 Southern Pacific Pipeline Rd.,1.4 miles SW of 355th Ave.

9 TLD (b) OR SS Southern Pacific Pipeline Rd.,2.5 miles SW of 355th Ave.

10 TLD (b) OR SE5 SE mrner of 355th Ave. and Elliot Rd.

11 TLD (b) OR ESE5 NW corner of 339th Ave. and Dobbins Rd.

12 TLD (b) OR E5 NE mrner of 339th Ave. and Buckeye.Salome Rd.

13 TLD (b) IR N1 N site boundary 14 TLD (b) IR NNE2 NNE site boundary 14A Air (b) NNE2 SW corner of 371st Ave. and Buckeye-Salome Rd.

15 TLD (b) IR NE2 NE site boundary, WRF access road 15 Air (b) NE2 Same as TLD 16 TLD (b) 1R ENE2 ENE site boundary 17 TLD (b) IR E2 E site boundary l

17A Air E4 351st Ave., I mile S of Buckeye-Salome Rd. I 72 ODCM Rev. 7 l i

i l

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

l l

l t

TABLE 6-4 l RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS LOCATION i SAMPLE SAMPLE NOTE (d) DESIGNATION LOCATION DESCRIPTION (c) ;

SITE TYPE (a) ;

18 TLD (b) IR ESE2 ESE site boundary f

19 TLD (b) IR SE2 SE site boundary l

20 TLD (b) 1R SSE2 SSE site boundary i 21 TLD (b) IR S3 S site boundary l l

21 Air (b) S3 Same as TLD !

22 TLD (b) 1R SSW3 SSW site boundary i

23 TLD (b) OR W5 2 miles N of Elliot Rd.,3 miles W of Wintersburg Rd. l t

l 24 TLD (b) OR SW4 Elliot Rd.,2 rniles W of Wintersburg Rd.

25 TLD (b) OR WSW5 Elliot Rd.,3 miles W of Wintersburg Rd. at cattleguard 26 TLD (b) OR SSW5 Shepard farm,13202 S. 383rd Ave.,0.5 miles W of l house ,

27 TLD (b) IR SW1 SW site boundary l 28 TLD (b) 1R WSW1 WSW site boundary !

29 TLD (b) IR W1 W site boundary 29 Air (b) W1 Same as TLD :

1 30 TLD (b) IR WNW1 WNW site boundary i 31 TLD (b) IR NW1 NW site boundary l

! 32 TLD (b) IR NNW1 NNW site boundary 33 TLD (b) OR NW4 Buckeye Rd.,0.5 miles W of 395th Ave.

34 TLD (b) OR NNW5 SE corner of 395th Ave. and Van Buren St.

l 35 TLD (b) SP NNW8 Fire Station,40901 W. Osborn Rd., Tonopah l 35 Air NNW8 Same as TLD 36 TLD (b) OR N5 SW corner of Winten, burg Rd. and Van Buren St.

37 TLD (b) OR NNES SE corner of 363rd Ave. and Van Buren St.

38 TLD (b) OR NES SW corner of 355th Ave. and Buckeye Rd.

73 ODCM Rev. 7 l

f TABLE 6-4

( RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLECTION LOCATIONS i

l LOCATION SAMPLE SAMPLE NOTE (d) DESIGNATION LOCATION DESCRllTION (c)

SITE TYPE (a) 39 TLD (b) OR ENE5 343rd Ave.,0.5 miles S of lower Buckeye Rd.

l 40 TLD (b) SP N3 Wintersburg. Transmission Rd. S of trailer park 40 Air (b) N3 Same as TLD I

41 TLD (b) SP WNW 20 Harquahala Valley School, Van Buren St., I mile W of ;

Steve Martori Dr. f 42 TLD (b) SP N8 Ruth Fisher School, Indian School and Wintersburg Rds. l 43 TLD (b) SP N45 Vulture Peak School, I mile S of US 60, Wickenburg ?

t 44 11.D(b) Control ENE35 APS El Mirage Office,12313 W. Grand Ave.

45 TLD (b) Transit E16 APS Buckeye Office,615 N. 4th St., REMP trailer ,

l Control (lead pig) 46 TLD ENE30 Litchfield Park School,13825 W. Indian School Rd. ;

j 46 Water (b) WD NW9 McArthur residence,41701 W. Indian School Rd.. .

Tonopah 47 TLD E35 Littleton School,115th Ave. and Hwy 85, Cashion !

47 Vegetation ENE3 Adams' residence, NW corner of 355th Ave. and l (b) Buckeye-Salome Rd.

48 TLD E24 Jackrabbit Trail, S of I-10, N of Filmore St.

l l 48 Water (b) WD SS Shepard farm,13202 S. 383rd Ave. '

49 TLD ENE11 Palo Verde Rd.,0.25 miles S of 110 49 Water (b) WD NNE2 Chowanez residence,371st Ave.,0.5 miles S of u

Buckeye-Salome Rd.

l l 50 TLD (b) OR WNWS 3.5 miles W of Wintersburg Rd.,2 miles S of Buckeye-Salome Rd.

50 Milk (b) ENE12 Crosswinds Dairy,295th Ave. and Van Buren St.

51 Milk (b) E11 Butler Dairy, Palo Verde Rd and Southern Ave.

52 Vegetation WD SW3 Gavette residence,39326 W. Elliot Rd.

3 (b), Water j 53 Milk (b) E19 Kerr Dairy, Dean and Baseline Rds.

74 ODCM Rev. 7 l

TAllt.E 6-4 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE COLLEPON LOCATIONS LOCATION l SAMPLE SAMPLE NOTE (d) DESIGNATION IDCATION DESCRIPTION (c) !

SITE TYPE (a) i 54 Milk E17 Dickman Dairy, Broadway and Apache (Cemetery)

Rds.

l 55 CHANGED TO SITE 52 l 56 Milk (b) Control EMI Pew Dairy, McQueen and Ryan Rds., Chandler j 57 Ground WG onsite Well 27dde ;

i Water (b) 58 Ground WO onsite Well 34abb l Water (b) j 59 Surface WS onsite Evaporation Pond #1 [

Water (b) l 60 Surface WS onsite Reservoir f Water (b) [

j 62 Vegetation Control E35 Tolleson Produce Co.,91st Ave. and Van Buren St. l (b) l t

63 Surface WS onsite Evaporation Pond #2 j Water (b) i l .!'

l i

l NOTES: (a) Distance and direction are relative to the Unit 2 containment, rounded to the nearest mile.

(b) These samples fulfill the requirements of the ODCM, Table 6-1.

, (c) Refer to Figures 6-1,6-2, and 6-3 for relative locations of sample sites.

i (d) IR - inner ring OR - outer ring SP - school or population center WS - waterborne surface WG - nterborne ground WD - waterborne drinking I

i 75 ODCM Rev. 7

=. ,,

OffSite Dose Calculation Manual Palo Verde Nuclear Generating Station '

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i ceneranno station Boundary Radiolog.ical ' '

a H o H Gas Pipeline Environmental Monitoring

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8i i

76 ODCM Rev. 7

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77 ODCM Rev. 7

1 Offsite Dose Calculation Manual Palo Verde Nuclear Generating Station !

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l Offsi:e Dese Calculation Manua! Palo Verde Nuclear Generating Station cm -

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1 79 ODCM Rev. 7

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. Ficure 6-6 81 ODCM Rev. 7 l

7.0 Radiological Reports

7.1 Requirement

Semlannual Radioacthe Emuent Release Report

Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January I and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gascous Effluents from Light. Water-Cooled Nuclear Power Plants," Revision 1. June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),

or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability". This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous emuents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous emuents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 6-4) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of rarliation doses shall be performed in accordance with the rnethodology and parameters in the ODCM.

l A single submittal may be made for a multiple unit station. The submittal should combine those i sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. !

l In lieu of submission with the first half year Semiannual Radioactive Emuent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

l 82 ODCM Rev. 7

The Semiannual Radioactive Emuent Release Report to be submitted 60 days after January 1 of each year shall also inchde an assessment of radiation doses to the likely most crposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary emuent pathways and direct radiation, for the presions calendar year to show conformance with 40 CFR Part 190, Emironmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contributions are given Section 5.0 and Regulatory Guide 1.109 Rev.1, October 1977.

The Semiannual Radioactive Etiiuent Release Reports shall include the following information for l cach class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

l a. Container volume, l

b. Total curic quantity (specify whether determined by measurement or estimate),

, i

c. Principal radionuclides (specify whether determined by measurement or estimate), .

l l d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, cvaporator bottoms),

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l e. Type of c(mtainer (e.g., LSA, Type A, Type B, Large Quantity), and l l

f. Solidification agent or absorbent (e.g., cement, urea formaldehyde).

{

The Semiannual Radioactive Emuent Release Repons shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous .

i and liquid emuents made during the reporting period. !

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the l reporting period to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE ,

l CALCULATION MANUAL, as well as a listing of new locations for dose calculations and/or emironmental monitoring identified by the land use census pursuant to Section 6.2.

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7.2 Requirement

Annual Radiok>gical Environmental Operating Report

Routine Annual Radiological Emironmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May I of the year following criticality.

The Annual Radiological Emironmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological emironmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous emironmental surveillance reports, and an assessment of the obsen'ed impacts of the plant operation on the emironment. The reports shall also include the results ofland use censuses required by Section 6.2.

I

The Annual Radiological Emironmental Operating Reports shall include the results of analysis of l all radiological emironmental samples and of all emironmental radiation measurements taken during the period pursuant to the locations specified in Table 6-4 and Figures 6-1,6-2, and 6-3., as well as summarired and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. He missing data shall be submitted as soon as possible in a supplementary report.

De reports shall also include the following: a summary description of the radiological environmental ,

I monitoring program; at least two legible maps" covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Section 6.3; discussion of all desiations from the sampling schedule of Table 6-1; and discussion of all analyses in which the LLD required by Table t

6-3 was not achievable.

l A single submittal may be made for a multiple unit station.

One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant '

stations.

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84 ODCM Rev. 7

APPENDIX A DETERMINATION OF CONTROLLING LOCATION The controlling location is the location of the MEMBER OF THE PUBLIC who receives the highest doses.

The determination of a controlling location for implementation of 10CFR50 for radiciodines and particulates is known to be a function of:

(1) Isotopic release rates (2) Meteorology (3) Exposure pathway (4) Receptor's age The incorporation of these parameters into Equation 5-2 results in the respective equations at the controlling location. The isotopic release rates are based upon the source terms calculated using the PVNGS Emironmental Report, Operating License Stage, Table 3.5-12, without carbon.

All of the locations and exposure pathways, identified in the 1984 Land Use Census, have been evaluated.

These include cow milk ingestion, goat milk ingestion, vegetable ingestion, inhalation, and ground planc exposure. An infant is assumed to be present at all milk pathway k) cations. A child is assumed to be present at all vegetable garden locations. The ground plane exposure pathway is only considered to be present where an infant is no present. Naturally, inhalation is present everywhere an individual is present.

For the determination of the controlling k) cations, the highest X/Q and D/Q values, based on the 9 year meteorological data base, for the vegetable garden, cow milk, and goat milk pathways, are selected for each unit. The receptor organ doses have been calculated at each of these locations. Based upon these calculations, it is determined that the controlling receptor pathway is a function of unit location. For Unit 1, the controlling receptor is a garden-child pathway; for releases from Unit 2 and Unit 3 the controlling zeceptor is a cow milk-infant pathway. These determinations are based upon Table 4-16 which,in turn, is based upon the 1984 land Use Census. Locations of the nearest residences, gardens and milk animals, as determined in the 1984 Land Use Census, are given in Table 4-16.

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1 85 ODCM Rev. 7 i

I APPENDIX B Bases for Requirements i

B-2.1 RADIOACTIVE GASEOUS EITLUENT MONITORING INSTRUMENTATION The radioactive gaseous ef0uent instrumentation is prcwided to monitor and control, as applicable, the releases of radioactive materials in gaseous emuents during actual or potential releases of gaseous efiluents. The alarm / trip setpoints for these instruments shall be calculated and adjusted in '

accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is mnsistent with the requirements of Oeneral Design Criteria 60,63,64 of Appendix A to 10 CFR PART 50.

There are two separate radioactive gaseous effluent monitoring systems: the low range effluent monitors for normal plant radioactive gaseous emuents and the high range emuent monitors for post-acrident plant radioactive gaseous effluents. The low range monitors operate at all times until the concentration of radioactivity in the effluent bemmes too high during post-accident mnditions.

The high range monitors only operate when the concentration of radioactivity in the effluent is above the setpoint in the low range monitors.

11- 3.1 GASEOUS EFFLUENT - DOSE RATE This requirement is provided to ensure that the dose at any time at and beyond the SITE ,

BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits l of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specifled in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who !

may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC ,

will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor i above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE !

PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the mrresponding gamma and beta dose rates above background to l a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 ,

mremshear to the total body or to less than or equal to 3000 rirems/ year to the skin. 'Ihese release rate limits also restrict , at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year.

This requirement applies to the release of radioactive materials in gaseous effluents from all reactor units at the site.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD,and other detection limits can be found in HASL Procedures Manual, H ASI 300 (revised annually), Currie, L. A.,' Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal.

Chem. 40, 586-93 (1968), and Hartwell, J. K-, " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

86 ODCM Rev. 7 i

l B-3.2 SECONDARY SYSTEM LIQUID WASTE DISCllARGE TO ONSITE EVAPORATION PONDS -

CONCENTRATION This requirement is provided to ensure that at any time during the life of the nuclear station, the !

annual total body dose due to ground contamination of an UNRESTRICTED AREA, arising from transportation and deposition by wind of the accumulated activity discharged to the pond from the secondary system of the plant (if the pond gets dried up) on the UNRESTRICTED AREA,is within the guidelines of 10 CFR Part 20 for the above-mentioned postulated event Restricting the mncentrations of the secondary liquid wastes discharged to the onsite evaporation ponds will restrict the quantity of radioactive material that can get accumulated in the ponds. His, i in turn, provides assurance that in the event of an uncontrolled release of the pond's mntents to an i UNRESTRICTED AREA, the resulting total body annual exposure from ground contamination to a MEMBER OF THE PUBLIC at the nearest exclusion area boundary will be within 0.5 rem.

This requirement applies to the secondary system liquid waste discharges of radioactive materials from ;

all reactor units to the onsite evaporation ponds. Since the chemical neutralizer tank concentrations

! will bound concentrations in other secondary waste discharges, surveillance requirements stipulate that sampling and analysis of other secondary waste discharges need be performed only if the sampling and analysis of the contents of the chemical neutralizer tank shows that the neutralizer tank concentration exceeds the specified LLD.

The required detection capabilities for radioactive materials in the secondary liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL 300 (revised annually), Currie,

! L A., 'L,imits for Qualitative Detection and Quantitative Determination - Application to l Radiochemistry," Anal. Chem. 40. 586-93 (1968), and Hartwell, J. K., "Detectian Limits for ;

l Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

B-4.1 GASEOUS EFFLUENT - DOSE, Noble Gases This requirement is provided to implement Sections ll B, III.A and IV.A of Appendix I,10 CFR Part

50. This requirement implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable.' The surveillance requirements implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. De dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,' Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50. Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111," Methods for Estimating Atmospheric Transport and Dispersion of Gaseous EfHuents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. De ODCM equations provided for determining the air doses at and beyond the SITE i BOUNDARY are based upon the historical average atmospheric conditions.

i This requirement applies to the release of radioactive materials in gaseous effluents from each reactor !

unit at the site.

l 87 ODCM Rev. 7 l

B-4.2 GASEOUS EITLUENT - DOSE - Iodine - 131, Iodine-133, Tritium, and All Radionuclides in i Particulate Form With Italf-Lives Greater "Ihan 8 Days l This requirement is prosided to implement the requirements of Sections ll.C, III.A, IV.A of Appendix 1,10 CFR Part 50. 'Ihis requirement is the guide set forth in Section II.C of Appendix I. The ACTION statements preside the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept *as low as is reasonably achievable.' The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix 1 be shouti by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109," Calculation of Annual Doses to Man from Routine Releases of Reactor Efiluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,* Revision 1, October 1977 and Regulatory Guide 1.111,

Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efiluents in Routine Releases for Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric cxmditions. The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were l

examined in the development of these calculations were: (1) indisidual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.

This requirement applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

B-4.3 GASEOUS RADWASTE TREATMENT The OPERABILITY of the G ASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use w henever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used,when specified, provides reasonable assurance that the releases of radioactive materials in gaseous efiluents will be kept *as low as reasonably achievable." This requirement implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The l specified limits governing the use of appropriate portions of the systems were specified as a suitable l fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix 1,10 CFR Part

! 50, for gaseous effluents.

This requirement applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

The minimum analysis frequency of 4/M (i.e. at least 4 times per month at intervals no greater than ,

9 days and a minimum of 48 times a year) is used for certain radioactive gaseous waste sampling in l Table 3-1. This will climinate taking doubic samples when quarterly and weekly samples are required ,

at the same time. )

1 88 ODCM Rev. 7 I

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B-4.4 SECONDARY SYSTEM IJQUID WASTE DISCILARGE TO ONSITE EVAPORATION PONDS -

DOSE nis requirement is provided to implement the requirements of Sections ll.A, Ill.A and IV.A of Appendix 1,10 CFR Part 50. This requirement implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required operating ficxibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive material in liquid effluents to UNRESTRICPED AREAS will be kept 'as low as is l reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially l affected by plant operations, there is reasonable assurance that the operation of the facility will not

( result in radionuclide concentrations in the finished drinking water that are in excess of the l

requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section Ill.A of Appendix 1 that mnformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual

! exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology l provided in Regulatory Guide 1.109,

Calculation of Annual Doses to Man from Routine Releases j of Reactor Efnuents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,*

Revision 1 October 1977 and Regulatory Guide 1.113,

Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose ofimplementing Appendix 1," April 1977.

This requirement applies to the relcase of liquid effluents from each reactor at the site. For units l with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.

! B-5.1 TOTAL DOSE AND DOSE TO PUBLIC ONSITE l This requirement is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. He requirement specifies the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix 1, and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the

, Special Report, it may be assumed that the dose mmmitment to the MEMBER OF THE PUBLIC l

from other uranium fuel cycle soura.s is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part j 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected),in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405e,is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is complcted. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Section 3.2 and 3.1 of the ODCM. An individual is not ccmsidered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

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89 ODCM Rev. 7 l

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l B-6.1 RADIOlhGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP) l l

l De Radiological Emironmental Monitoring Program required by this requirement prosides j representative measurements of radiation and of radioactive materials in those exposure pathways and J l for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OFTHE ,

PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 l of Appendix 1 to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring i l

program by verifying that the measurable concentrations of radioactive materials and levels of l

radiation are not higher than expected on the basis of the effluent measurements and the modeling ;

of the emironmental exposure pathways. Guidance for this monitoring program is prosided by the Radiological Assessment Branch Technical Position on Emironmental Monitoring. He initially i

specified monitoring program will be effective for at least the first 3 years of commercial operation.

l Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). De LLDs required by Table 6-3 are considered optimum for ,

routine emironmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

( Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures

! Manual, HASI 300 (revised annually), Currie, L A.,

Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., ' Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

B-6.2 1AND USE CENSUS This requirement is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological emironmental monitoring program are made if required by the results of this census. De best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B3 of Appendix 1 to 10 CFR Part 50.

Restricting the census to gardens of greater than 50 m 2provides assurance that significant exposure l pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kgtear) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following ,

assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., i similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, l B-6.3 INTERIABORATORY COMPARISON PROGRAM l

The requirement for participation in an approved Interlaboratory Comparison Program is presided l to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in emironmental sample matrices are performed as part of the quality assurance program for emironmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix 1 to 10 CFR Part 50.

90 ODCM Rev. 7

i APPENDIX C t Definitions Note:

The folk) wing definitions are from the ANPP Technical Specifications. These selected definitions support i those portions of the Technical Specifications which were transferred to the ODCM and have been l incorporated into the Requirements sections of the ODCM.

l Definitions:

The defined terms of this section appear in capitalized type and are applicable throughout the Requirements sections of this ODCht ACTION l

ACT!ON shall be that part of a requirement which prescribes remedial measures required under designated ,

conditions.

CH ANNEL CALIBRATION l A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it l responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK l A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall inclede, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST l A CHANNEL FUNCTIONAL TEST shall be:

l a. Analog channels - the injection of a simulated signal into the channel as close to the sensor l as practicable to verify OPERABILITY including alarm and/or trip functions. ;

I

b. Bistable channels - the injection of a simulated signal into the sensor to verify l OPERABILITY including alarm and/or trip functions.

c. Digital computer channels - the exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY including alarm and/or trip functions.

l 91 ODCM Rev. 7

APPENDIX C Definitions (Continued)

d. Radiological effluent process monit) ring channels - the CHANNEL FUNCrlONAL TEST may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is functionally tested. i ne CHANNEL FUNCTIONAL TEST shall include adjustment, as necessary, of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.

DOSE EOUIVALENT I-131 DOSE EQUIVALENT l-131 shall be that concentration of I-131 (microcuries/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131,1-132,1-133,1-134 and 1-135 actually present. The thyroid dose com'ersion factors used for this calculation shall be those listed in Table 111 of TID-14844, Calculatien of Distance Factors for Power and Test Reactor Sites.

FREOUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table C-1. l GASEOUS RADWASTE SYSTEM A G ASEOUS CADWASTE SYSTEM shall be any system designed and installed to red uce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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

OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL shall contain the methodologyand parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents,in the calculation of gaseous and liquid effluent monitoring AlarmTTrip Setpoints, and in the conduct of the Emironmental Radiological Monitoring Program. The ODCM shall also contain:

(1) the Radioactive Efiluent Controls and Radiological Emironmental Monitoring Programs required by Technical Specification Section 6.8.4, and ,

1 (2) descriptions of the information that should be included in the Annual Radiological Emironmental Operating and Semiannual Radioactive Efflucat Release Reports required by Technical Specifications 6.9.1.7 and 6.9.1.8.

92 ODCM Rev. 7 l

l J

i APPENDIX C Definitions (Continued) ,

OPERABLE-OPERABILITY l

l 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 equipment that are required for the I system, subsystem, trair, mmponent, or device to perform its function (s) are also capable of performing their related support function (s).

OPER ATIONAL MODE-MODE An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive mmbination of mre reactivity condition, power level, and cold leg reactor mo' ant temperature specified in Table C-2.

PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be aaomplished in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste. i PURGE-PURGING ;

PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition,in such a manner that ,

replacement air or gas is required to purify the confinement.

RATED THERM AL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor molant of 3800 MWL SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SOLIDIFICATION SOLIDIFICATION shall be the ccmversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (hee-standing).

SOURCE CHECK I

A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is !

exposed to a source of increased radioactisity.

93 ODCM Rev. 7 i

i

a APPENDIX C ,

t 4

Definitions (Continued)

[

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THERMAL POWER l

THERMAL POWER shall be the total reactor core heat transfer rate 13 the reactor molant. '

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for the purposes of protection of individuals from erposure to radiation and '

radioactive materials, or any area within the SITE BOUNDARY used for residential q zarters or for industrial, mmmercial, institutional, and/or recreational purposes.

i VENTTLATION EXH AUST TREATMENT SYSTEM l t A VENTILATION EXHAUSTTREATMENT SYSTEM shall be any system designed and installed to reduce !

gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charmal adsorbers and/or HEPA filters for the purpose of remosing iodines or particulates from the gaseous exhaust stream prior to the release to the emironment. Such a system is not -

considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that j replaament air or gas is not provided or required during VENTING. Vent, used in system names, does not !

j imply a VENTING process. i a ,

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. 94 ODCM Rev. 7

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1 TABLE C 1 1 FREQUENCY NOTATION j NOTATION FREOUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

^

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days. ,

4/M At least 4 times per month at intervals no greater than 9 days and a minimum of 48 times per year ;

M At least once per 31 days.

O At least once per 92 days.

SA At least once per 184 days. ;

R At least once per 18 months.

P Completed prior to each release.

'! SiU Prior to reactor startup.

N.A. Not Applicable.

TABLE C-2 l Operational Modes Operational Mode Reactivity % of Rated Cold Irg Condition,1Q Thermal Power

Temperature (T )

1. POWER OPERATION a 0.99 >5% a350*F

2. STARTUP a 0.99 s5% a 350" F l

3. HOT STANDBY < 0.99 0 a 350" F

4. HOT SHUTDOWN < 0.99 0 350* > Ta > 210"F

5. COLD SHUTDOWN < 0.99 0 s 210' F

6. REFUELING *

s 0.95 0 s 135* F Excluding decay heat.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

95 ODCM Rev. 7

. . __ __. . _ _ _ . _ . _ _ m _ ___ _ . = _

Appendix D Disposillon of NRC Generic letter 89-01 Items from the PVNGS Technical Specifications to the ODCM NUREG 0472 Tech Spec # PVNGS T.S. # ODCM ltem Disposition Table 1.2 Table 1.1 Tabic C-1 FREQUENCY NOTATION Table retained in Technical Specifications and duplicated in the ODCM.

N/A Table 1.2 Table C-2 OPERATIONAL MODES Table retained in Technical Specifications and duplicated in the ODCM.

1.17 1.18 Apx C OFFSITE DOSE CALCULATION Definition incorporated in Tech nical Specifications MANUAL and the ODCM definitions.

130 1.24 Apx C PROCESS CONTROL PROGRAM Definition incorporated in Tech nical Specifications and the ODCM definitions.

131 132 Apx C SOLIDIFICATION Definition deleted from Technical Specifications and relocated to the ODCM and PCP.

3/4 3.3.10 N/A N/A RADIOACTIVE LIQUID EFFLUENT This item does not exist in the PVNGS MONITORINO INSTRUMENTATION Technical Specifications since there are no liquid effluents.

3/433.11 3/433.8 2.1 RADIOACTIVE OASEOUS EFFLUENT Relocated to the ODCM.

MONITORINO INSTRUMENTATION Existing requirements for explosive gas monitoring instrument-action are retained in the Technical Specifications.

Tabic 33-13 Table 33-12 Table 2-1 RADIOACTIVE OASEOUS EFFLUENT Relocated to the ODCM.

MONITORINO INSTRUMENTATION Table 43-13 Table 4.3-8 Table 2-2 RADIOACTIVE OASEOUS EFFLUENT Relocated to the ODCM.

MONITORING INSTRUMENTATION

% ODCM Rev. 7

l l

Appendix D (Continued)

Disposition of NRC Generic Ixiter 89-01 Items from the PVNGS Technical Specincations to the ODCM NUREG 0472 Tech Spec # PVNGS T.S. # ODCM Item Disposition 3/4.11.1.1 3/4.11 1.1 3.2 LIQUID EFFLUENTS: CONCENTRATION Relocated to the ODCM.

Table 4.11-1 Table 4.11-1 Table 3-5 RADIOACTIVE LIQUID WASTE Relocated to the ODCM.

SAMPLING AND ANALYSIS PROGRAM 3/4.11.1.2 3/4.11.1.2 4.4 LIQUID EFFLUENTS: DOSE Relocated to the ODCM.

3/4.11.1.3 N/A LIQUID EI-TLUENTS: LIQUID 'lliis item does not exist in the PVNGS RADWASTE TREATMENT SYSTEM Technical Specifications since there are no liquid efiluents.

3/4.11.1.4 3/4.11.1.3 N/A LIQUID HOLDUP TANKS Existing specification requirements are retained in the Technical Specifications.

3/4.11.2.1 3/4.11.2.1 3.1 GASEOUS EFFLUENTS Relocated to the ODCM.

DOSE RATE Table 4.11-2 Table 4.11-2 Table 3-1 RADIOACTIVE GASEOUS WASTE Relocated to the ODCM.

SAMPLING AND ANALYSIS PROGRAM 3/4.11.2.2 3/4.11.2.2 4.1 GASEOUS EFFLUENTS: Relocated to the ODCM.

DOSE-NOBLE GASES 3/4.11.2.3 3/4.11.2.3 4.2 GASEOUS EFFLUENT 3: DOSE- 1-131, Rekx:ated to the ODCM.

1-133, Tritium, and Radioactive Material in Particulate form.

97 ODCM Rev. 7

Appendix D (Continued)

Disposition of NRC Generic letter 89-01 Items from the PVNGS Technical Specifications to the ODCM NUREG 0472 Tech Spec # PVNGS T.S. # ODCM Item Disposition 3/4.11.2.4 3/4.11.2.4 4.3 GASEOUS EFFLUENTS: Relocated to the ODCM.

Gascous Radwaste Treatment or Ventilation Exhaust Treatment System 3/4.11.2.5 3/4.11.2.5 N/A EXPLOSIVE GAS MIXTURE Retained in the Technical Specifications.

3/4.11.2.6 3/4.11.2.6 N/A GAS STORAGE TANKS Retained in the Technical Specifications.

3/4.11.3 3/4.11.3 N/A SOLID RADIOACTIVE WASTES Relocated to the PCP.

3/4.11.4 3/4.11.4 5.1 RADIOACTIVE EFFLUENTS: Relocated to the ODCM.

Total Dose 3/4.12.1 3/4.12.1 6.1 RADIOLOGICAL ENVIRONMENTAL Relocated to the ODCM.

MONITORING: Monitoring Program Tabic 3.12-1 Tabic 3.12-1 Tabic 6-1 RADIOLOGICAL ENVIRONMENTAL Rekx:ated to the ODCM.

MONITORING PROGRAM Table 3.12-2 Table 3.12-2 Table 6-2 REPORTING LEVELS FOR RADIO- Relocated to the ODCM.

ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Table 4.12-1 Table 4.12-1 Table 6-3 DETECTION CAPABILITIES FOR Relocated to the ODCM.

ENVIRONMENTAL SAMPLE ANALYSIS 3/4.12.2 3/4.12.2 6.2 RADIOACTIVE ENVIRONMENTAL Relocated to the ODCM.

MONITORING: Land Use Census 98 ODCM Rev. 7

Appendix D (Continued)

Disposition of NRC Generic Ixtter 89-01 Items from the PVNGS Technical Specifications to the ODCM NUREG 0472 Tech Spec # PVNGS T.S. # ODCM Item Disposition 3/4.12.3 3/4.12.3 6.3 RADIOACTIVE ENVIRONMENTAL Relocated to the ODCM.

MONITORING: Interlaboratory Comparison Program DESIGN FEATURES:

Figure 5.1 1 Figure 5.1-1 Figurc 6-4 SITE AND EXCLUSION BOUNDARIES Figure revised in Technical Specifications and duplicated in the ODCM.

Figure 5.1-2 Figure 5.1-2 Figurc 6-6 LOW POPULATION ZONE Figure revised in Technical Specifications and duplicated in the ODCM.

Figurc 5.1-3 Figure 5.1-3 Figure 6-5 GASEOUS RELEASE POINTS Figure revised in Technical Specifications and duplicated in the ODCM.

N/A 6.8.6.g N/A Radioactive Effluent Controls Program New Section is ad&d to Technical Specifications to address programmatic controls being relocated to the ODCM.

N/A 6.8.6.h N/A Radiological Emironmental Monitoring New Section is added to Technical Specifications Program to address programmatic controls being relocated to the ODCM.

6.9.1.3 6.9.1.7 7.2 REPORTING REQUIREMENTS: Rekxated to the ODCM and simplified in Technim!

Annual Radiological Environmental Specifications.

Operating Report 99 ODCM Rev. 7

Appendix D (Continued)

Disposition of NHC Generic letter 89-01 Items from the PVNGS Technical Specifications to the ODCM NUREG 0472 Tech Spec # PVNGS T.S. # ODCM Item Disposition 6.9.1.4 6.9.1.8 7.1 REPORTING REQUIREMENTS: Relocated to ODCM and simplified in Technical Semiannual Radiological Effluent Specifications.

Release Report N/A 6.10.2.q N/A RECORD RETENTION New section is added to Technical Specifications to address records of reviews performed for changes made to the ODCM and PCP.

6.13 6.13 N/A PROCESS CONTROL PROGRAM Technical Specification requirements simplified.

6.14 6.14 N/A OFFSITE DOSE CALCULATION Technical Specification requirements simplified.

MANUAL 6.15 6.15 N/A MAJOR CHANGES TO LIQUID, No changes, retained in Technical Specifications.

GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS 100 ODCM Rev. 7

Appendix D (Continued)

Disposition of NRC Generic letter 89-01 Items from the l'VNGS Technical Specifications to the ODCM NUREG 0472 Tech Spec # l'VNGS T.S. # ODCM ltem Disposition i

BASES He BASES for the above sections that were relocated from the Technical Specifications to the ODCM are also relocated to the ODCM, Appendix B. For convenience, the section references arc included below.

t 3/4.3.3.10 3/43.3.8 2.1 3/4.11.1.1 3/4.11.1.1 3.2 3/4.11.1.2 3/4.11.1.2 4.4 3/4.11.2.1 3/4.11.2.1 3.1 3/4.11.2.2 3/4.11.2.2 4.1 3/4.11.2.3 3/4.11.2.3' 4.2 3/4.11.2.4 3/4.11.2.4 4.3

3/4.11.4 - 3/4.11.4 5.1 3/4.12.1 3/4.12.1 6.1 3/4.12.2 3/4.12.2 6.2 1

3/4.12.3 3/4.12.3 6.3 i

l 101 ODCM Rev. 7 I

_ _ . - _ . . _ . _ _ . _ _ ,_ . _ _ _ . _ _ _ . - - . _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ _ _ __ _a.- - . _ _ _ _ . _ _ . - - - - - - - - - . - - . . . , , . . . - . - . . . . -. 4 e, .-, .- ..m.,-.e... _.--,w,. . ,-. . . . _ - . . ._m .-=_ - - ,

l REVISION REQUEST FORM l DATE: 5-11-93 i ORIGINATOR: Louis Drinovsky EXT: 6055 PAGE 1 OF 3 l

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i Description and Justification of Revision: !

i I. Introduction j The Radiological Em-ironmental Monitoring Program (REMP) is required to be performed as per !

the Offsite Dose Calculation Manual (ODCM). Section 6.0 of the ODCM defines the REMP, l including specified sample locations. Based on findings during the annual performance of the Land .

Use Census, some changes in sample locations were deemed necessary. Additionally, one new vegetation sample was added to the program. Each change along with its justification is documented ;

below (editorial changes were also made and are listed, but without justifications). None of the changes affect the level of radioactive effluent control since the REMP is designed to verify the ,

effectiveness of the in-plant measures used for controlling the release of radioactive materials.-

l Changing sample locations will not adversely impact the accuracy or reliability of effluent, dose, et j l setpoint calculations. ;

l II. Table 6-1 changes -

A. Direct radiation section l

i

1. The total number of required stations for TLDs was incteased from 40 to 41 with the i addition of site #50. This TLD site is located between four and five miles from PVNGS in the WNW sector (considered the

outer ring *) and has been in use since 1985. This site was previously identified as being inaccessible and was excluded as ,

a required outer ring TLD location in the WNW sector. l B. Waterborne (surface) section

1. Evaporation Pond 2 (site #63) was not included as a sample location in Technical I Specification Table 3.12-1 because the pond was constructed after the initiallicense I was issued. Therefore, when Table 3.12-1 was moved to the ODCM from the !

Technical Specification in September,1992. Evaporation Pond 2 was not included. l This sample location was added to the table. Even though the sample location was

. l not specified in the table,it has always been included in Table 6-4 of the ODCM and I is sampled when the pond contains water. ;

1 C. Food Products section l

1. The I-131 analysis in the past was performed by a radiochemical separation technique in order to meet the analytical sensitivity requirements. The gamma isotopic analysis can now determine the 1-131 concentration of vegetation well within the required 4

sensitivity as required in Table 6-3. Specific reference to the I-131 analysis was deleted since it is performed as part of the gamma isotopic analysis which is still a requirement of Table 6-1.

REVISION REOUEST FORM (continued) t i PAGE 2 OF 3 !

Table 6-1 Changes (continued) ;

D. Table Notations j

1. The table notation previously contained a reference to the potential for channeling in the charcoal canisters. This notation has been deleted. There is no apparent i

problem with channeling in canisters at the low volume flow rates used nor is there j a visual examination which can be used to make such a determination. This note was [

also deleted from USNRC Branch Technical Position, Rev.1 (Nov.1979) upon its l incorporation into NUREG-1301. !

/

! 2. The table notation previously contained n statement in note 'a' which required a l J submittal 'for acceptance' of alternative sample media or sample locations in e instances where the most desired samples were not available. This statement has !

been deleted. The ODCM contains a mechanism for obtaining replacemem samples. {

]

i This statement was also deleted from USNRC Branch Technical Position, Rev.1 )

(Nov.1979) upon its incorporation into NUREG-1301. l j

III. Table 6-4 changes !

t i A. The following changes are considered editorialin nature and require no explanation (sample l l

locations remain the same as previous, some location designations were edited to reflect map - j locations from Figures 6-1,6-2 and 6-3). ;

) .

samtile site chance j 5 location description f 6 location description 24 location description 35 location designation and descriptir t - !

! 40 location description {

45 changed ' transit' to ' transit control' i 46 location designation and description j 48 location description

50 location description, added 'OR' notation

, 53 location designation and description 56 location designation B. The following changes constitute additions or replacement of sample locations and include justifications:

1. Siw s, formerly Scott residence drinking water The Scott residence drinking water location was replaced with the Chowanez residence, effective 4-1-93. The previous location was ESE4 while the new location is NNE2. It has been determined that the Scott water supply is a community water supply which is more distant from PVNGS. The desired pathway would be an untreated water supply that could be affected by discharges from PVNGS. Even though PVNGS does not discharge liquid i effluents. the Chowanez well is nearer PVNGS and untreated which is closer l to fitting a potential pathway to man.

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REVISION REOUEST FORM (continued)

PAGE 3 OF 3 j Table 6-4 changes (continued) l l

l Additionally, this new sample point is presently sampled by the Arizona '

l Radiation Regulatory Agency and will provide a second drinking water duplicate sample location.

2. Site #52, formerly unused 1

Drinking water samples have been obtained from the Gavette residence as a supplemental sample location for many years. This resident was identified as a new broad leaf vegetation sample location within 5 miles of PVNGS during the performance of the 1.snd Use Census. The former designation as site #55 was changed since the ODCM designates site #52 for use as a vegetation sample location.

3. Site #55, formerly Gavette residence drinking water See discussion for no. 2 above.

4. Site #62, formerly J.A. Wood Co. vegetation The J.A. Wood Co. (E75) was replaced with the Tolleson Produce Co. as the vegetation control sample location effective 1-1-93. The previous location did not prove to be a dependable source of broad leaf vegetation after management of the property was changed. The new location is more readily accessible, has proven to be a more dependable source of samples, and more closely meets the ODCM requirement for a control sample ,

location 15-30 km from PVNGS.

IV. Figures 6-1,6-2 and 6-3 The sample locations identifled on Figures 6-1,6-2 and 6-3 were changed to reflect changes made in Tables 6-1 and 6-4.

V. Additional Editorial Changes Sections 6.1.2 and 6.2.2:

Changed " Station Manual Procedures" to ' Nuclear Administrative and Technical Manual Procedures."

Approved By: m / M -- Date: 5-29-93 R'adiological !donitoring Sgptisor (Site Chemistry) 1 i

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REVISION REOUEST FORM DATE: 05-17-93 i

ORIGINATOR: Kevin Kutner EXT: 82-6154 PAGE 1, OF 2 Description and Justification of Revision: -

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Title page >

Rev 6 to Rev. 7 :

I Page v :

Deleted figures 2-1 and 2 2. There is no requirement or reason to have these figures in the ODCM. l Page 1 Section 1.0, First paragraph Added discussion to indicate that Technical Specification section 3/4.0 applies to the ODCM (i.e 25% l exte sion for surveillance testing). This information is part of NUREG-1301, April 1991. ICR 60743 i l

Page 5, Section 2.1, Applicability i Added "This requirement does not apply to RU-141 or RU-142 if DCP-13-PJ-SQ-065 has been '

implemented." RU-141 and RU-142 are no longer effluent monitors after DCP-13-PJ-SQ-065 has been i implemented.

Page 12, Section 2.1.2, second paragraph !

Added *The RU-141 alert alarm setpoint may be further reduced to provide early indication of steam j generator tube leakage." Lowering the RU-141 alert alarm setpoint is consistent with the discussion (

provided in NRC Information Notice No. 91-43.

l Page 16, Section 2.1.2.4, bottom of page Added

Note - The RU-141 setpoints may be further reduced to provide early indication of steam generator tube leakage." Lowering the RU-141 alert alarm setpoint is consistent with the discussion j provided in NRC Information Notice No. 91-43. t 2.1.2.4 'I g

2.1.2.5 :

3.1' . .

3.2.2 Changed ' Station Manual Procedures

to " Nuclear Administrative and Technical Manual procedures". l This is an editorial change only. !

Page 17. Section 2.1.2.5 !

Deleted the last two paragraphs. This information is not required in the ODCM. I i

l Pages 18 and 19, Figures 2-1 and 2 ' '

Deleted. This information is not required in the ODCM. j l ;

) Page 18, Section 3.1, applicability l Added "This requirement does not apply to the Condenser Vacuum Pump Exhaust if DCP 13-PJ-SQ- l 1 065 has been implemented." RU-141 and RU-142 are no longer effluent monitors after DCP-13-PJ- l SQ-065 has been implemented. l i

Page 31, Section 4.1.2 Changed reference from Table 3-1 to Table 3-3 (typo). ICR 55494 i

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REVISION REQUEST FORM l

DATE: 05-17-93 ORIGINATOR: Kevin Kutner EXT: 82-6154 PAGE2OF2 Description and Justification of Revision:

l Page 80, Figure 6-5 Indicates that the Condenser Exhaust (RU-141) discharge point is combined with the Plant Vent if DCP-13-PJ-SQ-065 has been implemented.

i These changes will maintain the level of radioactive effluent control required by 10 CFR 20.106,40 CFR Part 190,10 CFR 50.363, and Appendix I to 10 CFR Part 50 and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

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l Approved By: m MMr RMS/ Effluents Supervisor (gChemistry)

Date: I-29-7J 1

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10CFRSO.59 SCREENING AND EVALUATION 1 y 7 OCTION UNDE R REVif W: H E vi$lON. PCN CFFSITE M6 c4KULA1)oO it4NU4L _ IMLEDENQi?oll o!L td g).pp$Q.cL9 m CRtd 0%oR4f AT 72 7 M4 CESCR; PYLON OF PAOPOLED cMANGE SEE DEscrirpa1 ON ME 2 f

10CFR50.59 SCREEN (Provide References on Response Justification Page) NO YES Does the proposed change:

1. Make changes in the facility as it is described in the UFSAR? X

2. Make changes in procedures as they are described in the UFSAR7 5

3. Involve test or experiments not described in the UFSAR? .3 -

4. Require a change to the technical specifica'..ons? 1 -

Any answer to questions 1 throvoh 3 "YES." then a 10CFR50.59 evaluation is required. Contact Document Control at ext. 82-6633 to obtain a tracking tog number and enter tne number in the Evaluation Log number block above. UFSAR Change Request per procedure 93AC-OLC01 may also be required.

Answer 4 is "YES." then Technical Specification Change Request per procedure 93AC-OLCD1 and NRC approval is required prior to implementation.

X All answers 1 through 4 are 'NO " no 10CFR50.59 Evaluation required or Technical Specification change required. recommend action approval.

80CFR50.59 EVALUATION (Provide Response Justification with References) 5 May the probability of an accident previously evaluated in the UFSAR be increased?

6. May the consecuences of an accident previously evaluated in the UFSAR be increased?

7. May the probability of a malfunction of equipment important to safety be increased? _ _

8. May the consecuences of a malfunction of equipment important to safety be increased?

9. May the possibility of an accident of a different type than any previously evaluated in the UFSAR be created? _ _

10. May tne possibility of a different type of malfunction than any previously evaluated in the UFSAR be created? _ -

11. Is the margin of safety as defined in the basis for any technical specification reduced? - -

Any answer to questions 5 through 11 "YES," then an unreviewed safety question is identified. Proceed to procedure 93AC-OLCO3 prior to implementation.

All answers 5 th1ough 11 are "NO." there is no unreviewed safety cuestion and action approval is recommenced.

If UFSAR Chapter 6/ Chapter 15 is potentially affected, forward a copy of evaluation to Nuclear Fuels Management.

I versfy that the above screening! evaluation is adequate and accurate and that the undersigned has received required training.

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schEENER Evews rOR (Pam15 V Pewso m 5.oi M ta AthE *E A rPRmT) l 93AC-oNSD1 l

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10CFR50.59 REVIEW AND EVALUATION RESPONSE JUSTIFICATION r... .

g_. a v ACTION UN::ER RIVIEW:(NAMEl!ITLE) REVISION PCN OFFSITE DOSE CALCOLATION MANUAL, IMPIN.ATION OF DCP 13-PJ-SQ-065 AND CRDR C80245 ACTION 72 7 N/A PROCEDURE /PCPl!IMPORARY MODIFICATION NO:

N/A QUESTION RISPONsE JUSTIFICATION Description of proposed change DCP 13-PJ-SQ-065, in part, reroutes the Condenser Vacuum Exhaust from a separate release point, to a combined release point with the Plant Vent. The ODCM needs to be revised to reflect this change. The actual release will now be monitored by the Plant Vent effluent monitors, RU-143 and/or RU-144. RU-141 and RU-142 are no longer effluent monitors. RU-141 will become an in-duct monitor and RU-142 will be removed.

This ODCM revision removes RU-141 and RU-142 as effluent monitors following the implementation of DCP 13-P3-SQ-065.

In addition to the changes required to implement DCP 13-PJ-SQ-065, the setpoint methodology for RU-141 has been enhanced for Units that have not implemented DCP 13-PJ-SO-065. Additional flexibility has been added to allow the RU-141 setpoint to be lowered to provide early indication of steam generator tube leakage .

(CRDR 080245 Action 72).

1. This action does not require a change to the facility as described in the UFSAR. The changes to the facility have been identified in DCP 13-PJ-SQ-065, and justification is provided in 10CFR50.59 i Evaluation Img No. 92-00035. This action simply implements what has already been identified in the DCP.

Reference:

10CFR50.59 Evaluation Irg No. 92-00035 (attached).

2. There are no specific references to procedures for RU-141 or RU-142 setpoints in the UFSAR.

Reference UFSAR, Rev. 5, sections: 11.5, 12.5, 13.2, 13.5.

3. This action does not involve any tests or experiments.

4. Figure 5.1-3 will be revised as part of DCP 13-PJ-SQ-065. The Technical Specification change nas been identified in the 10CFR50.59 Evaluation Log No. 92-00035 as part of DCP 13-PJ-SQ-065. By procedure, the change to the technical specifications must be approved by the NRC prior to the implementation of this DCP. The changes to the ODCM do not require a change to the technical specifications. The ODCM revision will address condenser vacuum exhaust release points and setpoints whether DCP 13-PJ-SO-065 has been implemented or not.

References:

Technical Specifications Amendments- 69 (U-1),55 (U-2),42 (U-3)

NUREG-1301, April 1991 10CFR50.59 Evaluation log No. 92-00035 (attached).

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Evaluation Log No.: 97 00035 '

10CFR50.59 l

l SCREENING AND EVALUATION Pagelof 1 l l

cN uNoER REVIEW: REVistoN: PcN:

DCP 13-PJ-SQ-065 0 cescasca or eaceosto cannae: l Provide a functional separation of the Condenser Exhaust High Range High (HRH) and High Range j Normal (HRN) Efiluent Radiation Monitors (13JSQNRUO141 and 13JSQNRU0142) by rerouting the l l

condenser exhaust to the Plant Vent, removing Monitor 142, and converting Monitor 141 to in-duct.

(

10CFR50.59 SCREEN (Provide References on Response Justification Page) NO YES Does the proposed change: l

1. Make changes in the f acility as it is describedin the UFSAR?

2. Make changes in procedures as they are described in the UFSAR? X '

3. Involve test or experiments not described in the UFSAR? X

4. Require a change to thc technical specifications? X i X Any answer to questions 1 through 3 "YES" then a 10CFR50.59 evaluation is required. Contact Document Control at ext. 82-6633 to obtain a tracking tog number and enter the number in the .

Evaluatien Log number block above. UFSAR Change Request per procedure 93AC-OLC01 may also be required.

X Answer 4 is "YES", then Technical Specificat' on Change Request per procedure 93AC-OLC01 and NRC approval is required prior to implementation

~

All answers 1 through 4 are *NO", no 10CFR50.59 Evaluation required or Technical Specircation i change required. recommend action approval. !

10CFR50.59 EVALU ATION (Provide Response Justification with References) i

5. May the probability of an accident previously evaluated in the UFSAR be increased? X i t

6. May the consequences cf an accident previously evaluated in the UFSAR be increased?

X

7. May the probsbility of a ma'fune: ion of equipment important to safety be increased?

X i"

8. May the consequences of a malfunction of equipment importantto safety be increased?

X ;

9. May the possibility cf an accident of a different type than any previously evaluated in the UFSAR be created?

X :

10. May the possibility of a different type of malfunction than any proveusly evaluated in the UFSAR be created?

11. Is the margin of safety as defined in the basis for any technical specification reduced?

X l

Any answer to cuestions 5 through 11 "YES*, then an unreviewed safety question is icentified. Proceed to proceduro 93AC-OLC03 prior to implementation. .,

X All answers 5 through 11 are *NO," there is no unreviewed safety question and action approvalis l

i

)

recommended.

if UFSAR Chapter 6/Chacter 15 is potentially affected, forward a ecpy of evaluation to Nuc!aar Fueis Management.

'erify that the above screeningtevaluation is adequate and accurate and that the undersigned have received required training.

SCREENEREVAluAToR

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I PISPONSE JUSTIFICATION :

FAGE ? JF 5 ACTION UNDER REVIEWDCP 13-PJ-SQ-065 REVISION: 0 PCN:

Name/ Title -

PROCEDURE /PCP/ TEMP MOD NO .DCP 13-PJ-SQ-065 '

i '

OUESTIONSI FISPLJSE JUSTIFICATION

! 1.

This DCP will change equipment as it is described in the Updated Fmal Safety Analysis l l Report, dated March 20,1991 (including the Technical Specifications).

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This DCP affects the condenser air mmoval system (CARS) vent routing and the effluent i radiation monitors in that vent. The CARS cffluen: will no longer be vented separately .

but will be combined with the plant vent effluent. The existing plant effluent radiation ;

monitors will then serve to monitor both the plant and the CARS efficent. These existing '

plant effluent monitors consist of a low (or "nonnal") range unit and a high range unit. '

As with the existing high range CARS monitor (being removed), the high range plant effluent monitor is designated as a post accident monitor. The existing CARS normal :

range monitor will remain but will be changed to an in-duct monitor, t i

This is consistent with an overall philosophy of having, at the outlet of each contributing !

system, monitors which are capable oflocating local soumes of radiation while providing ;

for the necessary high mnge monitoring where the combined effluents leave the plant.

The following sections of the UFSAR wem reviewed for change as a result of this i modification:

9.4.2 '

Auxiliary Building (AC, Heating, Cooling, and Ventilation Systems) 9.4.4 Turbine Building (AC, Heating, Cooling, and Ventilation Systems) ;

9A Section 9A.37 (NRC Question 460.5) Response 10.4.2 Main Condenser Evacuation System !

10.4.2.1.1 Safety Design Basis 10.4.2.1.2 I Power Generation Design Bases 10.4.2.2 System Description (CARS) ,

10.4.2.3 - Safety Evaluation (CARS) 10.4.2.4 Tests and Inspections (CARS) '

10.4.2.5 Instrumentation Applications (CARS) 10.4.3 Turbine Gland Scaling System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 11.3 Gaseous Waste Management Systems 11.3.2 System Descriptions 11.3.2.1 Gaseous Radwaste System 11.3.2.2 Condenser Air Removal System

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10CFR50.59 SCFIENING AND EVALUATION RESPONSE JUSTIFICATION

, E PAGE 3 OF. 5 l i

ACTION UNDER REVIEW DCP 13-PJ-SQ-065 FIVISION: 0 PCN: I Na: e/ Title ;

I PROCEDU?I/PCP/TEF" MOD. NO.DCP 13-PJ-SQ-065 i

QUES TIONSli RESPONSE JUSTIFICATION j I

11.3.2.3 Turbine Gland Sealing System 11.3.3 Radioactive Releases 11.3.3.1 Plant Vent Stack 11.2.3.3 iurbine Building Ventilation Exhaust 11.3.3.4 Condenser Air Removal System 11.3.3.5 Turbine Gland Scaling System Exhaust 11.3.3.6 Dilution Factors 11.3.3.7 Estimated Concentrations 11.3.3.8 Estimated Doses i 11.5 Process and Effluent Radiological Monitoring and Sampling Systems ;

11.5.1 Design Bases ;

11.5.1.1 Normal Operation and Anticipated Operational Occurrences ,

11.5.1.2 Postulated Accidents !

11.5.2 System Description 11.5.2.1 Continuous Process, Effluent and Area Radiation Monitoring and !

Sampling Table 11.5-1 Radiation Monitors (

12.3.4 Area Radiation and Airbome Radioactivity Monitodng Instrumentation 18 TMI-2 I2ssons Leamed Implementation Report [

18.U.F.1.1 (NRC/PVNGS Positions on Plant Effluent Monitodng)

Also reviewed for change were the following separate documents:

Off-Site Dose Calculation Manual Emergency Plan (Table 5.1-1)

EPIP-02 Emergency Plan Implementing P ocedure #2 I EPIP-14 Emergency Plan Implementing Procedure #14 :

4xEP-xZZ01 Emergency Operations 4xEP-xR003 Emergency Operations, S/G Tube Rupture i 4xRO-xZZ06 Recovery Operations, S/G Tube Rupture 4xAO-xZZ08 Response Procedures, S/G Tube Leak t 4xOP-xAR01 Operating Procedures, CARS I 36ST-95Q09 RMS Calibration Test for RU141/142 i

) 36ST-9SQ04 RMS Quanerly Functional Tests l 74ST-95Q04 Effluent Monitoring, Monthly Check ,

74ST-9SQO2 RMS Surveillance Procedure, Gaseous Effluents i

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10CFR50.59 SCREENING AND EVALUATION RESPONSE JUSTIFICATICN l 1

PAGE 4 OF 5 ACTION UNDER REVIEWDCP 13-PJ-SQ-065 REVISION: 0 PCN:

Na:ne/ Title '

PROCEDURE /PCP/ TEMP MOD. NO .DCP 13-PJ-SQ-065 I

QUESTIONS; RESPONSE JUSTIFICATION 74RM-9EF41 Alarm Responses, RMS 74RM-9EF40 RMS Operating Procedures 74RM-9EF43 Actions forInoperable RMS Monitors 74RM-9EF42 RMS Alarm Serpoint Determination

, 2. This DCP will not change procedures as they are described in the UFSAR. The following !

2 UFSAR sections were reviewed in this regard and no changes found to be requirei 10.4.2.4 Tests and Inspections (CARS) 11.5.2.2.1 Sampling Equipment And Procedures 11.5.2.2.2 Analytical Procedures

. 13.5 Plant Procedures

3. This DCP does not involve tests or experiments not described in the UFSAR as defined in section 4.10 of 93AC-0NS01.

4. This DCP will require a change to Technical Specifications Figure 5.1-3 because it depicts a sepante release path for the CARS monitors. Note that the DCP is scheduled for implementation following the removal of discussion of these radiation monitors from the body of the Technical Specifications, per Generic letter 89-01. {

5. The probability of an accident previously evaluatedin the UFSAR will not be increased.

None of the radiation monitors involved in the change have any bearing on accidents described in UFSAR. The following UFSAR sections were myiewed in this regari 15.2 Dec=ase in Fleat Removal by the Seconda:y System

15.3 Loss of Condenser Vacuum The new piping will run in the same area as the old (the turbine building 176 ft. elevation level) and be supponed in the same manner, and failures ofit will create no conditions not previously analyzed and desc:ibed in the UFSAR. Tne pipe mounting has been designed to prevent st=ss on the pipe and provide adequate support Pipe or suppon failure would have a negligible effect on other plant equipment in the turbine building.

6. The consequences of an accident previously describedin UFSAR will not be increased.

None of the radiation monitors affected by this change provide engineered safety features or protection system acmation signals. The change in the manner of effluent exhaust and CARS effluent radiation monitoring has no affect on the ability of the monitoring system to pe: form its Tech Spec / UFSAR required function (s). The new piping will mn in the same area as the old (the turbine building 176 ft. elevation level) and be supponed in the same manner, and failures ofit will create no conditions not previously analyaed and

~ ** ~u .n *: *ti~ U itht:4 *f ,*CWCM 1 Y("* IW I 10CFR50.59 SCREENING AND EVALUATION RESPONSE JUSTIFICATION i

PAGE 5 or 5 ACTION UNDER REVIEWDCP 13-PJ-SQ-065 FIVISION: 0 PCN: ;

Name/ Title

PROCEDURE /?CP/ TEMP MOD. NO .DCP 13-PJ.SQ-065 i

i OUESTIONS RESPONSE JUSTIFICATION described in the UFS AR.

t

7. The probability of a malfunction of equipment important to safety will not be incmased.

The new piping will nm in the same area as the old (the turbine building 176 ft. elevation level) and be supponed in the same manner, and failures ofit will c= ate no conditions not previously analyzed and described in the UFSAR. The pipe mounting has been ;

designed to pmvent stmss on the pipe and provide adequate support. Pipe or support failum would have a negligible effect on other plant equipment in the turbine building.

8. The consequences of a malfunction of equipment imponant to safety will not be increased

None of the radiation monitors affected by this change provide engineered safety features !

or protection system actuation signals. The change in the manner of effluent exhaust and

(

CARS effluent radiation monitoring has no affect on the ability of the monitoring system ;

to perform its Tech Spec / UFSAR requimd function (s). Calculation 13-JC-SQ-211

}

shows that additional moisture from the CARS effluent will not affect the plant effluent ;

monitors. The new piping will run in the same area as the old (the turbine building 176 ft. elevation level) and be supported in the same manner, and failures of it will create no l i

conditions not previously analyzed and described in the UFSAR.

9. (

The possibility of an accident of a different type than previously evaluated in the UFSAR r will not be created because the new piping will present no different kind of threat than the !

previous piping. The piping is simply extended on the same floor and creates no potential hazard other than that almady evaluated in the UFSAR as a dec=ase in heat removal or i loss of condenser vacuum.

10.

The possibility of a malfunction of a different type than previously evaluated in the UFSAR will not be cmated. The malfunction reviewin the UFSAR considers loss of heat; removal and loss of condenser vacuum. These remain the only equipment malfunctions ,

that the new piping scheme could engender.

l 11.

The margin of safety as defined in the basis for any technical specification is not i

reduced.'Ihis modification involves the mmoval of one high range monitor in the condenser vent and relegation ofits role to the high range radiation monitorin the plant ;

vent. The ranges of the monitors are the same. Because the normal range monitors have ;

the ability to adequately detect radiation over five decades and these monitors will stay 2 in place, they have the ability to perform the anticipated radiation release detec: ion and activate filter systems as requimd to reduce further radiation. ;

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10CFR50.59 SCREENING AND EVALUATION , 1 ,, 2 ACTaON UhDER Rivd W: E E vtSiON PCN ors,4e he G k daden A%eal lobed) 7

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l SOCFRSO.59 SCREEN (Provide References on Response Justification Page) NO YES Does the proposed change:

1. Make changes in the facility as it is described in the UFSAR? X

2. Make changes in procedures as they are described in the UFSAR? _

3. Involve test or experiments not cescribed in the UFSAR? . _

4. Require a change to the technical specifications? _

Any answer to questions 1 through 3 "YES," then a 10CFR50.59 evaluation is required. Contact

! Document Control at ext. 82-6633 to obtain a tracking log number and enter the number in the Evaluation Log number block above. UFS AR Change Request per procedure 93AC-OLC01 may also be required.

Answer 4 is "YES."then Technical Specification Change Request per procedure 93AC-OLC01 and NRC approvalis required prior to implementation.

All answers 1 through 4 are 'NO." no 10CFR50.59 Evaluation required or Technical Specification change required, recommend action approval.

l 10CFRSO.59 EVALUATlDN(Provide Response Justification with References)

5. May the probability of an accident previously evaluated in the UFSAR be increased? - -

6 May the consecuences of an accident previously evaluated in the UFSAR be increased? _ _

7. May the probability of a malfunction of equipment important to safety be increased? - -

I B. May the consequences of a malfunction of equipment important to safety be increased? - - -

9. May the possibility of an accident of a different type than any previously evaluated in the UFSAR be created"> .

10. May the possibility of a different type of malfunction than any previously evaluated in the UFSAR be created? _ _

l 11. Is the margin of safety as defined in the basis for any technical specification reduced? _ _

Any answer to questions 5 through 11"YES."then an unreviewed safety question is identified. Proceed to procedure 93AC-OLC03 prior to implementation.

All answers 5 through 11 are "NO," there is no unreviewed safety question and action approval is recommended.

If UFSAR Chapter 6/ Chapter 15 is potentially affected, forward a copy of evaluation to Nuclear Fuels [

Management.

l I verify that the above screenanglevaluation is adequate and accurate and that the undersigned has received required training. I

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l 10CFRf0.59 RD'IDV AND EVALUATION l

RESPONSE JUSTIFICATION r.. 2 a z ACTION UNOER REVIEiis(NAME/TITLI) REVISIDN PCN OFFSITE DOSE CANTION MANUAL, EECTION 6.0 CEANCES WHICH AFFECT THE RADIDIDCICAL ENVIROKKENTAL MONITORING PRCCPJut 7 N/A PROCEOUR.E/PCP/ TEMPORARY MODIFICA!!ON N3r RfA QUESTION RESPONSE JUSTIFICATION l 1. The Radiological Environmental Monitoring Program (REMP) is not specifically described !

I in the UFSAR. The REMP is described in the ER-OL, ODCM and NUREG-1301. The l REMP is required to be implemented via the ODCM as required by Section 6 of the i

technical specifications. ODCM changes are allowed as long as the changes are made and l

reported as required. The proposed changes affect the REMP sampling locations and will not make changes to the facility as it is described in the UFSAR.

References:

ER-OL 6.1.5 UFSAR 3.1.55,11.5,12.3,13.1, Rev. 5 Technical Specification 6.8.4h, Amendments- 69 (U-1), 55 (U-2), 42 (U-3)

USNRC Regulatory Guide 4.1, Rev.1,1975 NUREG-1301, April 1991 i

2. The REMP is not specifically described in the UFSAR. Nuclear Administrative and '

Technical Manual procedures implement the ODCM, but the procedures are not described in the UFSAR. Consequently, no changes will be made in procedures as they are described in the UFSAR.

References:

see no.1 above

3. Changes in REMP sample locations does not involve any tests or experiments and, therefore, would not involve tests or experiments not described in the UFSAR. r

References:

see no. I above

4. The Technical Speci5 cations requires that a radiological environmental monitoring program ;

be implemented in accordance with the methodology and parameters in the ODCM. The 3 intent is to monitor all potential pathways for dose to man due to plant effluents. The addition of, or changes to, sampling locations will enhance the monitoring program but will not require that a technical specification change be made.

References:

see no. I above i

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i TECHNICAL SPECIFICATION REFERENCE A. Periodic Review and/or Revision Requirements:

Technical Specification, Section 6.8.4.g and Section 6.8.4.h have been reviewed. The program elements required to be contained in the ODCM are present in this review / revision of the ODCM.

ODCM Revision No. 7 Initiator Name (printed) Kevin Kutner Signature Vlf Date 5-134 3 Technical Reviewer IJ/ #/M Date 5-M-f3 l 9 B. Additional Revision Requirements:

This ODCM revision submittal contains;

1. Sufficient iniormation to support the change together with the appropriate analyses or evaluations justifying the change (s) (RCTS 011072-01) and;

2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106,40 CFR Part 190,10 CFR 50.36a, and Appendix i to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. (RCTS 011050-01).

l 3. Each change shall be identified by markings in the margin of the affected pages, clearly l indicating the area if the page that was changed, and sha!! indicate the date(e.g., month / year) !

l the change was implemented.

Initiator // #[f Date S 1i't)

Technical ReviewerIfl MfM Date C-w.ov V j l

1 i

1

1 1

i TECHNICAL SPECIFICATION REFERENCE )

1 i

A. Periodic Review and/or Revision Requirements:

Technical Specification , Section 6.8.4g and Section 6.8.4h have been reviewed. The program elements required to be contained in the ODCM are present in this review! revision of the ODCM.

ODCM Revision No. 7 l

Initiator Name (printed) Louis Drinovskv Signature Date 6"80-f3 M, .

Technical Reviewer b). 9 7/ d _ Date 8 - S e,r- 9 7 l

/ ;

B. Additional Revision Requirements.

This ODCM revision submittal contains; i 1. Sufficient information to support the change together with the appropriate analyses or j evaluations justifying the change (s) (RCTS 011072-01) and; -

P

2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106,40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR 50 }

and not adverselyimpact the accuracy or reliability of effluent, dose, or setpoint calculations.

(RCTS 011050-01)

3. Each change shall be identified by markings in the margin of the affected pages, clearly i indicating the area if the page that was changed , and shall indicate the date (e.g.,

month / year) the change was implemented. i Initiator

/

er Date 980"N f ,

Technical Reviewer T R/ MM/ Date A~ Sv- 9 ? [

07 ;

i 1

ML20056G011

Text

3.2 Requirements

REFERENCES:

2.1 Requirements

3.1 Requirements

3.2 Requirements

4.1 Requirements

4.2 Requirements

4.4 Requirements

5.1 Requirements

6.1 Requirement

7.1 Requirement

7.2 Requirement

1.0 INTRODUCTION

2.1 Requirements

3.1 Requirements

Reference:

Reference:

3.2 Requirements

4.1 Requirements

4.2 Requirement

4.3 Requirements

References:

References:

References:

4.4 Requirements

5.1 Requirement

6.1 Requirements

6.2 Requirement

6.3 Requirements

7.1 Requirement

7.2 Requirement

Reference:

References:

References:

References:

References:

References:

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