Text

CY-OC-170-301, Revision 4, Offsite Dose Calculation Manual
	ML101270345
Person / Time
Site:	Oyster Creek
Issue date:	08/19/2009
From:	Oltmans D; Exelon Nuclear
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	AD-AA-101-F-01, Rev 1, CY-AA-170-3100, RA-10-002 CY-OC-170-301, Rev 4
	Download: ML101270345 (201)
	v • d • e

Procedure Approval Form AD-AA-101-F-01 Page 1 of 1 Revision 1 Document Number: L C - ) 70 -301 Revision:

Title, O6 FS'it' Po c , 1 c, -41. ,,.. P.M-tvc.. I FO O C (2ý5 C] New C1 Cancel [] Cancel **Revision EC#: PCR#: PPIS#I Document Revision 03 Editorial

[' Batch ER#: AR#: #:

M IC 7C7-3 13 o Supersede corporate document(s) List: PJ / .

Revision Summary:

Attach addlidescript, if req'd A.r0tvu.n.D..err Impact on Operating I* N/A and Design Margins:

Attach add') descript. if req'd CONFIRM that W commitments (i.e., those steps annotated with CM-X) have been changed or deleted uQless evaluated via completion of LS-AA-1 10 commitment change/deletion form and INITIAL [Preparer]:

Preparer: -- s /e Oi_ / 0'?

FC)1,7o-'s O che" 1.-5 e Print Date Location/Ext Applicable 13R D DR 0

QCt _

Site Contacts BY 03 LA ["0 . CL [3 Check box and PB D OcI F R. r9t-A LG 3 provide name TMiW ZNr-I Other r-_

Validation Req'd: 0 No £3 Yes (attach) pIn / mon Training Reqd: 0 No 0 Yes (Validation requirement see AD-AA-101) Print/Signature Site Specific Training Req'd: M* No £3 Yes Change Management: C] HU-AA-1101 Change Checklist Attached [] Document Traveler E] None Required Level of Use: £3 Level 1 - Continuous Use -' Level 2 - Reference Use 91 Level 3 - Information Use Approval CFAM (Standard Procedures) Print/Sign Date Location/Ex:

Approval Site Document(s) to be superseded: A) IA Location: 5 f C fce..( IUse additional sheets as necessary. Assure that all pending changes are di6positioned.

[I Temp. Change £3 Interim Change Temp or Interim Change #: Interim Change expiration:

10CFR50.59 Applicable: M No [£ Yes Tracking Number:_

10CFR72.48 Applicable; N] No[3 Yes Other Regulatory Process Applicable: [£ No U Yes Other Regulatory Process Number. C y- /1 * -070 -3/Op PORC Required: £] No M] Yes PORC Number (after PORC Approved):

Environmental Review Required: [] No 00 Yes If "Yes" then attach completed EN-AA-103 Attachment 1.

£3 Itsuperceding a document containing commitments, notify the Commitment Tracking Coordinatot per LS-M-110 so the CTO can be updated as appropliate.

Cross Discipline Reviews: (list below) Survet c C ordi ator Review Feq'd C No Q Yes, list below r_ Date D-pIlnoor Org 7A..4; at t~S~

eorr I

Print Signature Data Diacipline or Org.

Attach ad*itional f req'd Temp Change Authorization Only SRO Print/Sign/Date SOR Print/Sign/Date Impl. Date Exp. Date SOR Approval indicates that all required Cross-Disciplinary reviews have been performed and the reviewers have signed this form. This procedure is technically and functionally accurate for all functional areas. (Refer to - .2)

SOR Approval: -

,n" t/Siqr ,

A 1. / I I I O i ci o Or g Site Authorization: 9 Manage r S7PriAMrin Sig prpl.'Dt a 7 e Plant Manager Prinit/Sign (w en require bV TW cdur

PORC ITEM

SUMMARY

COVER SHEET

SUBJECT:

Offqite rln,- CaIrlcultinn Manial Revisinn 4 PREPARED BY: Dennis Oltmans IS A 50.59172.48 SAFETY EVALUATION REQUIRED? [] Yes X ] No ODCM Change Determination performed per CY-AA-170-3100, Offsite Dose Calculation Manual Revisions ISSUE

SUMMARY

A Change Matrix is provided.

The changes to the ODCM are to either conform to NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Boiling Water Reactors, or provide justification in the Bases for deviations from NUREG-1302.

There are changes to the ODCM to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. These ensure that liquid and gaseous treatment systems are used to reduce radioactive releases to less than 2% of annual release limits.

The administrative changes are also identified in the Change Matrix.

SAFETY IMPACT: There is no impact to the accuracy or reliability of effluent, dose or setpoint calculations.

The level of radioactive effluent control is maintained as required by 10CFR20.1301, 10CFR20.1302, 40CFR1 90, 10CFR50.36a, 10CFR72.104, Appendix I to 10CFR50 and the SAR.

The Presentation Material is Ready for PORC Review Dpnnim Oltman Q /

A #- 91231*00Q Presenter's Signature Date

.ftansiKandasaa - 21-.lenAAC19-N 9/23/2009 Supervisor's Sign ure Date

EN-AA-103 Revision 3 Page 6 of 8 Attachment 1 SAMPLE - Environmental Review Checklist Page 1 of 2 Exelkn,. EnvironmentalReview Checklist NucleaT CJ7 0 - Rev. No. L4 Station/Unit o -

(~eej(

Summarize Desc'ription of Proposed Activity:

I Doc. No.

I YES responses require Environmental Evaluation by Site Environmental Personnel per EN-AA-103-0001.

following?

Does the proposed activity involve any of the for any system listed in Attachment 2 IF1 Yes W No I Configuration or operational changes e

2. Removal from service of instrument air or electrical components that energize or otherwise affect on, Priority 1 or 2 environmental equipment listed in EN-AA-103-00020YeNo 31 Increased noise levels at the site property boundary from the installation of new permanent LI YesailiAo EeYent LI Yes 1 No

4. equip Purchase, lease or sale of any land or property Yes No

5. Increase in the maximum reactor power level

6. Affect operations of fish hatcheries, recreational areas, parks, or other public domains Yes 9 No Technical

7. Non-editorial change to the Environmental Protection Plan or Environmental 0 No Specifications, NRC Operating License, Appendix B F1 Yes

8. Oil-filled Iransformers or oil circuit breakers > 55 gallons in volume Yes N No

9. Equipment containing PCBs INo Yes

10. Station equipment that bums fossil fuels, i.e. internal combustion engines (gasoline or diesel), L ] Yes No boilers, gas turbines, furnaces, heating sources, etc.

11. Chillers or refrigeration equipment containing over 50 Ibs of refrigerant [*Yes No

12. Halon ytm Yes No

13. Instruments on the meteorological tower :) Yes [] No Yes No Vents or exhaust systems designed to exhaust vapors, fumes, mists, internal combustion

14. engi

15. Tanks or tank vents that contain a chemical or fuel []Yes [] No Yes No

16. Cooling towers or cooling lakes -1Yes *No.

17. Floor drains and plumbing

18. Sandlsigeupet - Yes RINo 19, Painting operations that use more than 5,000 gallons of paint (including thinner) per year ; Yes No

20. Storm drains, ditches or swales . Yes *'No

21. Oil seprtrolitretrges rp- Yes No 22.

23.

Dikes, dams or appurtenances (i.e. equipment or structures required for operation)

Construction, demolition or abandon in place of any site buildings or structures H Yes Yes Yes No No No

24. Permanent or temporary storage areas for hazardous or regulated wastes

25. Chemical or oil containment or oil-filled transformer fire rock containment Yes No

26. Dredging or removing silt from intake structures Yes No

EN-AA-103 Revision 3 Page 7 of 8 a Attachment I SAMPLE - Environmental Review Checklist Page 2 of 2 Exelkn. EnvironmentalReview Checklist Nuclear

27. Corrective maintenance WRANO or change / deletion of preventative maintenance for Priority [3 Yes *J No, 1 or 2 environmental equipment listed in EN-AA-103-0002

28. Management of habitat, wildlife or vegetation (other than landscaping) on the site property, i.e. El Yes g] No tra in .hunting, extermination, fish electroshock, etc.

29. Temporary / portable equipment containing internal combustion engines and fuel tanks 0I"Yes 0 No

> 55 allons

30. Land disturbance of >1 acre (e.g. excavation, tilling, clearing away top layer of dirt for construction of building, etc.), well drilling, soil boring or change to storm water runoff, i.e. -- Yes [ No more or less pavement

31. Paving of previously unpaved areas or chipping / demolition of old pavement Yes No to protect soil or

32. Work near waterways or storm drains or work not on an impervious surface, No surface water, as described in EN-AA-103-0003 E Yes

33. Friable asbestos work (non-friable gaskets, packing, etc. excluded) E] Yes *] No

34. Changes to process plant chemicals or concentrations and flow rates ot existing chemicals 1'Yes

)) No

35. Increase in surface water or groundwater withdrawal or increased withdrawal pump run times Yes No Yes No

36. Increase in fossil fuel usage on site

37. Changes to the amount of water or effluent location discharging to the environment

.J Yes No licensed material to reachR Yes R No

38. Changes that could create a new credible mechanism for Yes_ No groundwater

39. Pesticide / herbicide application _- r_

__Yes _No

40. Open burning of wood, brush, weeds, oil, fossil fuels, propane, etc. Yes XNo

41. Maintenance on domestic water, potable water or well water systems :- Yes .*No

42. Affects to Significant Environmental Aspects (SEA) Yes X No Prepared By:

Print Name: -D- 'n 'S if Signature: , ýae /Ie C/O For any boxes checked "YES":

Environmental Personnel Contacted: Date / Time:

D Copy of completed Attachment 1 provided to Environmental Personnel contacted.

Comments:

-EN-AA-103 Revision 3 Page 8 of 8 ATTACHMENT 2 SYSTEMS WHERE CONFIGURATION OR OPERATIONAL CHANGES MAY REQUIRE ENVIRONMENTAL EVALUATIONS Page 1 of 1 Acid Feed & Handling .' Laundry Equipment Radwaste Reprocessing.,'

Auxiliary Building Floor Draine< and Disposal Auxiliary Steam Boilers-' Makeup Dermineralizer I Pretreatment Systemr Carbon Dioxide/ Miscellaneous Drains Caustic Handling, Miscellaneous Outside Equipment' Chemical Feed & Handling / Chlorination /"/ Oil Drain Disposal -

Hypochlorite Primary Containment Purge / Nitrogen Chemical Radwaste Disposal' Process Sampling (For NPDES only) /

Closed Cooling Water Chromate Addition I Radwaste Floor Drains /

Removal (PWR) / Reactor Floor Drains and Sumps/

Circulating Water / Refrigeration Piping / HVAC I Coolers I Chillers-Diesel Generator/ Screen Wash -'

Diesel Oil / Fuel Oil (including associated piping Service Water,'

and storage tanks)-' o Emergency'.

Dilution Water System (Oyster Creek)-' o Essential' Domestic Water / o Non-Essential.,-

Drains - Station Heating Cond. / HVAC / Air o High Pressure,-`

Washers - Sewage Treatment / Tie-in Radiation Monitor/

Fire Protection Systems - Solid Radwaste Reprocessing and Disposal,'

Fuel Storage (Refers to fossil fuels only)' Standby LiquidControl (SBLCY Grounding and Cathodic Protection,- Station Heating / Boilers / Evaporators '

Halon,- Switchyard' Heat Exchangers (Raw Water Cooled," Transformers (Oil-filled only)-'

Hydrogen,-, Treated Water, Hydrogen Water Chemistry " Turbine Building Floor Drains and sumps, Independent Spent Fuel Storage Installation" Turbine Electro-Hydraulic Control (EHC)'

(ISFSI) Turbine Oil / Turbine Dirty Oil Tank ,

Lake Makeup or Blowdown" Wastewater Treatment/.

Laundry and Floor Drains" Well Water -"

Ofmans, Dennis:(GenCo-Nuc)

FVom: Jordan, Francis:(GenCo-Nuc)

Sent: Wednesday, September 23, 2009 10:13 AM To: Oltmans, Dennis:(GenCo-Nuc)

Cc: Kandasamy, Jhansi R.:(GenCo-Nuc); Greiner, David:(GenCo-Nuc)

Subject:

ODCM

Dennis, I reviewed the ODCM, the proposed changes I have no comments on.

However, from a ODCM content, entire document content standpoint I had question about the following:

Liquid effluents: The overboard discharge methodology and calculations are still present in the document, however, there is no controls defining minimium dilution flow. I know in the calculation you would get a zero allowable release without dilution but what bothers me about the calculation. Without knowing better some one could assume the dilution flow as creek.

only SW flow or about 6000 gpm and I believe this is inadequate based on previous OE from Oyster Specifically the older version required at least one circulating water pump or equivelent to be in service during liquid discharge to prevent exceeding concentrations at the route 9 bridge. For example in the early 90's the station crossconnected a contaminated system with Service water systems and some large volume of hotwell water/CST was discharged directly to the discharge canal. The only thing running was the SW system. When we sampled the discharge canal some nuclides were above 10cfr2O limits because of not mixing. I believe there should be a control on minimum dilution flow during a discharge.

MET tower:

Met tower instrumentation section does not include joint data recovery requirements for the year. I.e. 90% ANSI requirement the current specification as written does not ensure compliance or proper compensatory action for non-compliance with meeting data recovery.

Chip Jordan 1

ODCM Change Determination Station: Oyster Creek Generating Station Page 1 of 15 ODCM Revision No. 4 Determination Identifier T I. Determination Questions (Check correct response)

1. Does the ODCM change maintain the level of radioactive X YES _NO effluent control required by 10CFR20.1301?

Explain:

10 CFR 20.1301 establishes the dose limits for individual members of the public. The current limit is 100 mrem in a year. In addition 10 CFR 20.1301.d establishes that a licensee must meet the requirements of 40 CFR 190. 40 CFR 190 establishes annual limits to a real individual of 25 mrem to the total body or any organ other than thyroid and 75 mrem to the thyroid.

The changes marked with an "A"in the "Change Type" column are administrative in nature and therefore have no negative impact on exceeding these limits.

items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. The changes in item 17 are consistent with 40 CFR 190. Therefore, there will be no change to the level of effluent control required by 10CFR20.1301.

Items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6.

6.8.4. a.5. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

6.8.4. a.6. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in the 31 day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR 50, Since the ODCM changes are the same as the existing Technical Specifications, there will be no change to the level of radioactive effluent control required by 10CFR20.1301.

items 18,19 and 21 I

These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1302.

Since there are no changes to current practices, the level of radioactive effluent control required by 10CFR20.1301.

2. Does the ODCM change maintain the level of radioactive X YES _NO effluent control required by 10CFR20.1302?

Explain:

10CFR20.1302 compliance with dose limits for individual members of the public requires that surveys of radiation levels in unrestricted and controlled areas and radioactive materials in effluents released to unrestricted and controlled areas to demonstrate compliance with the dose limits for individual members of the public in

§20.1301. The licensee shall show compliance to the annual dose limit in §20.1301 by demonstrating by measurement or calculation that the total effective dose equivalent to the individual likely to receive the highest dose.. .does not exceed the annual dose limit; or demonstrate that the annual average concentrations of radioactive material released in gaseous and liquid effluents at the boundary of the unrestricted area do not exceed the values specified in Table 2 of Appendix B to Part 20; and if an individual were continuously present in an unrestricted area, the dose from external sources would not exceed 0.002 rem (0.02mSv) in an hour and 0.05 rem (0.5mSv) in a year.

The changes marked with an "A"in the "Change Type" column are administrative in nature and therefore have no negative impact on the requirements of 10CFR20.1302:

Items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, there will be no change to the level of effluent control required by 10CFR20.1302.

Items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

Since the ODCM changes are the same as the existing Technical Specifications, there will be no change to the level of radioactive effluent control required by 10CFR20.1302.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1302.

Since there are no changes to current practices, the level of radioactive effluent control required by 10CFR20.1302.

2

3. Does the ODCM change maintain the level of radioactive X YES _NO effluent control required by 40CFR190 and 10CFR72.104?

Explain:

40CFR1 90 requires the annual dose equivalent not exceed 25 millirems to the whole body, and 75 millirems to the thyroid, and 25 millirems to any other organ of any member of the public as the result of exposures to planned discharges of radioactive materials, radon and its daughters excepted, to the general environment from uranium fuel cycle operations and to radiation from these operations.

10CFR72.104 has the same limits with the following requirements: Operational restrictions must be established to meet as low as is reasonably achievable objectives for radioactive materials in effluents and direct radiation levels associated with ISFSI and Operational limits must be established for radioactive materials in effluents and direct radiation levels associated with ISFSI.

The changes marked with an "A" in the "Change Type" column are administrative in nature and therefore have no negative impact on requirements of 40CFR190 or 10CFR72.104.

Items 5,6 and "17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, there will be no change to the level of effluent control required by 40CFR190 or 10CFR72.104.

'Items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

Since the ODCM changes are the same as the existing Technical Specifications, there will be no change to the level of radioactive effluent control required by 40CFR1 90 or 10CFR72.104.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences toNUREG-1 302.

Since there are no changes to current practices, the level of radioactive effluent control required by 40CFR190 or 10CFR72.104.

4. Does the ODCM change maintain the level of radioactive X YES _NO effluent control required by 10CFR50.36a?

Explain:

10CFR50.36a requires that nuclear power reactors keep releases of radioactive materials to unrestricted areas during normal conditions, including expected occurrences, as low as is reasonably achievable by establishing operating procedures developed pursuant to §50.34a(c) for the control of effluents and that the radioactive 3

waste system, pursuant to §50.34a, be maintained and used .... (b) in establishing and implementing the operating procedures described in paragraph (a) of this section, the licensee shall be guided by the following considerations: Experience with the design, construction, and operation of nuclear power reactors indicates that compliance with the technical specifications described in this section will keep average annual releases of radioactive material in effluents and their resultant committed effective dose equivalents at small percentages of the dose limits specified in §20.1301 and in the license. At the same time, the licensee is permitted the flexibility of operation, compatible with considerations of health and safety... It is expected that in using this flexibility under unusual conditions, the licensee will exert its best efforts to keep levels of radioactive material in effluents as low as is reasonably achievable.

The changes marked with an "A" in the "Change Type" column are administrative in nature and therefore have no negative impact on the requirements of 10CFR50.36a.

Items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, there will be no change to the level of effluent control required by 10CFR50.36a.

items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

Since the ODCM changes are the same as the existing Technical Specifications, there will be no change to the level of radioactive effluent control required by 10CFR50.36a.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1302.

Since there are no changes to current practices, the level of radioactive effluent control required by 10CFR50.36a.

5. Does the ODCM change maintain the level of radioactive X YES _NO effluent control required by Appendix I to 10CFR50?

Explain:

Appendix I to 10CFR50 requires that A. The calculated annual total quantity of all radioactive material above background to be released from each light-water-cooled nuclear power reactor to unrestricted areas will not result in an estimated annual dose or dose commitment from liquid effluents for any individual in an unrestricted area from all pathways of exposure in excess of 3 millirems to the total body or 10 millirems to any organ.

B.

4

1. The calculated annual total quantity of all radioactive material above background to be released from each light-water-cooled nuclear power reactor to the atmosphere will not result in an estimated annual air dose from gaseous effluents at any location near ground level which could be occupied by individuals in unrestricted areas in excess of 10 millirads for gamma radiation or 20 millirads for beta radiation.

2. Not withstanding the guidance of paragraph B.A: (a) The Commission may specify, as guidance on design objectives, a lower quantity of 'adioactive material above background to be released to the atmosphere if it appears that the use of the design objectives in paragraph B.1 is likely to result in an estimated annual external dose from gaseous effluents to any individual in an unrestricted area in excess of 5 millirems to the total body; and (b) Design objectives based upon a higher quantity of radioactive material above background to be released to the atmosphere than the quantity specified in paragraph B.1 will be deemed to meet the requirements for keeping levels of radioactive material in gaseous effluents as low as is reasonably achievable if the applicant provides reasonable assurance that the proposed higher quantity will not result in an estimated annual external dose from gaseous effluents to any individual in unrestricted areas in excess of 5 millirems to the total body or 15 millirems to the skin.

C. The calculated annual total quantity of all radioactive iodine and radioactive material in particulate form above background to be released from each light-water-cooled nuclear power reactor in effluents to the atmosphere will not result in an estimated annual dose or dose commitment from such radioactive iodine and radioactive material in particulate form for any individual in an unrestricted area from all pathways of exposure in excess of 15 millirems to any organ.

The changes marked with an "A" in the "Change Type" column are administrative in nature and therefore have no negative impact on exceeding these limits.

Items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, there will be no change to the level of effluent control required by Appendix I to 10CFR50.

Items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

Since the ODCM changes are the same as the existing Technical Specifications, there will be no change to the level of radioactive effluent control required by Appendix I to 10CFR50.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1302.

Since there are no changes to current practices, the level of radioactive effluent control required by Appendix I to 10CFR50.

5

6. Does the ODCM change maintain the accuracy or reliability of X._._YES _NO effluent, dose, or setpoint calculations?

Explain:

The changes marked with an "A" in the "Change Type" column are administrative in nature and therefore have no negative impact on accuracy or reliability of effluent dose or setpoints.

Items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, these changes have no negative impact on accuracy or reliability of effluent dose or setpoints.

items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

Since the ODCM changes are the same as the existing Technical Specifications, the changes will have no negative impact on accuracy or reliability of effluent dose or setpoints.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1 302.

Since there are no changes to current practices, there will be no negative impact on accuracy or reliability of effluent dose or setpoints.

7. Does the ODCM change maintain the accuracy of radioactive ._ X YES -NO effluent control required by the SAR?

Explain:

The changes marked with an "A"in the "Change Type" column are administrative in nature and therefore have no negative impact on the accuracy of radioactive effluent control required by SAR.

Items 5,6 and 17 These changes to the ODCM are consistent with NUREG-1302 and are consistent with the practices at Oyster Creek. Therefore, these changes have no negative impact on the accuracy of radioactive effluent control required by SAR.

Items 8,9,13,14,16,23 and 25 These changes to the ODCM are to implement Technical Specification 6.8.4.a.5 and 6.8.4.a.6. See Question 1 Explain: for Technical Specifications 6.8.4.a.5 and 6.8.4.a.6 requirements.

6

Since the ODCM changes are the same as the existing Technical Specifications, the changes will have no negative impact on the accuracy of radioactive effluent control required by SAR.

Items 18,19 and 21 These changes to the ODCM BASES describe the current practices at Oyster Creek.

The changes to the BASES were to justify differences to NUREG-1302.

Since there are no changes to current practices, there will be no no negative impact on the accuracy of radioactive effluent control required by SAR.

II.If all questions are answered YES, then complete the ODCM Change Determination and implement the Change per this procedure.

IlI. It any question is answered NO, then a change to the ODCM is not permitted WV. Signoffs:

See AD-AA-101-F-01 7

Change Matrix for CY-OC-170-301 (ODCM) Revision 4 Change type - A Administrative Changes Change type - T TechnicaM Changes Rev.

item Rev. 3 4 Change No. Page Page Description of Change and Reason for Change Type Change header from Revision 3 to Revision 4 and footer from All All 040809 to 092909 for consistency A Added 1.5.6 PROJECTED DOSE - LIQUID for new section in 2 4 4 Calculation Methodologies A Added 2.6 PROJECTED DOSE - GASEOUS for new section in 3 6 6 Calculation Methodologies A 3.3.3.11 ACTION c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

Was changed to ACTION c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable. Report all deviations in the Radioactive Effluent Release Report.

The requirement to report all deviations in the Radiological Effluent Release Report was added to be consistent with 4 23 23 NUREG-1 302 A Added New Radwaste and AOG to the Service water Effluent section of Table 4.11.1.1.1-1, Radioactive Liquid Waste Sampling and Analysis Program. This was added to be consistent with NUREG-1302 and current Oyster Creek 5 31 31 practices T 8

NOTATION i. was changed to:

f. In the event a grab sample contains more than 5E-7 pCi/mL of 1-131 and principal gamma emitters or in the event the Reactor Building Service Water radioactivity monitor indicates more than 5E-7 pCi/mL radioactivity in the effluent, as applicable, sample the elevated activity effluent daily until analysis confirms the activity concentration in the effluent does not exceed 5E-7 pCi/mL. In addition a composite sample must be made up for further analysis for all samples taken when the activity was > 5E-7 pCi/mL.

Note f was changed to reflect the changes to Table 4.11.1.1.1-1, Radioactive Liquid Waste Sampling and Analysis Program, B. New Radwaste and AOG Service water Effluents were added to the table. Since New Radwaste and AOG Service water streams do not have radiation monitors, the note was modified to reflect these differences.

This was changed to be consistent with NUREG-1302 and current Oyster Creek practices.

6, 33 33 T In 3.11.1.2, (see Figure E-3) was changed to (see Figure E-4). Figure E-4 is the correct title for Area Plot Plan of Site, Site Map Defining UNRESTRICTED areas and SITE L-7-_--- .. 34 ---- BOUNDARY-for-Radioactive.-Gaseous and-Liquid Effluents. A 9

3.11.1.3 In accordance with the Oyster Creek Technical Specifications 6.8.4.a.6, the liquid radwaste treatment system shall be OPERABLE and appropriate portions of the system shall be used to reduce the radieactivo matc-ial; on liquid wastes pFrio tthir,4,d*,*,,g*

th when !hc radi*otiVitr Concflt~tiflexclucive Gf tritium and d-ircseolc ologc in !he batch oXceode 0.001 p~i4rL.

was changed to 3.11.1.3 In accordance with the Oyster Creek Technical Specifications 6.8.4.a.6, the liquid radwaste treatment system shall be OPERABLE and appropriate portions of the system shall be used to reduce the releases of radioactivity when projected doses due to the liquid effluent to UNRESTRICTED AREAS (see Figure E-4) would exceed 0.06 mrem to the Total Body or 0.2 mrem to any organ in a 31 day period.

The limits are equal to 2% of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR 50 as required by Technical Specification 6.8.4.a.6.

8 35 35 The limit are consistent with NUREG 1302. T 4.11.1.3.1 was changed as follows:

Doses due to liquid releases to UNRESTRICTED AREAS shall be dete-mie projected at least once per 31 days in accordance with the methodology and parameters in the ODCM Part II Section 1.5 in accordance with Technical Specifications 6.8.4.a.5.

The change was made to conform to Technical Specification 9 35 35 6.8.4.a.5 T 4.11.1.3.2 The installed liquid radwaste treatment system shall be demonstrated OPERABLE by meeting CONTROLS 3.11.1.1,3.11.1.2, and 3.11.1.4.

was changed to The installed liquid radwaste treatment system shall be demonstrated OPERABLE by meeting CONTROLS 3.11.1.1, 3.11.1.2, and 3.11.1.3.

This administrative change was made because 4.11.1.4 does not exist and 4.11.1.3 is the correct reference.

10 35 35 A 10

Deleted 4.11.2.1.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be determined at least once per 31 days in accordance with the methodology and parameters in the ODCM Part II Section 2.4.1 in accordance with Technical Specification 6.8.4.a.5.

This is the same requirement as 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM Part II Section 2.4.1 at least once per 31 days in accordance with Technical Specification 37 36 6.8.4.a.5. and is consistent with NUREG-1302. A Table 4.11.2.1.1-1, Radioactive Gaseous Waste Sampling and Analysis Program, was changed to ensure that tritium is required to be sampled and analyzed in the Stack, Turbine Building Exhaust Vents, Augmented Off Gas Building Vent to be consistent with NUREG-1302. This is consistent with the 12 38 37 existing sampling and analysis procedures.

3.11.2.4 In accordance with Oyster Creek Technical Specifications 6.8.4.a.6, the AUGMENTED OFF GAS SYSTEM shall be in operation.

was changed to 3.11.2.4 The AUGMENTED OFF GAS SYSTEM shall be in operation to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY (see Figure E-4) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any body organ The limits from are equal to 2% of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR 50 as required by Technical Specification 6.8.4.a.6.

13 43 42 This wording is also consistent with NUREG-1302. T II

3.11.2.4, ACTION: b was changed as follows:

With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 30 consecutive days and either CONTROL 3.11.2.1 or 3.11.2.2 4 exceeded, prepare and submit to the Commission within 30 days from the end of the quarter during which release occurred, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:

This changes the applicable dose limits from quarterly or annual limits (CONTROL 3.11.2) to 2% of the annual dose limits (CONTROLS 3.11.4) to be consistent with Technical 14 43 42 Specifications 6.8.4.a.6. T 4.11.2.4 changed to 4.11.2.4.1. This was an administrative 15 43 42 change to allow addition of a Surveillance Requirement A Added 4.11.2.4.2 Doses due to gaseous releases to UNRESTRICTED AREAS shall be determined at least once per 31 days in accordance with the methodology and parameters in the ODCM Part IISection 1.5 in accordance with Technical Specifications 6.8.4.a.5.

16 43 42 This requirement is consistent with NUREG-1302. T 3.11.4.1 In accordance with Oyster Creek Technical Specifications 6.B.4.a.10, the annual (calendar year) dose commitment to any MEMBER OF THE PUBLIC due to radioactive material in the effluent and direct radiation from the OCGS in the UNRESTRICTED AREA shall be limited to less than or equal to 75 mrem to the thyroid or less than or equal to 25 mrem to any other organ.

was changed to:

In accordance with Oyster Creek Technical Specifications 6.8.4.a.10, the annual (calendar year) dose commitment to any MEMBER OF THE PUBLIC due to radioactive material in the effluent and direct radiation from the OCGS in the UNRESTRICTED AREA shall be limited to less than or equal to 75 mrem to the thyroid or less than or equal to 25 mrem to the total body or any other organ.

17 45 44 This will make the limits the same as NUREG-1302. T 12

The following was added to BASES 3/4.11.1.1 Liquid, Concentration.

Weekly grab samples for Service Water Effluents are composited for monthly tritium and gross alpha analysis and quarterly Sr-89,90 and Fe-55 analysis if activity is detected. New Radwaste and AOG Service Water Effluents have no continuous radiation monitors. Weekly grab samples are performed to detect leaks. Plant design does not include composite samplers, so the weekly grab sample frequency is considered adequate.

These changes were added to justify the current practices at Oyster Creek. NUREG-1 302 assumes that continuous releases have in-line composite samplers.

Oyster Creek does not have composite samplers. Grab samples have been historically used to makeup 18 65 64 composite samples. T The following was added to BASES 3/4.11.1.1 Liquid, Concentration.

Circulating Water Effluent is not included in Table 4.1 1.1.1=,-. Radioactive Liquid Waste Sampling and Analysis Program, since the Circulating Water is sampled as a part of the Radiological Environmental Monitoring Program, Table 3.12.1-1, 3a, Waterborne, Surface downstream sample.

This change was to justify the current practice at Oyster Creek. The circ water is not currently sampled at the outlet of the condenser. The discharge canal flow is mostly circ water and should be sufficient for detecting leaks into the circ water system. This question came after the contamination of the circ water by flushing of 19 65 64 the condenser with condensate water, not demin water. T 13

The following was added to Bases 3/4.11.1.3, Liquid Effluent, Liquid Radwaste Treatment.

Figure U3-1 -1 a, Liquid Radwaste Treatment Chem Waste and Floor Drain System and Figure D-1-11b, Liquid Radwaste Treatment - High Purity and Equipment Drain System provide details of the Liquid Radwaste Treatment system.

20 66 65 This administrative change was to add information. A The following was added to Bases 3/4.11.2.1, Gaseous Effluents, Dose Rate.

Tritium is sampled quarterly for gaseous effluents.

Based on the consistency of the data from the quarterly sampling, the sampling frequency is adequate.

This change was to justify the current practices at Oyster Creek. NUREG-1302 has tritium performed monthly. The total tritium released in gaseous form was very consistent quarter to quarter for the last seven years. Monthly sampling would not increase the accuracy of the total tritium amount 21 66 66 released. T The following was added to Bases 3/4.11.2.4, Augmented Off Gas Treatment System Gaseous Effluents, Augmented Off Gas System.

Figure D-2-1, Gaseous Radwaste Treatment -

Augmented Off gas System, Figure D-2-2, Ventilation System provide details of the Augmented Off Gas Treatment System and Figure D-2-3, AOG Ventilation System.

22 67 67 This administrative change was to add information. A

Added 1.5.6 PROJECTED DOSE - LIQUID The projected doses in a 31 day period are equal to the calculated doses from the current 31' day period.

This addition was to provide a methodology in the ODCM to determine the projected doses as required by 23 80 79 Technical Specifications 6.8.4.a.5. T In Section 2.2.2 OTHER RELEASE POINTS, the following was added:

Symbos for this equation were defined in Section 2.2.1.

24 83 82 These words were added for clarity.

Added 2.6 The projected doses in a 31 day period are equal to the calculated doses from the current 31 day period.

This addition was to provide a methodology in the ODCM to determine the projected doses as required by 25 105 104 Technical Specifications 6.8.4.a.5 T Added Figure D-2-3 to show the gaseous release path for the AOG Ventilation System.

Along with D-2-1 and D-2-2 the three gaseous release paths (Stack, Turbine Building and AOG Building) are graphically 26 128 128 shown.

15

-301 Exek~

E eon,. CY-OC.170Revi:Sion Page 1 of 140 Nuclear OFFSITE DOSE CALCULATION MANUAL FOR OYSTER CREEK GENERATING STATION Revision of this document requires PORC approval and changes are controlled by CY-AA-170-3100 092909

CY-OC-170-301 Revision 4 Page 2 of 140 TABLE OF CONTENTS INTRODUCTION PART 1 - RADIOLOGICAL EFFLUENT CONTROLS 1.0 DEFINITIONS 3.0 APPLICABILIITY 3.3 INSTRUMENTATION 3.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3.3.3.11, RADIOACTIVE GASEOUS EFFLUENT MONTIORING INSTRUMENTATION 3.11.1 LIQUID EFFLUENTS 3.11.1.1 CONCENTRATION 3.11.1.2 DOSE 3.11.1.3 LIQUID WASTE TREATMENT SYSTEM 3.11.2 GASEOUS EFFLUENTS 3.11.2.1 DOSE RATE 3.11.2.2 DOSE NOBLE GASES 3.11.2.3 DOSE - IODINE -131, IODINE - 133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM 3.11.2.4 GASEOUS RADWASTE TREATMENT 3.11.3 MARK I CONTAINMENT 3.11.4 TOTAL DOSE 3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.12.1 MONITORING PROGRAM 3.12.2 LAND USE CENSUS 3.12.3 INTERLABORATORY COMPARISON PROGRAM 3.12.4 METEOROLOGICAL MONITORING PROGRAM 092909

CY-OC-1 70-301 Revision 4 Page 3 of 140 BASES FOR SECTIONS 3.0 AND 4.0 3.3 INSTRUMENTATION 3.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 3.11 RADIOACTIVE EFFLUENTS 3.11.1 LIQUID EFFLUENTS 3.11.1.1 CONCENTRATION 3.11.1.2 DOSE 3.11.1.3 LIQUID RADWASTE TREATMENT 3.11.2 GASEOUS EFFLUENTS 3.11.2.1 DOSE RATES 3.11.2.2 DOSE-NOBLE GAS 3.11.2.3 DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM 3.11.2.4 AUGMENTED OFFGAS TREATMENT SYSTEM 3.11.4 TOTAL DOSE 3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.12.1 MONITORING PROGRAM 3.12.2 LAND USE CENSUS 3.12.3 INTERLABORATORY COMPARISON PROGRAM 5.0 DESIGN FEATURES/SITE MAP 6.0 ADMINISTRATIVE CONTROLS 6.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (REOR) 6.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (RERR) 6.3 RESPONSIBILITES 092909

CY-OC-170-301 Revision 4 Page 4 of 140 PART II - CALCULATIONAL METHODOLOGIES 1.0 LIQUID EFFLEUNTS 1.1 RADIATION MONITORING INSTRUMENTATION AND CONTROLS 1.2 LIQUID EFFLUENT MONITOR SETPOINT DETERMINATION 1.2.1 LIQUID EFFLUENT MONITORS 1.2.2 SAMPLE RESULT SETPOINTS 1.2.3 ASSUMED DISTRIBUTION SETPOINTS 1.3 BATCH RELEASES 1.4 CONTINUOUS RELEASES 1.5 LIQUID EFFLUENT DOSE CALCULATION - 10CFR50 1.5.1 MEMBER OF THE PUBLIC DOSE - LIQUID EFFLUENTS 1.5.2 SHORELINE DEPOSIT DOSE 1.5.3 SHORELINE DOSE EXAMPLE 1.5.4 INGESTION DOSE - LIQUID 1.5.5 INGESTION DOSE CALCULATION EXAMPLE 1.5.6 PROJECTED DOSE - LIQUID 1.6 REPRESENTATIVE SAMPLES 2.0 GASEOUS EFFLUENTS 2.1 RADIATION MONITORING INSTRUMENTATION AND CONTROLS 2.2. GASEOUS EFFLUENT MONITOR SETPOINT DETERMINATION 2.2.1 PLANT VENT 2.2.2 OTHER RELEASE POINTS 2.2.3 RADIONUCLIDE MIX FOR SETPOINTS 2.3 GASEOUS EFFLUENT INSTANTANEOUS DOSE RATE CALCULATIONS 10CFR20 2.3.1 SITE BOUNDARY DOSE RATE - NOBLE GASES 2.3.1.1 TOTAL BODY DOSE RATE 2.3.1.2 EXAMPLE TOTAL BODY DOSERATE 092909

CY-OC-170-301 Revision 4 I Page 5 of 140 2.3.1.3 SKIN DOSE RATE 2.3.1.4 EXAMPLE SKIN DOSE RATE 2.3.2 SITE BOUNDARY DOSE RATE - RADIOIODINE AND PARTICULATES 2.3.2.1 METHOD - SITE BOUNDARY DOSE RATE -

RADIOIODINE AND PARTICULATES 2.3.2.2 EXAMPLE IODINE AND PARTICULATES DOSE RATE CALCULATION 2.4 NOBLE GAS EFFLUENT DOSE CALCULATION - 10CFR50 2.4.1 UNRESTRICTED AREA DOSE - NOBLE GASES 2.4.1.1 AIR DOSE METHOD 2.4.1.2 EXAMPLE NOBLE GAS AIR DOSE CALCULATION 2.4.1.3 INDIVIDUAL PLUME DOSE METHOD 2.5 RADIOIODINE PARTICULATE AND OTHER RADIONUCLIDES DOSE CALCULATIONS - 10CFR50 2.5.1 INHALATION OF RADIOIODINES, TRITIUM, PARTICULATES,AND OTHER RADIONUCLIDES 2.5.2 EXAMPLE CALCULATION - INHALATION OF RADIOIODINES, TRITIUM, PARTICULATES, AND OTHER RADIONUCLIDES 2.5.3 INGESTION OF RADIOIODINES, PARTICULATES AND OTHER RADIONUCLIDES 2.5.3.1 CONCENTRATION OF THE RADIONUCLIDE IN ANIMAL FORAGE AND VEGETATION -

OTHER THAN TRITIUM 2.5.3.2 EXAMPLE CALCULATION OF CONCENTRATION OF THE RADIONUCLIDE IN ANIMAL FORAGE AND VEGETATION -

OTHER THAN TRITIUM 2.5.3.3 CONCENTRATION OF TRITIUM IN ANIMAL FORAGE AND VEGETATION 2.5.3.4 EXAMPLE CALCULATION OF CONCENTRATION OF TRITIUM IN ANIMAL FORAGE AND VEGETATION 2.5.3.5 CONCENTRATION OF THE RADIONUCLIDE IN MILK AND MEAT 092909

CY-OC-1 70-301 Revision 41 Page 6 of140 2.5.3.6 EXAMPLE CALCULATION OF CONCENTRATION OF THE RADIONUCLIDE IN MILK AND MEAT 2.5.3.7 DOSE FROM CONSUMPTION OF MILK, MEAT, AND VEGETABLES 2.5.3.8 EXAMPLE CALCULATION - DOSE FROM CONSUMPTION OF MILK, MEAT, AND VEGETABLES 2.5.4 GROUND PLANE DEPOSITION IRRADIATION 2.5.4.1 GROUND PLANE CONCENTRATION 2.5.4.2 EXAMPLE GROUND PLANE CONCENTRATION CALCULATION 2.5.4.3 GROUND PLANE DOSE 2.5.4.4 EXAMPLE GROUND PLANE DOSE 2.6 PROJECTED DOSES - GASEOUS 3.0 TOTAL DOSE TO MEMBERS OF THE PUBLIC- 40CFR190 3.1 EFFLUENT DOSE CALCULATIONS 3.2 DIRECT EXPOSURE DOSE DETERMINATION 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM APPRENDIX A - DERIVED DOSE FACTORS AND RECEPTOR LOCATIONS Table A-i: Dose Conversion Factors for Deriving Radioactive Noble Gas Radionuclide-To-Dose Equivalent Rate Factors Table A-2: Noble Gas Radionuclide-To-Dose Equivalent Rate Factors Table A-3: Air Dose Conversion Factors for Effluent Noble Gas Table A-4: Locations Associated with Maximum Exposure of a Member of The Public Table A-5: Critical Receptor Noble Gas Dose Conversion Factors APPENDIX B - MODELING PARAMETERS Table B-1 :OCGS Usage Factors for Individual Dose Assessment Table B-2: Monthly Average Absolute Humidity APPENDIX C - REFERENCES TABLE C-1: REFERENCES 092909

CY-OC-1 70-301 Revision 4 I Page 8 of 140 OYSTER CREEK GENERATING STATION OFF SITE DOSE CALCULATION MANUAL INTRODUCTION The Oyster Creek Off Site Dose Calculation Manual (ODCM) is an implementing document to the Oyster Creek Technical Specifications. The previous Limiting Conditions for Operations that were contained in the Radiological Effluent Technical Specifications (RETS) are now included in the ODCM as Radiological Effluent Controls (REC). The ODCM contains two parts:

Part I - Radiological Effluent Controls, and Part II - Calculational Methodologies.

Part I includes the following:

" The Radiological Effluent Controls and the Radiological Environmental Monitoring Programs required by Technical Specifications 6.8.4

Descriptions of the information that should be included in the Annual Radioactive Effluent Release Report and the Annual Radiological Environmental Operating Report required by Technical Specifications 6.9.1.d and 6.9.1.e, respectively.

Part II describes methodologies and parameters used for:

The calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip set points; and

The calculation of radioactive liquid and gaseous concentrations, dose rates, cumulative yearly doses, and projected doses.

Part II also contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program (REMP), and the liquid and gaseous waste treatment systems.

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CY-OC-1 70-301 Revision 4 Page 9 of 140 PART - RADIOLOGICAL EFFLUENT CONTROLS 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these CONTROLS may be achieved. The defined terms appear in capitalized type and are applicable throughout these CONTROLS.

1.1 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s).

Implicit in the definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

A verification of OPERABILITY is an administrative check, by examination of appropriate plant records (logs, surveillance test records) to determine that a system, subsystem, train, component or device is not inoperable. Such verification does not preclude the demonstration (testing) of a given system, subsystem, train, component or device to determine OPERABILITY.

1.2 ACTION ACTION shall be that part of a CONTROL that prescribes remedial measures required under designated conditions.

1.4 CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds, with acceptable range and accuracy, to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including equipment actuation, alarm, or trip.

1.5 CHANNEL CHECK A CHANNEL CHECK shall be a qualitative determination of acceptable operability by observation of channel behavior during operation. This determination shall include, where possible, comparison of the channel with other independent channels measuring the same variable.

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CY-OC-170-301 Revision 4 Page 10 of 140 1.6 CHANNEL FUNCTIONAL TEST CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel to verify its proper response including, where applicable, alarm and/or trip initiating actions.

1.9 CONTROL The Limiting Conditions for Operation (LCOs) that were contained in the Radiological Effluent Technical Specifications were transferred to the OFF SITE DOSE CALCULATION MANUAL (ODCM) and were renamed CONTROLS. This is to distinguish between those LCOs that were retained in the Technical Specifications and those LCOs or CONTROLS that were transferred to the ODCM.

1.13 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

1.30 REPORTABLE EVENT A REPORABLE EVENT shall be any of those conditions specified Section 50.73 to 10CFR Part 50.

1.33 SOURCE CHECK SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

1.34 AUGMENTED OFF GAS SYSTEM (AOG)

The AUGMENTED OFF GAS SYSTEM is designed and installed to holdup and/or process radioactive gases from the main condenser off gas system for the purpose of reducing the radioactive material content of the gases before release to the environs.

1.35 MEMBER (S) OF THE PUBLIC MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with Exelon Generation and who do not normally frequent the Oyster Creek Generating Station site. This category does not include employees of the utility, its contractors, contractor employees, 092909

CY-OC-1 70-301 Revision 4 I Page 11 of 140 vendors, or persons who enter the site to make deliveries, to service equipment, work on site or for other purposes associated with plant functions. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant. An individual is not a member of the public during any period in which the individual receives an occupational dose.

1.36 OFF SITE DOSE CALCULATION MANUAL (ODCM)

The OFF SITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of Off Site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Set points, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radioactive Effluent Release Report AND Annual Radiological Environmental Operating Report required by Technical Specification Sections 6.9.1.d and 6.9.1.e, respectively.

1.37 PURGE - PURGING PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement and replacing it with air or gas.

1.38 SITE BOUNDARY The SITE BOUNDARY shall be the perimeter line around OCGS beyond which the land is neither owned, leased, nor otherwise subject to control by Exelon Generation. The area outside the SITE BOUNDARY is termed OFF SITE or UNRESTRICTED AREA.

1.39 OFF SITE The area that is beyond the site boundary where the land is neither owned, leased nor otherwise subject to control by Exelon Generation. Can be interchanged with UNRESTRICTED AREA.

1.40 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive 092909

CY-OC-170-301 Revision 4 I Page 12 of 140 materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. Can be interchanged with OFF SITE.

1.41 DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (micro curies per 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 conversion factors used for this calculation shall be those listed in Table E-7 of Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluences for the Purpose of Evaluating Compliance with 10CFR Part 40 Appendix I."

1.42 DEPOSITION (D/Q)

The direct removal of gaseous and particulate species on land or water surfaces. DEPOSITION is expressed as a quantity of material per unit area (e.g. m-2).

1.43 DOSE CONVERSION FACTOR (DCF)

A parameter calculated by the methods of internal dosimetry, which indicates the committed dose equivalent (to the whole body or organ) per unit activity inhaled or ingested. This parameter is specific to the isotope and the dose pathway. DOSE CONVERSION FACTORS are commonly tabulated in units of mrem/hr per picocurie/m 3 in air or water. They can be found in Reg Guide 1.109 appendices.

1.44 EFFLUENT CONCENTRATION (EC)

The liquid and air concentration levels which, if inhaled or ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem. LEC refers to liquid EFFLUENT CONCENTRATION.

1.45 ELEVATED (STACK) RELEASE An airborne effluent plume whose release point is higher than twice the height of the nearest adjacent solid structure and well above any building wake effects so as to be essentially unentrained. Regulatory Guide 1.111 is the basis of the definition of an ELEVATED RELEASE. Elevated releases generally will not produce any significant ground level concentrations within the first few hundred yards of the source.

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CY-OC-1 70-301 Revision 4 Page 13 of 140 ELEVATED RELEASES generally have less dose consequence to the public due to the greater downwind distance to the ground concentration maximum compared to ground releases. All main stack releases at the OCGS are ELEVATED RELEASES.

1.46 FINITE PLUME MODEL Atmospheric dispersion and dose assessment model which is based on the assumption that the horizontal and vertical dimensions of an effluent plume are not necessarily large compared to the distance that gamma rays can travel in air. It is more realistic than the semi-infinite plume model because it considers the finite dimensions of the plume, the radiation build-up factor, and the air attenuation of the gamma rays coming from the cloud. This model can estimate the dose to a receptor who is not submerged in the radioactive cloud. It is particularly useful in evaluating doses from an elevated plume or when the receptor is near the effluent source.

1.47 GROUND LEVEL (VENT) RELEASE An airborne effluent plume which contacts the ground essentially at the point of release either from a source actually located at ground elevation or from a source well above the ground elevation which has significant building wake effects to cause the plume to be entrained in the wake and driven to the ground elevation. GROUND LEVEL RELEASES are treated differently than ELEVATED RELEASES in that the X/Q calculation results in significantly higher concentrations at the ground elevation near the release point.

1.48 OCCUPATIONAL DOSE The dose received by an individual in a RESTRICTED AREA or in the course of employment in which the individual's assigned duties involve exposure to radiation and to radioactive material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other person. Occupational dose does not include dose received from background radiation, as a patient from medical practices, from voluntary participation in medical research programs, or as a member of the general public 1.49 "OPEN DOSE" A routine effluent dosimetry computer program that uses Reg.

Guides 1.109 and 1.111 methodologies.

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CY-OC-1170-301 Revision 4 Page 14 of 140 1.50 RAGEMS (RADIOACTIVE EMISSIONS MONITORING SYSTEM)

A plant system that monitors gaseous effluent releases from monitored release points. There is a RAGEMS system for the main stack (RAGEMS I) and one for the turbine building (RAGEMS II). They monitor particulates, iodine's, and noble gases.

1.51 SEMI-INFINITE PLUME MODEL Dose assessment model which is based on the assumption that the travel in air. The ground is considered to be an infinitely large flat plate and the receptor is located at the origin of a hemispherical cloud of infinite radius.

The radioactive cloud is limited to the space above the ground plane. The semi-infinite plume model is limited to immersion dose calculations.

1.52 SOURCE TERM The activity release rate, or concentration of an actual release or potential release. The common units for the source term are curies, curies per second, and curies per cubic centimeter, or multiples thereof (e.g., micro curies).

1.53 X/Q &- ("CHI over Q")

The dispersion factor of a gaseous release in the environment calculated by a point source Gaussian dispersion model. Normal units of X/Q are sec/m 3. The X/Q is used to determine environmental atmospheric concentrations by multiplying the source term, represented by Q (in units of pCi/sec or Ci/sec). Thus, the plume dispersion, X/Q (seconds/cubic meter) multiplied by the source term, Q (uCi/seconds) yields an!

environmental concentration, X (gCi/m 3). X/Q is a function of many parameters including wind speed, stability class, release point height, building size, and release velocity.

1.54 SEEDS (Simplified Effluent Environmental Dosimetry System)

A routine effluent dosimetry computer program that uses Reg.

Guides 1.109 and 1.111 methodologies.

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CY-OC-170-301 Revision 4 Page 15 of140 TABLE 1.1: SURVEILLANCE FREQUENCY NOTATION

NOTATION FREQUENCY 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.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

A At least once per 366 days.

R At least once per 18 months (550 days).

1/24 At least once per 24 months (refueling cycle)

StU Prior to each reactor startup.

P Prior to each radioactive release.

N.A. Not applicable.

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

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CY-OC-1 70-30'1 Revision 4 Page 16 of140 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY CONTROLS 3.0.1 Compliance with the CONTROLS contained in the succeeding CONTROLS is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the CONTROL, the associated ACTION requirements shall be met.

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

3.0.3 Except as provided in the associated ACTION requirements, when a CONTROL is not met or the associated ACTION requirements cannot be satisfied, action shall be initiated to place the unit into COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the CONTROL. Exceptions to these requirements are stated in the individual CONTROLS.

This CONTROL is not applicable in COLD SHUTDOWN or REFUELING.

3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made when the conditions of the CONTROLS are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual CONTROLS.

3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to CONTROL 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

092909

CY-OC-170-301 Revision 4 Page 17 of 140 3 /4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual CONTROLS unless otherwise stated in an individual Surveillance Requirement.

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

4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by CONTROL 4.0.2, shall constitute a failure to meet the OPERABILITY requirements for a CONTROL. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowed outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement(s) associated with the CONTROLS have been performed within the applicable surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

092909

CY-OC-170-301 Revision 4 I Page 18 of 140 3/4.3 INSTRUMENTATION 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CONTROLS 3.3.3.10 In accordance with Oyster Creek Technical Specifications 6.8.4.a.1, the radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.3.10-1 shall be OPERABLE with their Alarm/Trip set points set to ensure that the limits'of CONTROL 3.11.1.1 are not exceeded. The Alarm/Trip set points of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM Part II section

1.2.1. APPLICABILITY

During all liquid releases via these pathways.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip set point less conservative than required by the above CONTROL, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the set point so it is acceptably conservative, or provide for manual initiation of the Alarm/Trip function(s).

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.3.10-1. Make every reasonable effort to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report pursuant to Technical Specification 6.9.1 .d why the inoperability was not corrected in a timely manner.

c. The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable. Report all deviations in the Radioactive Effluent Release Report.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNL FUNCTIONAL TEST at the Frequencies shown in Table 4.3.3.10-1.

092909

CY=OCo170-301 Revision 4 Page 19 of 140 TABLE 3.3.3.10-1: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (DELETED) N/A 110

b. Turbine Building Sump No. 1-5 (DELETED) N/A 114

2. RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Reactor Building Service Water System Effluent Line 1 112

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line (DELETED) N/A 113 092909

CY-OC-1 70-301 Revision 4 Page 20 of 140 TABLE 3.3.3.10-1 (Continued)

TABLE NOTATIONS' ACTION 110 With no channels OPERABLE, effluent releases via this pathway may continue provided that:

a. At least two independent samples are taken, one prior to discharge and one near the completion of discharge and analyzed in accordance with SURVEILLANCE REQUIREMENT 4.11.1.1.1.

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

ACTION 112 With no channels OPERABLE, effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during the release, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 1E-6 jaCi/ml.

ACTION 113 With no channel OPERABLE, effluent releases via the affected pathway may continue provided the flow is estimated with the pump curve or change in tank level, at least once per batch during a release.

ACTION 114 With no channel OPERABLE effluent may be released provided that before initiating a release:

1. A sample is taken and analyzed in accordance with SURVEILLANCE REQUIREMENT 4.11.1.1.1.

2. Qualified personnel determine and independently verify the acceptable release rate.

092909

CY-OC170=301 Revision 4 Page 21 of 140 TABLE 4.3.3.10-1: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTSa CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (DELETED) N/A N/A N/A N/A

b. Turbine Building Sump No. 1-5 (DELETED) N/A N/A N/A N/A

2. RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Reactor Building Service Water System Effluent Line D M Re Qd

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line (DELETED) N/A N/A N/A N/A 0929 I

CY-OC-1 70-301 Revision 4 Page 22 of 140 TABLE 4.3.3.10-1 (Continued)

TABLE NOTATIONS

a. Instrumentation shall be OPERABLE and in service except that a channel may be taken out of service for the purpose of a check, calibration, test or maintenance without declaring it to be inoperable.

d. The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm set point.

2. Instrument indicates a downscale failure.

'3. Instrument controls not set in operate mode.

4. Instrument electrical power loss.

e. The CHANNEL CALIBRATION shall be performed according to established calibration procedures.

0929 1

CY-OC-170-301 Revision 4 Page 23 of 140 3/4.3 INSTRUMENTATION 3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS 3.3.3.11 In accordance with Oyster Creek Technical Specifications 6.8.4.a.1, the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.3.11-1 shall be OPERABLE with their alarm/trip set points set to ensure that the limits of CONTROL 3.11.2.1 are not exceeded. The alarm/trip set points of these channels meeting CONTROLS 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM Part II Section 2.2.

APPLICABILITY: As shown in Table 3.3.3.11-1 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip set point less conservative than required by the above CONTROL, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the set point so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.3.11-1. Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report pursuant to Technical Specification 6.9.1 .d why this inoperability was not corrected in a timely manner.

c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable. Report all deviations in the Radioactive Effluent Release Report.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.3.11-1.

0929 1

CY-OC-1 70o301 Revision 4 Page 24 of 140 TABLE 3.3.3.11-1: RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERAE ILEa APPLICABILITY ACTION

1. DELETED

2. STACK MONITORING SYSTEM

a. Radioactive Noble Gas Monitor (Low Range) 1 b,e 124

b. Iodine Sampler 1 b,e 127

c. Particulate Sampler 1 b,e 127

d. Effluent Flow Measuring Device 1 b 122

e. Sample Flow Measuring Device 1 b 128

3. TURBINE BUILDING VENTILATION MONITORING SYSTEM

a. Radioactive Noble Gas Monitor (Low Range) 1 b 123

b. Iodine Sampler 1 b 127

c. Particulate Sampler 1 b 127

d. Effluent Flow Measuring Device 1 b 122

e. Sample Flow Measuring Device 1 b 128 0929 I

CY-OC-1 70-301 Revision 4 Page 25 of 140 TABLE 3.3.3.11-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLEa APPLICABILITY ACTION

4. AUGMENTED OFF GAS BUILDING EXHAUST VENTILATION MONITORING SYSTEM

a. Radioactive Noble Gas Monitor 1 b 123

b. Iodine Sampler 1 b 127

c. Particulate Sampler 1 b 127

d. Sample Flow Measuring Device 1 -b 128 0929 1

CY-OC-170-301 Revision 4 Page 26 of 140 TABLE 3.3.3.11-1 (Continued)

TABLE NOTATIONS

a. Channels shall be OPERABLE and in service as indicated except that a channel may be taken out of service for the purpose of a check, calibration, test maintenance or sample media change without declaring the channel to be inoperable.

b. During releases via this pathway

e. Monitor / sampler or an alternate shall be OPERABLE to monitor / sample Stack effluent whenever the drywell is being purged.

ACTION 122 With no channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated whenever the exhaust fan combination in this system is changed.

ACTION 123 With no channel OPERABLE, effluent releases via this pathway may continue provided a grab sample is taken at least once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and is analyzed for gross radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter or provided an alternate monitoring system with local display is utilized.

ACTION 124 With no channel OPERABLE, effluent releases via this pathway may continue provided a grab sample is taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or provided an alternate monitoring system with local display is utilized. Drywell purge is permitted only when the radioactive noble gas' monitor is operating.

ACTION 127 With no channel OPERABLE, effluent releases via this pathway may continue provided the required sampling is initiated with auxiliary sampling equipment as soon as reasonable after discovery of inoperable primary sampler(s).

ACTION 128 With no channel OPERABLE, effluent releases via the sampled pathway may continue provided the sampler air flow is estimated and recorded at least once per day.

0929 1

CY-OC-170-301 Revision 4 Page 27 of 140 TABLE 4.3.3.11-1: RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVIELLANCE CHECK CHECK CALIBRATION TEST 'REQUIREDa INSTRUMENT

1. DELETED

2. MAIN STACK MONITORING SYSTEM

a. Radioactive Noble Gas Monitor (Low Range) D M 1 /2 4 t Qh b

b. Iodine Sampler W N.A. N.A. N.A. b

c. Particulate Sampler W N.A. N.A. N.A. b

d. Effluent Flow Measuring Device D N.A. 1/24 Q b

e. Sample Flow Measuring Device D N.A. R Q b

3. TURBINE BUILDING VENTILATION MONITORING SYSTEM

a. Radioactive Noble Gas Monitor (Low Range) D M 1/24' Qi b

b. Iodine Sampler W N.A. N.A. N.A. b

c. Particulate Sampler W N.A. N.A. N.A. b

d. Effluent Flow Measuring Device D N.A. 1/24 Q b

e. Sample Flow Measuring Device D N.A. R Q b 0929

CY-OC-170=301 Revision 4 Page 28 of 140 TABLE 4.3.3.11-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVIELLANCE IS INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED a

4. AUGMENTED OFF GAS BUILDING EXHAUST VENTILATION MONITORING SYSTEM

a. Radioactive Noble Gas Monitor D M Rf Qe b

b. Iodine Sampler W N.A. N.A. N.A. b

c. Particulate Sampler W N.A. N.A. N.A. b

d. Sample Flow Measuring Device D N.A. R N.A. b 0929 1

CY-OC-170-301 Revision 4 Page 29 of 140 TABLE 4.3.3.11-1 (Continued)

TABLE NOTATIONS

a. Instrumentation shall be OPERABLE and in service except that a channel may be taken out of service for the purpose of a check calibration, test or maintenance without declaring it to be inoperable.

b. During releases via this pathway.

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

1. Instrument indicates measured levels above the alarm set point.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mode.

4. Instrument electrical power loss.

f. The CHANNEL CALIBRATION shall be performed according to established calibration procedures.

h. The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm set point.

2. Instrument indicates a low count rate/monitor failure.

3. Switch cover alarm shall be verified to alarm when the cover is opened; and clear when the cover is closed after the faceplate switches are verified in their correct positions.

i. The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm set point.

2. Instrument indicates a low count rate/monitor failure.

0929 1

CY-OC-170-301 Revision 4 Page 30 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION CONTROLS 3.11.1.1 In accordance with the Oyster Creek Technical Specifications 6.8.4.a.2 and 3, the concentration of radioactive material, other than noble gases, in liquid effluent in the discharge canal at the Route 9 bridge (See Figure E-3) shall not exceed the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. The concentration of noble gases dissolved or entrained in liquid effluent in the discharge canal at the Route 9 bridge shall not exceed 2E-4 microcuries/milliliter.

APPLICABILITY: At all times.

ACTION:

a. In the event the concentration of radioactive material in liquid effluent released into the Off Site area beyond the Route 9 bridge exceeds either of the concentration limits above, reduce the release rate without delay to bring the concentration below the limit.

b. The provisions of CONTROLS 3.0.3, 3.0.4 and Technical Specification 6.9.2 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program in Table 4.11.1.1.1-1.

Alternately, pre-release analysis of batches(es) of radioactive liquid waste may be by gross beta or gamma counting provided a maximum concentration limit of 1E-8 pCi/mI in the discharge canal at the Route 9 bridge is applied.

4.11.1.1.2 The results of the radioactivity analyses shall be used' in accordance with the methodology and parameters in the ODCM Part II Section 1.2 to assure that the concentrations at the point of release are maintained within the limits of CONTROL 3.11.1.1 and 3.11.1.2.

4.11.1.1.3 The alarm or trip set point of each radioactivity monitoring channel in Table 3.3.3.10-1 shall be determined on the basis of sampling and analyses results obtained according to Table 4.11.1.1.1.-1 and the set point method in ODCM Part 111.2.1 and set to alarm or trip before exceeding the limits of CONTROL 3.11.1.1.

0929 1

CY-OC-170-301 Revision 4 Page 31 of 140 TABLE 4.11.1.1.1-1: RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Liquid Release Type Sampling Minimum Type of Activity Analysis Lower Limit Frequency Analysis Detectiona (LLD)

Frequency (lCi/ml)

A. Batch Waste P PC Principal Gamma Emitters 5E-07 Release Tanks Each Each Batch 1-131 5E-07 Batchb p M Dissolved and Entrained 1E-05 One Gases (Gamma Emitters)

Batch/Mb P M H-3 1E-05 Each Composited Gross Alpha 1 E-07 Batchb P Q Sr-89, Sr-90 5E-08 Each Composited Fe-55 1E-06 Batchb B. Reactor Building, W W Principal Gamma Emitters 5E-07 New Radwaste, and AOG Service Grab 1-131 5E-07 Water Effluent Sample6 (note f) M H-3 1E-05 Compositeg Gross Alpha 1E-07 (note f) Q Sr-89, Sr-90 5E-08 Composites Fe-55 1E-06 0929 1

CY-OC-1 70-301 Revision 4 Page 32 of 140 TABLE 4.11.1.1-1 (CONTINUED)

TABLE NOTATIONS

a. The Lower Limit of Detection (LLD) is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

The LLD is applicable to the capability of a measurement system under typical conditions and not as a limit for the measurement of a particular sample in the radioactive liquid waste sampling and analyses program.

For a particular measurement system, which may include radiochemical separation:

LLD = 4.66

Sb E

V

2.22E6

Y

exp(-IAt)

Where:

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

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

E is the counting efficiency, V is the sample size (units of mass or volume),

2.22E+6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between the end of the sample collection and the time of counting.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions with typical values of E, V, Y, and t for the radionuclides Mn-54, Fe-59, Co-58, Co-60, Zn-65, Ce-141, Cs-I 34, Cs-1 37; and an LLD of 1E-5 iCi/ml should typically be achieved for Mo-99 and Ce-144.

0929 1

CY-OC-170-301 Revision 4 Page 33 of 140 TABLE 4.11.1.1.1-1 (CONTINUED)

TABLE NOTATIONS Occasionally, background fluctuations, interfering radionuclides, or other uncontrollable circumstances may render these LLD's unachievable.

When calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background may include the typical contributions of other radionuclides normally present in the sample. The background count rate of a semiconductor detector (e.g. HPGe) is determined from background counts that are determined to be within the full width of the specific energy band used for the quantitative analysis for the radionuclide.

The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.

This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall be identified and reported.

The LLD for Mo-99 and Ce-144 is 1 E-5 giCi/mL Whereas the LLD for the other gamma emitters is 5E-7 gICi/mL. Nuclides that are below the LLD for the analysis should not be reported.

b. A batch release is the discharge of liquid wastes of a discrete volume. Before sampling for analysis, each batch should be thoroughly mixed.

c. In the event a gross radioactivity analysis is performed in lieu of an isotopic analysis before a batch is discharged, a sample will be analyzed for principal gamma emitters afterwards.

d. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

e. Analysis may be performed after release.

f. In the event a grab sample contains more than 5E-7 igCi/mL of 1-131 and principal gamma emitters or in the event the Reactor Building Service Water radioactivity monitor indicates more than 5E-7 liCi/mL radioactivity in the effluent, as applicable, sample the elevated activity effluent daily until analysis confirms the activity concentration in the effluent does not exceed 5E-7 gCi/mL. In addition a composite sample must be made up for further analysis for all samples taken when the activity was > 5E-7 t.Ci/mL.

g. A composite sample is produced combining grab samples, each having a defined volume, collected routinely from the sump or stream being sampled 0929 I

CY-OC-1 70-301 Revision 4 Page 34 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1.2 DOSE CONTROLS 3.11.1.2 In accordance with Oyster Creek Technical Specifications 6.8.4.a.4 and 5, the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see Figure E-4) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the Total Body and to less than or equal to 5 mrem to any body organ, and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days from the end of the quarter, 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 that have been taken and/or will be 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.

b. The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM Part IISection 1.5 at least once per 31 days in accordance with Technical Specification 6.8.4.a.5.

0929 I

CY-OC-170-301 Revision4 I Page 35 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1.3 LIQUID WASTE TREATMENT SYSTEM CONTROLS 3.11.1.3 In accordance with the Oyster Creek Technical Specifications 6.8.4.a.6, the liquid radwaste treatment system shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when projected doses due to the liquid effluent to UNRESTRICTED AREAS (see Figure E-4) would exceed 0.06 mrem to the Total Body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of the above, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:

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

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

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

b.The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases to-UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM Part II Section 1.5 in accordance with Technical Specifications 6.8.4.a.5.

4.11.1.3.2 The installed liquid radwaste treatment system shall be demonstrated OPERABLE by meeting CONTROLS 3.11.1.1, 3.11.1.2, and 3.11.1.3.

0929 I

CY-OC-170-301 Revision 4 Page 36 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE CONTROLS 3.11.2.1 In accordance with the Oyster Creek Technical Specifications 6.8.4.a.5 and 7, the dose rate due to radioactive materials released in gaseous effluents in the UNRESTRICTED AREA (see Figure E-3) shall be limited to the following:

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

b. For iodine-1 31, iodine-1 33, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any body organ.

APPLICABILITY: At all times.

ACTION:

a. With the dose rate(s) exceeding the above limits, immediately restore the release rate to within the above limit(s).

b. Ifthe gaseous effluent release rate cannot be reduced to meet the above limits, the reactor shall be in at least SHUTDOWN CONDITION within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless corrective actions have been completed and the release rate restored to below the above limit.

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM Part II Section 2.3.1.

4.11.2.1.2 The dose rate due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM Part II Section 2.3.2 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2.1.2-1.

4.11.2.1.3 Dose rates due to tritium, Sr-89, Sr-90, and alpha-emitting radionuclides are averaged over no more than 3 months and the dose rate due to other radionuclides is averaged no more than 31 days.

0929 I

CY-OC-1 70-301 Revision 4 Page 37 of 140 TABLE 4.11.2.1.2-1: RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Sampling Minimum Type of Activity Lower Limit Type Frequency Analysis Analysis Detectiona (LLD)

Frequency (4Ci/ml)

Stack; Turbine Q Q H-3 1 E-06 Building Exhaust Vents; Augmented Grab Sample' Off gas Building Vent M M Principal 1E-04 Grab Sample Gamma b Grab SEmittersb (Noble c~d,f Gases)

Continuousf W 1-131 1E-12 Charcoal 1-133 1E-10 Sample _

Continuousf W Principal 1E-11 Particulate Gamma Sample Emittersb (particulates)

Continuousf Me Gross Alpha 1E-11 Composite Particulate Sample Continuous Qe Sr-89, Sr-90 1E-11 Composite Particulate Sample Continuous Noble Gas Noble Gases 1E-06 Monitor Gamma Radioactivity 0929 I

CY-OC-1 70-301 Revision 4 Page 38 of 140 TABLE 4.11.2.1.2-1 (Continued)

TABLE NOTATIONS

a. The Lower Limit of Detection (LLD) is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

The LLD is applicable to the capability of a measurement system under typical conditions and not as a limit for the measurement of a particular sample in the radioactive liquid waste sampling and analyses program.

For a particular measurement system, which may include radiochemical separation:

LLD= 4.66 Sb E

V

2.22E6 *Y

exp(-AAt)

Where:

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

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

E is the counting efficiency, V is the sample size (units-of mass or volume),

2.22E+6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, X is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between the end of the sample collection and the time of counting.

Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions with typical values of E, V, Y, and t for the radionuclides Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137, and Ce-141. Occasionally background fluctuations, or other uncontrollable circumstances may render these LLD's unachievable.

When calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background may include the typical contributions of other radionuclides normally present in the 0929 1

CY-OC-1 70-301 Revision 4 Page 39 of 140 TABLE 4.11.2.1.2-1 (Continued)

TABLE NOTATIONS samples. The background count rate of a HpGe detector is determined from background counts that are determined to be within the full width of the specific energy band used for the quantitative analysis for that radionuclide

b. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-1 33m, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report consistent with CONTROL 3.11.2.1. The LLD for Mo-99 and Ce-144 is 1E-10 gCi/ml whereas the LLD for other principal gamma emitting particulates is 1E-1 1 gCi/ml.

Radionuclides which are below the LLD for the analysis should not be reported.

c. The noble gas radionuclides in gaseous effluent may be identified by taking a grab sample of effluent and analyzing it.

d. In the event the reactor power level increases more than 15 percent in one hour and the Stack noble gas radioactivity monitor shows an activity increase of more than a factor of three after factoring out the effect due to the change in reactor power, a grab sample of Stack effluent shall be collected and analyzed.

e. A composite particulate sample shall include an equal fraction of at least one particulate sample collected during each week of the compositing period.

f. In the event a sample is collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less, the LLD may be increased by a factor of 10.

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CY-OC-170-301 Revision 4 I Page 40 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.2 DOSE - NOBLE GASES CONTROLS 3.11.2.2 In accordance with the Oyster Creek Technical Specification 6.8.4.a.5 and 8, the air dose due to noble gases released in gaseous effluents in the UNRESTRICTED AREA (see Figure E-3) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days from the end of the quarter during which the release occurred, 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 that have been taken to reduce the release and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM Part IISection 2.4.1 at least once per 31 days in accordance with Technical Specification 6.8.4.a.5.

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CY-OC-1 70-301 Revision 4 I Page 41 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.3 DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM CONTROLS 3.11.2.3 In accordance with Oyster Creek Technical Specification 6.8.4.a.5 and 9, the dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-1 33, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released in the UNRESTRICTED AREA (see Figure E-3) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, iodine-133 and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, 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 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.

b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for odine-1 31, iodine-1 33, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM Part II Section 2.5 at least once per 31 days in accordance with Technical Specification 6.8.4.a.5.

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CY-OC-170-301 Revision 4 Page 42 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.4 GASEOUS RADWASTE TREATMENT CONTROLS 3.11.2.4 The AUGMENTED OFF GAS SYSTEM shall be in operation to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY (see Figure E-4) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any body organ APPLICABILITY: Whenever the main condenser steam jet air ejector is in operation except during startup or shutdown with reactor power less than 40 percent of rated. In addition, the AUGMENTED OFF GAS SYSTEM need not be in operation during end of cycle coast-down periods when the system can no longer function due to low off gas flow.

ACTION:

a. Every reasonable effort shall be made to maintain and operate charcoal absorbers in the AUGMENTED OFF GAS SYSTEM to treat radioactive gas from the main condenser air ejectors.

b. With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 30 consecutive days and either CONTROL 3.11.2.1 or 3.11.2.4 exceeded, prepare and submit to the Commission within 30 days from the end of the quarter during which release occurred, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:

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

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

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

c. The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Operation of the Augmented Off gas System charcoal absorbers shall be verified by verifying the AOG System bypass valve, V-7-31, alignment or alignment indication closed at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the main condenser air ejector is operating.

4.11.2.4.2 Doses due to gaseous releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM Part II Section 1.5 in accordance with Technical Specifications 6.8.4.a.5.

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CY-OC-170-301 Revision 4 Page 43 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.3. MARK I CONTAINMENT CONTROLS 3.11.3.1 Venting or purging of the containment Drywell may be through normal Reactor Building.

Ventilation if the following requirements are met:

APPLICABILITY:

If the Station year-to-date radiological effluent releases (either iodine or noble gas) are less than 10% of the ODCM limit, then Standby Gas Treatment is NOT required for purging the contents of the Drywell.

ACTION:

If the Station year-to-date radiological effluent releases (either iodine or noble gas) are greater than '10% of the ODCM limit, then the Standby Gas Treatment System must be used for purging the contents of the Drywell.

SURVEILLANCE REQUIREMENTS 4.11.3.1 The Standby Gas Treatment System is OPERABLE and available whenever the purge system is in use.

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CY-OC-1 70-301 Revision 4 Page 44 of 140 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3.11.4.1 In accordance with Oyster Creek Technical Specifications 6.8.4.a.10, the annual (calendar year) dose commitment to any MEMBER OFTHE PUBLIC due to radioactive material in the effluent and direct radiation from the OCGS in the UNRESTRICTED AREA shall be limited to less than or equal to 75 mrem to the thyroid or less than or equal to 25 mrem to the total body or any other organ.

APPLICABILITY: At all times ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of CONTROLS 3.11.1.2a, 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a, or 3.11.2.3b, perform an assessment to determine whether the limits of CONTROL 3.11.4.1 have been exceeded. Ifsuch is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report shall include information specified in 10CFR20.2203. Ifthe estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SURVEILLANCE REQUIREMENT 4.11.1.2, 4.11.2.2, 4.11.2.3, and in accordance with the methodology and parameters in the ODCM Part II Section 3.0 at least once per year.

4.11.4.2 Cumulative dose contributions from direct radiation from the facility shall be determined in accordance with the methodology and parameters in the ODCM Part II Section 3.2. This requirement is applicable only under conditions set forth in CONTROL 3.11.4, ACTION a.

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CY-OC-170-301 Revision 4 Page 45 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM CONTROLS 3.12.1. In accordance with Oyster Creek Technical Specifications 6.8.4.b, the radiological environmental monitoring program shall be conducted as specified in Table 3.12.1-1. For specific sample locations see Table E-1. Revisions to the non-ODCM required portions of the program may be implemented at any time. Non-ODCM samples are those taken in addition to the minimum required samples listed in Table 3.12.1-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification 6.9.1.e, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the Commission within 60 days of the end of the quarter, 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 CONTROLS 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) 1.0 reporting level (I) reporting level (2)

When radionuclides other than those in Table 3.12.1-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose* to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of CONTROLS 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report pursuant to Section 6.1.2.1.

The methodology used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

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CY-OC-170-301 Revision 4 Page 46 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM CONTROLS (Continued)

ACTION: (Continued)

c. Ifgarden vegetation samples are unobtainable due to any legitimate reason, it is NOT ACCEPTABLE to substitute vegetation from other sources. The missed sample will be documented in the annual report, with no further actions necessary. Ifa permanent sampling location becomes unavailable, follow Table 3.12.1-1 Table Notation (1) to replace the location.

d. The provisions of CONTROLS 3.0.3, 3.0.4 and Technical Specification 6.9.2 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in Table E-1, and shall be analyzed pursuant to the requirements of Table 3.12.1-1, and the detection capabilities required by Table 4.12.1-1.

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CY-OC-170-301 Revision 4 Page 47 of 140 TABLE 3.12.1-1: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTAIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(1 ) COLLECTION FREQUENCY OF ANALYSIS

1. DIRECT RADIATION(2) Routine monitoring Quarterly Gamma dose quarterly stations with two or more dosimeters placed as follows:

An inner ring of stations one in each meteorological sector in the general area of the SITE BOUNDARY (At least 16 locations);

An outer ring of stations, one in each land-based meteorological sector in the approximately 6- to 8-km range from the site (At least 14 locations); and At least 8 stations to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

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CY-OC-1 70-301 Revision 4 Page 48 of 140 TABLE 3.12.1-1(Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(1 ) COLLECTION FREQUENCY OF ANALYSIS

2. AIRBORNE Radioiodine and Samples from 5 locations: Continuous sampler Radioiodine Canister:

Particulates operation with sample 1-131 analysis weekly.

Three samples from close to collection weekly or the SITE BOUNDARY in different more frequently if sectors of the highest calculated required by dust Particulate Sampler annual average ground- loading. Gross beta radioacti-level D/Q. vity analysis follow-ing filter change(3);

One sample from the vicinity Gamma isotopic of a community having the highest analysis(4) calculated annual average ground- of composites (by level D/Q; and location) quarterly.

One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction (6).

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CY-OC-1 70-301 Revision 4 Page 49 of 140 TABLE 3.12.1-1(Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(1 ) COLLECTION FREQUENCY OF ANALYSIS

3. WATERBORNE

a. Surface One sample upstream Grab sample weekly, Gamma isotopic and One sample downstream Combine into monthly tritium analysis(4 ).

composite.

b. Ground(5) Samples from one or two sources Grab sample quarterly. Gamma isotopic and tritium if likely to be affected. analysis(4 ).

c. Drinking 1 sample of each of 1 to 3 of the Grab sample weekly, Gross beta, gamma nearest water supplies that could combine into monthly isotopic and tritium analysis be affected by its discharge. composite. monthly (4)(7)

One sample from a background location.

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CY-OC-170-301 Revision 4 Page 50 of 140 TABLE 3.12.1 -1 (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(') COLLECTION FREQUENCY OF ANALYSIS

d. Sediment One sample from downstream area Semiannually Gamma isotopic with existing or potential recreational analysis(4) semiannually.

value.

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CY=OC-1 70-301 Revision 4 Page 51 of 140 TABLE 3.12.1-1(Cont'dI RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(1 ) COLLECTION FREQUENCY OF ANALYSIS

4. INGESTION

a. Milk (6)

No milking animals Semimonthly when on pasture; Gamma isotopic (4) and Iodine -131 If milk animals are identified: monthly at other times semimonthly when Samples from milking animals in animals are on pasture; three locations within 5km having monthly at other times the highest dose potential. If there are none, then one sample from milking animals in each of three areas between 5 an 8 km distant where doses are calculated to be greater than 1 mrem per year. One sample from milking animal at a control location 15 to 30 km distant and in the least prevalent wind direction

b. Fish One sample of available species Semiannually, Gamma isotopic (4) consumed by man in plant when available analysis discharge canal. on edible portions.

One sample of available species consumed by man not influenced by plant discharge.

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CY-OC-1 70-301 Revision 4 Page 52 of 140 TABLE 3.12.1-1 (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS(1 ) COLLECTION FREQUENCY OF ANALYSIS

c. Clams One sample of available species Semiannually, Gamma isotopic consumed by man within the (4) when available analysis influence of the facility discharge. on edible portions.

One sample of available species consumed by man not influenced by plant discharge.

d. Vegetation (8) 3 samples of broad leaf Monthly during Gamma isotopic vegetation grown nearest each growing season analysis (4) and 1-131 on of two different Off Site locations of edible portion.

highest predicted annual average combined elevated and ground level release D/Q One sample of each of the similar broad leaf vegetation grown at least 15 to 30 km (9.3-18.6 miles) distant in the least prevalent wind direction.

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CY-OC-170-301 Revision 4 Page 53 of 140 TABLE 3.12.1-1 (Continued)

TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of the reactor, and additional description where pertinent, are provided for each and every sample location in Table 3.12.1-1 and Table E-1. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.1.2.4. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM.

Pursuant to Technical specification 6.19, submit in the next Radioactive Effluent Release Report documentation for a change in the ODCM including revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for the pathway and justifying the selection of the new location(s) for obtaining samples. This applies to changes/deletions/additions of permanent sampling locations. This does not apply to one-time deviations from the sampling schedule. In those cases, it is NOT ACCEPTABLE to substitute sample media from other sources. The missed sample will be documented in the annual report, with no further actions necessary.

(2) One or more instruments, such as pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in'a packet are considered as two or more dosimeters. The number of direct radiation monitoring stations has been reduced from the NUREG 1302 recommendation due to geographical limitations; e.g.,

some sectors are over water and some sectors cannot be reached due to lack of highway access, therefore the number of dosimeters has been reduced accordingly.

(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 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 particulate is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

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CY-OC-170-301 Revision 4 Page 54 of 140 (5) Groundwater samples shall 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. Extensive studies of geology and groundwater in the vicinity of the OCGS (Reference 21 and 31) have demonstrated that there is no plausible pathway for effluents from the facility to contaminate offsite groundwater, including the local drinking water supplies. Samples of groundwater, including local drinking water wells, are collected in order to provide assurance to the public that these water resources are not impacted.

(6) The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites which provide valid background data may be substituted.

(7) 1-131 analysis on each composite when the dose calculated for the consumption of the water is greater than 1 mrem per year.

(8) Ifgarden vegetation samples are unobtainable due to any legitimate reason (see (1) above), it is NOT ACCEPTABLE to substitute vegetation from other sources. The missed sample will be documented in the annual report, with no further actions necessary.

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CY-OC-170-301 Revision 4 Page 55 of 140 TABLE 3.12.1-2: REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES - REPORTING LEVELS Surface and Airborne Fish Milk Vegetation Analysis Ground Particulate (pCi/Kg, (pCi/I) (pCi/Kg, Water(pCi/I) and Iodine wet) wet)

(pCi/m3)

H-3 20000*

Mn-54 1000 30000 Fe-59 400 10000 Co-58 1000 30000 Co-60 300 10000 Zn-65 300 20000 Zr-Nb-95 400 1-131 2** 0.9 3 100 Cs-134 30 10 1000 60 1000 Cs-1 37 50 20 2000 70 2000 Ba-La-140 200 300

For drinking water samples (this is the 40 CFR Part 141 value).

If no drinking water pathway exists, a value of 30,000 pCi/L may be used.

- If no drinking water pathway exists, a value of 20 pCi/L may be used.

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CY-OC-170-301 Revision 4 Page 56 of 140 TABLE 4.12.1-1: DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS('1 )(2) LOWER LIMITS OF DETECTION (LLD)(3)

Surface and Air Vegetation Sediment Fish and Analysis Ground Particulate (pCi/Kg, (pCi/Kg, dry) Clams Water and Air wet) (pCi/Kg, wet)

(pCi/I) Iodine (pCi/m 3)

Gross Beta 4 0.01 H-3 2000(')

Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 1(4) .07(5) 60 Cs-134 15 .05(6) 60 150 130 Cs-137 18 .06(6) 80 180 150 La-140 15 Ba-140 60 0929 1

CY-OC-170-301 Revision 4 Page 57 of 140 TABLE 4.12.1-1 (Continued)

TABLE NOTATIONS (1) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.1.2.3.

(2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

(3) The LLD is-defined, for purposes of these CONTROLS as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD 4.66

Sb E

V

2.22

Y

exp(-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per Pico curie, Y is the fractional radiochemical yield, when applicable, X is the radioactive decay constant for the particular radionuclide (secl), and At for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting (sec).

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

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CY-OC-170-301 Revision 4 Page 58 of 140 TABLE 4.12.1-1 (Continued)

TABLE NOTATIONS 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 Radiological Environmental Operating Report pursuant Technical Specification 6.9.1.e and Control 6.1.2.6.4.

(4) If no drinking water pathway exists, a value of 3000 pCi/L for tritium and 15 pCi/L for iodine-131 may be used.

(5) For the air iodine sample (6) For the air particulate sample 0929 I

CY-OC-170-301 Revision 4 Page 59 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS CONTROLS 3.12.2 In accordance with the Oyster Creek Technical Specifications 6.8.4.b, a land use census shall be conducted and shall identify within a distance of 5 miles the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden* of greater than 500 ft2 producing broad leaf vegetation. The census shall also identify within a distance of 3 miles the location in each of the 16 meteorological sectors all milk animal and all gardens greater than 500 square feet producing broadleaf vegetation.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in SURVEILLANCE REQUIREMENT 4.11.2.3, identify the new location(s) in the next Radioactive Effluent Release Report, pursuant to Control 6.2.2.4.

b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with CONTROL 3.12.1, add the new location(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may then be deleted from this monitoring. Pursuant to CONTROL 6.2.2.4, identify the new location(s) in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by door-to-door survey, visual survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.1.2.2.

Broad leaf 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 elevated release D/Q's in lieu of the garden census. Controls for broadleaf vegetation sampling in Table 3.12.1-1, Part 4.c shall be followed, including analysis of control samples.

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CY-OC-170-301 Revision 4 Page 60 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM CONTROLS 3.12.3 In accordance with Oyster Creek Technical Specifications 6.8.b.3, analyses shall be performed on radioactive materials supplied as part of an Interlaboratory comparison program which has been approved by the Commission.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the reason and corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.1.2.6.3.

b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above-required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.1.2.6.3.

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CY-OC-170-301 Revision 4 Page 61 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.4 METEOROLOGICAL MONITORING PROGRAM CONTROLS 3.12.4 The meteorological monitoring instrumentation channels shown in Table 3.12.4.-i shall be operable.

APPLICABILITY: At all times.

ACTION:

a. With less than the minimum required instrumentation channels OPERABLE for more than 7 days, initiate an Issue Report outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status.

c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.

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CY-OC-170-301 Revision 4 Page 62 of 140 TABLE 3.12.4-1 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT ELEVATION MINIMUM INSTRUMENT OPERABLE

1. Wind Speed a 380 feet 1

b. 150 feet 1

c. 33 feet 1

2. Wind Direction

a. 380 feet 1

b. 150 feet 1

c. 33 feet 1

3. AT

a. 380-33 1

b. 150-33 1 0929 I

CY-OC-170-301 Revision 4 Page 63 of 140 BASES FOR SECTIONS 3.0 AND 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS NOTE: The BASES contained in the succeeding pages summarize the reasons for the CONTROLS of Sections 3.0 and 4.0, but are not considered a part of these CONTROLS.

3/4.3 INSTRUMENTATION BASES 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The reactor service water system discharge line radioactivity monitor initiates an alarm in the Control Room when the alarm set point is exceeded. The alarm/trip set points 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 consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip set points for each of the noble gas monitors shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM. This will ensure the alarm/trip will occur prior to exceeding the limits of 10 CFR Part

20. The radioactive gas monitors for the stack effluent and the Augmented Off gas Building exhaust ventilation have alarms which report in the Reactor Control Room. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

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CY-OC-170-301 Revision 4 Page 64 of 140 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This CONTROL is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix 1,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(a) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-1 35 is the controlling radioisotope and its concentration limit in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP)

Publication 2.

The value 1E-8 is the limit for unidentified gross gamma or beta releases as per 10 CFR 20 Appendix B, Table 2, Column 2 "any single radionuclide.. .other than alpha or spontaneous fission ... half life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ". This provides operational flexibility while providing reasonable assurance that dose will remain less than 0.1 rem/yr.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in references 25, 26, and 27.

Weekly grab samples for Service Water Effluents are composited for monthly tritium and gross alpha analysis and quarterly Sr-89,90 and Fe-55 analysis if activity is detected. New Radwaste and AOG Service Water Effluents have no continuous radiation monitors. Weekly grab samples are performed to detect leaks. Plant design does not include composite samplers, so the weekly grab sample frequency is considered adequate.

Circulating Water Effluent is not included in Table 4.11.1.1.1-1, Radioactive Liquid Waste Sampling and Analysis Program since the Circulating Water is sampled as part of the Radiological Environmental Monitoring Program, Table 3.12.1-1, 3a, Waterborne, Surface downstream sample.

3/4.11.1.2 DOSE This CONTROL is provided to implement the requirements of Sections II.A, Ill.A, and IV.A of Appendix 1,10 CFR Part 50. The CONTROL implements the guides set forth in Section I.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is 0929 I

CY-OC-170-301 Revision 4 Page 65 of 140 reasonably achievable." The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109 and Regulatory Guide 1.113.

3/4.11.1.3 LIQUID RADWASTE TREATMENT The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to their release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This CONTROL implements the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section I.A of Appendix 1,10 CFR Part 50, for liquid effluents. Figure D-1-la, Liquid Radwaste Treatment Chem Waste and Floor Drain System and Figure D-1-lb, Liquid Radwaste Treatment - High Purity and Equipment Drain System provides details of the Liquid Radwaste Treatment system.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This CONTROL is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits 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 2, 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 specified in Appendix B, Table 2 of 10 CFR Part 20. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the individual will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor 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 corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/yr to the skin. These 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 mrem/year.

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CY-OC-170-301 Revision 4 Page 66 of 140 The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in references 25, 26 and 27.

Tritium is sampled quarterly for gaseous effluents. Based on the consistency of the data from the quarterly sampling, the sampling frequency is adequate.

3/4.11.2.2 DOSE - NOBLE GASES This CONTROL is provided to implement the requirements of Section 11.3, III.A and IV.A of Appendix 1,10 CFR Part 50. The CONTROL implements the guides set forth in Section 11.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The SURVEILLANCE REQUIREMENTS implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109 and Regulatory Guide 1.111. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

3/4.11.2.3 DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This CONTROL is provided to implement the requirements of Section II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The CONTROLS are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRERSTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in SURVEILLANCE REQUIREMENTS implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM 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, and Regulatory Guide 1.111. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for iodine-1 31, iodine-1 33, tritium, and radionuclides in particulate form with half-life greater than 8 days are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

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CY-OC-170-301 Revision 4 Page 67 of 140 3/4.11.2.4 AUGMENTED OFF GAS TREATMENT SYSTEM The OPERABILITY of the AUGMENTED OFF GAS TREATMENT SYSTEM (AOG) ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This CONTROL implements the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents. Figure D-2-1, Gaseous Radwaste Treatment - Augmented Off gas System, Figure D-2-2, Ventilation System provide details of the Augmented Off Gas Treatment System and Figure D-2-3, AOG Ventilation System.

3/4.11.4 TOTAL DOSE This CONTROL is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The CONTROL requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the doses remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the unit, including outside storage tanks, etc. 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 purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190 and 10 CFR Part 20, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in CONTROLS 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

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CY-OC-1 70-301 Revision 4 Page 68 of 140 314.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this CONTROL provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the' highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Position on Environmental Monitoring, Revision 1, November 1979.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a Driori (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 references 25, 26, and 27.

Site-specific research, which included the installation of a groundwater monitoring well network, has demonstrated that the groundwater pathway is not a potential pathway to man from the OCGS. The surface water into which the OCGS discharges is a marine estuary containing saline water that is not used as drinking water or irrigation water by man.

3/4.12.2 LAND USE CENSUS This CONTROL 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 environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey, from visual survey or consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix Ito 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 (500 ft2) provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) 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 growin 9 broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/mr.

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CY-OC-1 70-301 Revision 4 Page 69 of 140 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.5.0 5.0 DESIGN FEATURES I SITE MAP (Provided FOR INFORMATION ONLY. Technical Specifications are controlling.)

5.1 Site map which will allow identification of structures and release points shall be as shown in Figure E-4.

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CY-OC-170-301 Revision 4 Page 70 of 140 6.0 ADMINISTRATIVE CONTROLS 6.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (REOR) 6.1.1 In accordance with Oyster Creek Technical Specifications 6.9.1 .e, a routine radiological environmental operating report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of the following year.

6.1.2 The Annual Radiological Environmental Operating Reports shall include:

6.1.2.1 Summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities (Radiological Environmental Monitoring Program -REMP) for the report period. This will include a comparison with preoperational studies, with operational controls (as appropriate), and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

6.1.2.2 The reports shall also include the results of land use censuses required by CONTROL 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

6.1.2.3 The Annual Radiological Environmental Operating Reports shall include summarized and tabulated results similar in format to that in Regulatory Guide 4.8, December 1975 of all the radiological environmental samples taken during the report period.

6.1.2.4 Deviations from the sampling program identified in CONTROL 3.12.1 shall be reported.

6.1.2.5 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. The missing data shall be submitted as soon as possible in a supplementary report.

6.1.2.6 The reports shall also include the following:

6.1.2.6.1 A summary description of the radiological environmental monitoring Program; 6.1.2.6.2 Map(s), covering sampling locations, keyed to a table giving distances and directions from the reactor; 6.1.2.6.3 The results of licensee participation in the Inter-laboratory Comparison Program, as required by CONTROL 3.12.3; 6.1.2.6.4 Identification of environmental samples analyzed when the analysis instrumentation was not capable of meeting the detection capabilities in Table 4.12.1-1.

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CY-OC-1 70-301 Revision 4 Page 71 of 140 6.2 ANNUAL ROUTINE RADIOACTIVE EFFLUENT RELEASE REPORT (RERR) 6.2.1 Routine radioactive effluent release reports covering the operation of the unit shall be submitted prior to May 1 of each year and in accordance with the requirements of 10CFR50.36a and section IV.B.1 of 10CFR 50 Appendix I.

6.2.2 The Radioactive Effluent Release Report shall include:

6.2.2.1 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 Gaseous 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.

6.2.2.2 An annual summary of hourly meteorological data collected over the previous year. This annual summary may be in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. Alternatively, summary meteorological data may be retained and made available to the NRC upon request.

6.2.2.3 An assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. The historical annual average meteorology or 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 radiation doses shall be performed in accordance with this OFF SITE DOSE CALCULATION MANUAL (ODCM).

6.2.2.4 Identify those radiological environmental sample parameters and locations where it is not possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In addition, the cause of the unavailability of samples for the pathway and the new location(s) for obtaining replacement samples should be identified. The report should also include a revised figure(s) and table(s) for the ODCM reflecting the new location(s).

6.2.2.5 An assessment of radiation doses to the likely most exposed MEMBER, and 3.2.OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. The assessment of radiation doses shall be performed in accordance with this OFF SITE DOSE CALCULATION MANUAL (ODCM) Part II Sections 1.5, 2.4, 2.5 0929 I

CY-OC-170-301 Revision 4 Page 72 of 140 6.2.2.6 The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped Off Site during the report period (see Figure D-1-2):

a. Total volume shipped

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Type of waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms) 6.2.2.7 Unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents on a quarterly basis.

6.2.2.8 Changes to the PROCESS CONTROL PROGRAM (PCP) 6.2.2.9 Changes to the OFF SITE DOSE CALCULATION MANUAL (ODCM) in the form of a complete, legible copy of the ODCM.

6.3 RESPONSIBILITIES

6.3.1 Chemistry / Radwaste - Responsible for:

6.3.1.1 Implementing approval.

6.3.1.2 Compliance with specifications regarding routine dose assessment.

6.3.1.3 Radiological Environmental Monitoring Program 6.3.1.4 Technical consultation and review 6.3.2 Operations - Responsible for compliance with specifications regarding operation of the OCGS.

6.3.3 Engineering - Responsible for compliance with specifications regarding set point determination and implementation 6.3.4 Radiological Engineering - Responsible for technical consultation and review.

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CY-OC-170-301 Revision 4 Page 73 of 140 PART 1D- CALCULATIONAL METHODOLOGIES 1.0 LIQUID EFFLUENTS 1.1 RADIATION MONITORING INSTRUMENTATION AND CONTROLS The liquid effluent monitoring instrumentation and controls at Oyster Creek for controlling and monitoring normal radioactive material releases in accordance with the Oyster Creek Radiological Effluent Technical Specifications are summarized as follows:

(1) Alarm (Only) - The Reactor Building Service Water Effluent Line Monitor provides an Alarm function only for releases into the environment.

Liquid radioactive waste flow diagrams are presented in Figures D-1-la and D 1b.

1.2 LIQUID EFFLUENT MONITOR SET POINT DETERMINATION Per the requirements of CONTROL 3.3.3.10, alarm set points shall be established for the liquid monitoring instrumentation to ensure that the release concentration limits of CONTROL 3.11.1.1 are met (i.e., the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS at the U.S. route 9 bridge over the discharge canal shall not exceed the concentrations specified in 10 CFR 20 Appendix B. Table 2, Column 2, for radionuclides and 2.OE-04 IaCi/ml for dissolved or entrained noble gases).

1.2.1 LIQUID EFFLUENT MONITORS The set points for the liquid effluent monitors at the Oyster Creek Generating Station are determined by the following equation:

_ A F2 FLEC FI Where:

S = radiation monitor alarm set point (cpm)

A = activity concentration (pCi/ml) of sample in laboratory: A =1C g = the primary conversion factor for the instrument - the ratio of effluent radiation monitor counting rate to laboratory activity concentration in a sample of liquid (cpm per gCi/mL).

F1 = flow in the batch release line (e.g. gal/min). Value not greater than the discharge line flow alarm maximum set point.

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CY-OC-170-301 Revision 4 Page 74 of 140 F2 = flow in the discharge canal (e.g. gal/min). Value not less than the discharge canal minimum flow.

BKG = Monitoring instrument background (cpm)

FLEC = fraction or multiple of unrestricted area LEC in aqueous effluent based on sample analysis. FLEC is the ratio between the LECi and Ci. FLEC is unitless. For example:

LEC for Co-60 is 3E-6 ýtCi/mL. If the concentration in a expected release is 6E-6 gCi/mL; then FLEC is 6E-6/3E-6 = 2.

The term A represents the count rate of a solution having the same nuclide FLEC distribution as the sample and the LEC of that mixture.

Ci = concentration of radionuclide i in effluent, i.e., in a liquid radwaste sample tank, in reactor building service water (kCi/mL).

LECi = The unrestricted area liquid effluent concentration (LEC) of radionuclide i, i.e., 10 CFR 20, Appendix B, Table 2, Column 2 quantity for radionuclide i ([tCi/mL).

In the event gross radioactivity analysis alone is used to determine the radioactivity in an effluent stream or batch, FLEC is C/1E-8 (see 4.11.1.1.1),

Where:

C = The gross radioactivity concentration in effluent (pICi/mL).

1E-8 = The unrestricted area LEC for unidentified radionuclides (gCi/mL) from 4.11.1.1.1.

If the gross activity concentration, C, is below the lower limit of detection for gross activity, the value, 1E-8 gCi/mL, or the equivalent counting rate (cpm/mL) may be substituted for the factor

.A FLEC A = 1E-8 ItCi/mL FLEC 1.2.2 SAMPLE RESULT SET POINTS Usually, when the concentration of specific radionuclides is determinable in a sample(s), i.e., greater than the LLD, the alarm/trip set point of each liquid effluent radioactivity monitor is based upon the measurement of radioactive material in a batch of liquid to be released or in a continuous aqueous discharge.

1.2.3 ASSUMED DISTRIBUTION SET POINTS Alternatively, a radionuclide distribution that represents the distribution expected to be in the effluent if the concentration were high enough to be detectable, i.e.,

greater than the LLD, may be assumed. The representative distribution may be 0929 I

CY-OC-170-301 Revision 4 Page 75 of 140 based upon past measurements of the effluent stream or upon a computed distribution.

1.3 BATCH RELEASES A sample of each batch of liquid radwaste is analyzed for 1-131 and other principal gamma emitters or for gross beta or gross gamma activity before release. The result of the analysis is used to calculate the trip set point of the radioactivity monitor on the liquid radwaste effluent line to apply to release of the batch.

1.4 CONTINUOUS RELEASES The Reactor Building Service Water Effluent is sampled and analyzed weekly for 1-131 and other principal gamma emitters. Results of analyses for the preceding week or for a period as long as the preceding 3 months are used to calculate the alarm/trip set point of the corresponding effluent radioactivity monitor in order to determine a representative value. In each case, whether batch or continuous, the monitor alarm/trip set point may be set at lower activity concentration than the calculated set point.

1.5 LIQUID EFFLUENT DOSE CALCULATION - 10 CFR 50 Doses resulting from the release of radioiodines and particulates must be calculated to show compliance with Appendix I of 10CFR50. Calculations will be performed at least monthly for all liquid effluents as stated in SURVEILLANCE REQUIREMENT 4.11.1.2 and SURVEILLANCE REQUIREMENT 4.11.1.3.1 to verify that the dose to MEMBERS OF THE PUBLIC is maintained below the limits specified in CONTROL 3.11.1.2 The maximum dose to an individual from radioiodines, tritium, and radioactive particulates with half-lives of greater than eight days in liquid effluents released to unrestricted areas is determined as described in Reg. Guide 1.109. Environmental pathways that radioiodine, tritium, and particulates in liquid effluent follow to the maximally exposed MEMBER OF THE PUBLIC are assumed to be: exposure to shoreline deposits, ingestion of fish, and ingestion of shellfish. To assess compliance with CONTROL 3.11.1.2, the dose due to radioactive iodine, tritium, and particulates in liquid effluent is calculated to a person at the Route 9 bridge who consumes fish and shellfish harvested at that location.

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CY-OC-170-301 Revision 4 Page 76 of 140 1.5.1 MEMBER.OF THE PUBLIC DOSE - LIQUID EFFLUENTS CONTROL 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from Oyster Creek Generating Station to those listed in Table 1.5.1-1.

TABLE 1.5.1-1 LIQUID PATHWAY DOSE LIMITS During Any Calendar Quarter During Any Calendar Year

< 1.5 mrem to total body < 3.0 mrem to total body

< 5.0 mrem to any organ < 10.0 mrem to any organ Per the SURVEILLANCE REQUIREMENTS of 4.11.1.2, the following calculation methods shall be used for determining the dose or dose commitment due to the liquid radioactive effluents from Oyster Creek. Applicable liquid pathways to man for Oyster Creek include shoreline exposure, and ingestion of saltwater fish and shellfish. The receptor location is provided in Table A-4.

1.5.2 SHORELINE DEPOSIT DOSE The shoreline exposure pathway dose is calculated generally in the form (based on Reg. Guide 1.109):

Rapj = 110000 Ua*WM

QiTiDaipj(1- exp(-AiTb))

F Where:

110000 = a constant that accounts for time and flow conversions Rapj = the annual dose to organ j (including the total body), through pathway p, to age group a Uap = the age dependent usage factor for the specific pathway. Usage factors for shoreline exposure are residence time on the shoreline (hours). Usage factors are provided in Reg. Guide 1.109 Table E-5. Usage factors specifically selected for Oyster Creek are presented in Table B-I.

W = the shore width factor. This adjusts the infinite plane gamma or beta dose factors for the finite size and shape of the shoreline. Different factors apply to different bodies of water. A factor of 0.1 is used for OC for 'discharge canal bank'.

M = the recirculation factor. The recirculation factor is a multiplier of 3.76 to account for recirculation of discharge water back into the intake. Although this occurs infrequently, it is assumed to occur for each liquid release.

F = the flow rate in the discharge canal in cubic feet per second 0929 I

CY-OC-170-301 Revision 4 Page 77 of 140 Qi = the activity of the ith isotope in the release in curies Ti = the half life of the ith isotope in days Daipj = the age a, isotope i, pathway p, and organ j, specific dose conversion factor.

Pathway, isotope, age, and organ specific dose factors are obtained from Regulatory Guide 1.109 Appendix E, Tables E-6 through E-14 ki = the decay constant of the ith isotope in years Tb the long term buildup time, assumed to be 30 years Note: Xi and Tb can use any time units as long as they are both the same.

No transit delay (Tp from Reg. Guide 1.109) is assumed.

1.5.3 SHORELINE DOSE EXAMPLE The following provides an example of the liquid dose calculation:

Initial parameters:

Canal flow rate 1E6 gpm (typical of normal full power operation)

Release: 10,000 gallons of water at 1 E-3 gCi/ml Co-60 Problem: calculate shoreline whole body dose Uap = 67 (teenager) hours W =0.1 M = 3.76 F = 2228 [1 E6 gpm *3785 ml /gal /(60 sec/min

28316 ml/ft3) = 2228 CFS]

Qi = 0.03785 Ci [1E-3uCi/ml

10000gal

3785ml/gal = 0.03785 Cii Ti = 1930 [5.27 years*365.25days/yr = 1.93E3 days)

?1l = 1.31E-1 [0.693 / (5.27 yrs)]

Tb = 30 years Daipj = 1.7E-8 mrem/hr / pCi/m 2 Gamma dose factor Calculate Rapj for a = Teen, j = total body, p = shoreline dose for one isotope Rapj = I 10000 :-1 67*O.1"*37 Y.003785* 1930* 1.7E -8 * (1 - exp(-1.3E- 1*30))

2228 i Rapj = 1.5E-3 mrem: teen: wholebody 0929 I

CY-OC-1 70-301 Revision 4 Page 78 of 140 1.5.4 INGESTION DOSE - LIQUID Ingestion dose pathway calculations are similar to those for the shoreline dose, with minor changes in constants, removal of the shore width factor, and inclusion of the bioaccumulation factor:

Rap] = 1100 UapM QiBipDaipj Q

F Where:

Bip = the stable element bioaccumulation factor for pathway p for the ith isotope No transit delay is assumed Pathway, isotope, age, and organ specific dose factors are obtained from Regulatory Guide 1.109 Appendix E Tables E-7 through E-14. Bioaccumulation factors are provided in Reg. Guide 1.109 Table A-1. Usage factors are provided in Reg. Guide 1.109 Table E-5. Usage factors specifically selected for Oyster Creek are presented in Table B-1.

The radionuclides included in the periodic dose assessment per the requirements of CONTROL 3/4.11.1.2 are those as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of CONTROL 3/4.11.1.1, Table 4.11 .1 -.

Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Table 4.11.1.1.1-1.

1.5.5 INGESTION DOSE CALCULATION EXAMPLE The following provides an example of the liquid dose calculation:

Initial parameters:

Canal flow rate 1 E6 gpm (typical of normal full power operation)

Release: 10000 gallons of water at 1 E-3 I.Ci/mLI Co-60 Problem: calculate teen whole body dose from saltwater fish ingestion Rapj = 1100 UapM F QiBipDaipj F

Uap = 16 (teenager) Kg M = 3.76 F = 2228 [1E6 gpm *3785 ml / gal / (60 sec/min

28316 ml/ft3) = 2228 CFS]

Qi = 0.03785 Ci [1E-3uCi/mL

10000

3785 = 0.03785 Ci]

Bip = 100 Daipj = 6.33E-6 mrem / pCi 0929

CY-OC-1.70-301 Revision 4 Page 79 of 140 Calculate Rapj for a = Teen, j = total body, p = fish ingestion dose for one isotope Rapj = 1100 16

3.76 2228.03785 *100"633E 6 Rapj = 7.12 E - 4mrem teen wholebody 1.5.6 PROJECTED DOSE - LIQUID The projected doses in a 31 day period are equal to the calculated doses from the current 31 day period.

1.6 REPRESENTATIVE SAMPLES A sample should be representative of the bulk stream or volume of effluent from which it is taken. Prior to sampling, large volumes of liquid waste should be mixed in as short a time interval as practicable to assure that any sediments or particulate solids are distributed uniformly in the waste mixture. Recirculation pumps for liquid waste tanks (collection or sample test tanks) should be capable of recirculating at a rate of not less than two tank volumes in eight hours. Minimum recirculation times and methods of recirculation are controlled by specific plant procedures.

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CY-OC-170-301 Revision 4 Page 80 of 140 2.0 GASEOUS EFFLUENTS 2.1 RADIATION MONITORING INSTRUMENTATION AND CONTROLS The gaseous effluent monitoring instrumentation and controls at Oyster Creek for controlling and monitoring normal radioactive material releases in accordance with the Radiological Effluent CONTROLS are summarized as follows:

(1) Main Stack The main stack receives normal ventilation flow from the reactor building, new radwaste, old radwaste, process discharge flow from the augmented off gas system (AOG), condenser off gas flow if AOG is not in service, and normal ventilation flow from portions of the turbine building, predominantly the condenser bay area.

Reactor building and turbine building flow is not normally processed or filtered.

Reactor Building flow may be manually or automatically directed through the Standby Gas Treatment System (SBGTS) which has particulate and charcoal filtration. Off gas flow is processed through AOG or through a 30-minute delay pipe prior to release. Flow from the 'new' and 'old' radwaste buildings is HEPA filtered.

Releases through the main stack are monitored for noble gases using the RAGEMS I system and sampled for iodine, particulates and tritium. The plant stack is considered to be a true elevated release point.

(2) Turbine Building Vent The Turbine building vent is monitored for noble gases by the RAGEMS II system and sampled for iodine, particulates and tritium. It discharges on the west side of the turbine building approximately at roof height and is considered to be a ground level release. It ventilates the turbine floor and other areas of the turbine building.

Flow through this release point is not filtered.

(3) Feed Pump Room Vent The feed pump room vent is monitored by RAGEMS I1.It discharges on the east side of the turbine building below roof height and is considered to be a ground level release. It ventilates the reactor feed pump room. Flow through this release point is not filtered.

(4) Augmented Off Gas Building Vent Off gas Building HVAC is released through a ground level release from the building.

Off Gas process flow is not released through the building ventilation, but is routed To the stack plant. A ventilation monitoring system monitors for noble gas and samples for particulate and iodine.

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CY-OC-170-301 Revision 4 Page 81 of 140 (5) Isolation Condensers The isolation condensers are a ground level release. The predominant isotope through this potential release point is tritium as a consequence of the forced evaporation of condensate transfer water when the isolation condensers are initiated. Releases are neither monitored nor is the release process flow sampled.

Releases of tritium are evaluated based on liquid samples of the input and the volume used.

Gaseous radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation controls are presented in Figures D-2-1 and D-2-2.

2.2 GASEOUS EFFLUENT MONITOR SET POINT DETERMINATION 2.2.1 PLANT VENT Per the requirements of CONTROL 3.3.3.11, alarm set points shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of CONTROL 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem/year to the total body or 3000 mrem/year to the skin. Based on a grab sample analysis of the applicable release (i.e., of the Stack, Off gas process flow, etc.), the radiation monitoring alarm set points may be established by the following calculation methods. A set point of a monitor of an elevated release, e.g., from the stack, may be calculated using the equation:

S = 1.06 (CiDFSi) + Bkg where:

S = the alarm set point (cpm) h = primary conversion factor of the instrument - monitor response to activity concentration of effluent being monitored, cpm/(gjCi/cm 3 ). Each monitoring channel has a unique response, h, which is determined by the instrument calibration.

Ci = relative concentration of noble gas radionuclide i in effluent at the point of monitoring (pCi/cm 3 )

1.06 = 500 mrem/year /472 (conversion of cfm to cc/sec)

DFSi = factor converting elevated release rate of radionuclide i to total body dose equivalent rate at the location of potential exposure. Units are: mrem/(yr(jiCi/sec)).

From Table A-1.

f = flow of gaseous effluent stream being monitored, i.e., stack flow, vent flow, etc.

(ft 3/min) 0929

CY-OC-1 70-301 Revision 4 Page 82 of 140 BKG = Monitoring instrument background (cpm or mR/hr) 2.2.2 OTHER RELEASE POINTS The set point of a monitor of a ground-level or split-wake release, e.g., from the turbine building vent or the AOG building, may be calculated with the equation:

S= 1.06 _ h Bkg f (CDFM) +k X

Where:

DFVi = factor converting ground-level or split wake release of radionuclide i to the total body dose equivalent rate at the location of potential exposure. Units are:

mrem/(yr(j1Ci/m 3)). From Table A-i.

X/Q = atmospheric dispersion from point of ground-level or split-wake release to the location of potential exposure (sec/mi) from Table 2.2.2-1.

The atmospheric dispersion, X/Q, and the dose conversion factor, DFSi, depend upon local conditions. For the purpose of calculating radioactive noble gas effluent monitor alarm set points appropriate for the OCGS, the locations of maximum potential Off Site exposure and the reference atmospheric dispersion factors applicable to the derivation of set points are given in Table 2.2.2-1.

Symbols for this equation were defined in Section 2.2.1.

TABLE 2.2.2-1 RECEPTOR LOCATIONS AND DISPERSION FOR GASEOUS MONITOR SET POINTS Discharge Point Receptor Location Atm. Dispersion Sector Distance(m) (sec/m3)

Ground-level or vent SE 522 1.36 E-5 Stack SE 522 1.09 E-8 2.2.3 RADIONUCLIDE MIX FOR SET POINTS For the purpose of deriving a set point, the distribution of radioactive noble gases in an effluent stream may be determined in one of the following ways:

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CY-OC-170-301 Revision 4 Page 83 of 140 2.2.3.1.1 Preferably, the radionuclide distribution is obtained by gamma isotopic analysis of identifiable noble gases in effluent gas samples. Results of the analyses of one or more samples may be averaged to obtain a representative spectrum.

2.2.3.2 In the event a representative distribution is unobtainable from recent measurements by the radioactive gaseous waste sampling and analysis program, it may be based upon past measurements.

2.2.3.3 Alternatively, the total activity concentration of radioactive noble gases may be assumed to be Xenon-133 as found in Reg Guide 1.97.

2.3 GASEOUS EFFLUENT INSTANTANEOUS DOSE RATE CALCULATIONS -

10 CFR 20 2.3.1 SITE BOUNDARY DOSE RATE - NOBLE GASES CONTROL 3.11.2.1a limits the dose rate at the SITE BOUNDARY due to noble gas releases to < 500 mrem/yr, total body and < 3000 mrem/yr, skin. Radiation monitor alarm set points are established to ensure that these release limits are not exceeded. In the event any gaseous releases from the station results in an alarm set point (as determined in Section 2.2) being exceeded, an evaluation of the SITE BOUNDARY dose rate resulting from the release shall be performed.

2.3.1.1 TOTAL BODY DOSE RATE The total body dose equivalent rate from radioactive noble gases discharged from an elevated point (stack above building wake) is calculated with the equation:

DG= "QiP)8i i

From a ground-level release (building vent) the total body dose equivalent rate is:

x DG =-XZ oiPrVi Qv where:

DG = total body dose equivalent rate due to irradiation by radioactive noble gas (mrem/hr)

Qi = average discharge rate of noble gas radionuclide i released during the averaging time (tLCi/hr)

PyVi = factor converting time integrated ground-level concentration of noble gas nuclide i to total body dose mrem - m3. See Table A-2.

gCi - sec 0929 1

CY-OC-170-301 Revision 4 Page 84 of 140 X = atmospheric dispersion factor from the OCGS to the Off Site location of interest Qv (sec/M3) from Table 2.3.1.3-1 PySi = factor converting unit noble gas nuclide i stack release to total body dose at ground level received outdoors from the overhead plume (mrem/fpCi). See Table A-2 The noble gas plume gamma-to-total body dose factors, PySi at designated locations are derived from meteorological dispersion data with the USNRC RABFIN software computer code or similar computer program implementing Reg Guide 1.109, Appendix B. The noble gas semi-infinite cloud gamma-to-total body dose factors, PySi, are derived from Reg Guide 1.109, Revision 1, Table B-I, Column 5.

2.3.1.2 EXAMPLE TOTAL BODY DOSE RATE Calculate the dose from a release of 100 Ci of Xel 33 in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from a ground level release X

DG = - QiP yVi Qv X/Qv = 1.36E-5 sec/m3 (Table 2.3.1.3-1)

Qi = 1E8 gCi/hr [100Ci*1E6 gCi/Ci]

PWi = 9.33E-6 mrem-m 3 / gCi-sec DG =t.36E-5 lE8*9.33E-6 DG=0.013 mrem/hr 2.3.1.3 SKIN DOSE RATE The dose equivalent rate to skin from radioactive noble gases is calculated by assuming a person at ground level is immersed in and irradiated by a semi-infinite cloud of the noble gases originating in airborne effluent. It is calculated for each air effluent discharge point with the equation:

DB = X Z Qi(SBi+1.11AyVi)

Q .

where:

DB = dose rate to skin from radioactive noble gases (mrem/hr)

X = Atmospheric dispersions from gaseous effluent discharge point to ground-level location of interest (sec/M3 ) from Table 2.3.1.3-1.

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CY-OC-1 70-301 Revision 4 Page 85 of 140 Qi = discharge rate of noble gas radionuclide i (ltCi/hr)

SBi = factor converting time integrated ground-level concentration of noble gas radionuclide i to skin dose from beta radiation mrem - m3 from Table A-2.

gCi

sec AVi = factor for converting time integrated, semi-infinite 3

concentration of noble gas radionuclide mrad - m i to air dose from its gamma

s from Table A-2.

gCi

sec The noble gas beta radiation-to-skin-dose factors, SBi and the noble gas gamma-to-air dose factors, AWV, are derived from Reg Guide 1.109, Revision 1, Table B-i, columns 3 and 4 respectively. A tabulation of these factors used to compute noble gas-to-dose equivalent rate at 522 meters SE of the OCGS is in Table A-2.

The dose equivalent rate is calculated with the meteorological dispersion data given in Table 2.3.1.3-1.

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CY-OC-170-301 Revision 4 Page 86 of 140 TABLE 2.3.1.3-1 RECEPTOR LOCATIONS AND DISPERSION FOR SITE BOUNDARY DOSE RATES Discharge Point Receptor Location Atm. Dispersion Sector Distance (m) (sec/m 3 )

Ground Level SE 522 1.36 E-5 or Vent Stack SE 522 1.09 E-8 Alternatively, an approved computer code (e.g., "SEEDS" or "Open EMS") that implements the requirements of Regulatory Guide 1.109 may be used.

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

2.3.1.4 EXAMPLE SKIN DOSE RATE Calculate the skin dose from a release of 100 Ci of Xe133 in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from a ground level release:

DB X Qi (SBi ++1.11ArVi)

Q=.

X/Q = 1.36 E-5 sec/m3 Qi = 1E8 jICi/hr SBi = 9.71 E-6 AyVi = 1.12E-5 DB = 1.36E 5X1E8(9.71E - 6 + 1.11 *1.12E -5)

DB = O.O030mrad/ hr 2.3.2 SITE BOUNDARY DOSE RATE - RADIOIODINE AND PARTICULATES 2.3.2.1 METHOD - SITE BOUNDARY DOSE RATE - RADIOIODINE AND PARTICULATES The dose rate Off Site due to the airborne release of 1-131, 1-133, tritium, and particulates with half-lives greater than 8 days is limited to no more than 1500 mrem/yr to any organ in CONTROL 3.11.2.1b. Evaluation of compliance with CONTROL 3.11.2.1b is based on the sampling and analyses specified in TABLE 4.11.2.1.2-1. Since the dose rate cannot be resolved within less than the sample integration or compositing time, the contribution of each radionuclide to the calculated dose rate will be averaged no more than 3 months for H-3, Sr-89, Sr-90, and alpha-emitting radionuclides and no more than 31 days for other radionuclides. These are their usual sample 0929 1

CY-OC-1 70-301 Revision 4 Page 87 of 140 integration or compositing times. The equation used to assess compliance of radioiodine, tritium, and radioactive particulate releases with the dose rate limit is:

DRp = IE6 RaDFAUaQe i-X e i Qe where:

I E6 = conversion pCi/pCi DRp = the average dose rate to an organ via exposure pathway, p (mrem/yr).

DFAija= inhalation dose factors due to intake of radionuclide i, to organ j age group a (mrem/pCi) from Reg. Guide 1.109 Appendix E.

Ra = age group dependent inhalation respiratory rate (usage factor) m3/yr from Table B-1

- annual average relative airborne concentration at an Off Site location due to a release Qe from either the Stack or a vent, i.e. release point, e (sec/m 3) from Table 2.3.2.1-1.

Qei = release rate of radionuclide i from release point, e during the period of interest (pCi/sec).

For real-time meteorology and on an annual average basis, the location of the maximum ground-level concentration originating from a vent release will differ from the maximum ground-level concentration from a stack release. When assessing compliance with CONTROL 3.11.2.1 b for tritium, iodine, and particulate, the air dispersion (X/Q) values are provided in Table 2.3.2.1-1.

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CY-OC-170-301 Revision 4 Page 88 of 140 TABLE 2.3.2.1-1 LOCATION OF MAXIMUM EXPOSURE RE BY INHALATION Discharge Point Receptor Location Atm. Dispersion Sector Distance (m) (sec/m3)

Ground Level or Vent SE 522 1.36 E-5 Stack SE 522 1.09 E-8 Alternatively, inhalation exposure to effluent from the stack may be evaluated at the closest hypothetical individual located at:

Stack SE 966m 1.19 E-8 Alternatively, an approved computer code (e.g., "SEEDS" or "Open EMS") that implements the methods of Regulatory Guide 1.109, may be used.

2.3.2.2 EXAMPLE IODINE AND PARTICULATES DOSE RATE CALCULATION Calculate the child thyroid dose rate from a release of 100 gCi/hr of 1131 from a ground level release:

DRp = IE6_SeY RaDFAi'aQeiX-i Qe Ra = 3700 m 3/yr DFAija= 4.39E-3 Qei = 0.028 ptCi/sec 11 OOpCi/hr /3600 sec/hr = 0.02778]

X/Qe = 1.36 E-5 DRp = 1E6Z 3700

4.39E - 3

0.028

1.36E - 5 DRp = 6 .2mrem / yr 2.4 NOBLE GAS EFFLUENT DOSE CALCULATIONS - 10 CFR 50 Doses resulting from the release of noble gases must be calculated to show compliance with Appendix I of 10CFR50. Calculations will be performed at least monthly for all gaseous effluents as stated in SURVEILLANCE REQUIREMENT 4.11.2.2 to verify that the dose to air is kept below the limits specified in CONTROL 3.11.2.2 and the dose to MEMBERS OF THE PUBLIC is maintained below the limits specified in CONTROL 3.11.2.3.

2.4.1 UNRESTRICTED AREA DOSE - NOBLE GASES CONTROL 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly air dose limits shown in Table 2.4.1-1.

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CY-OC-170-301 Revision 4 Page 89 of 140 TABLE 2.4.1-1 ANNUAL AIR DOSE LIMITS Duringq any calendar quarter Duringq any calendar year

<5 mad gamma-air _510 mad gamma-air

_<10 mad beta-air <20 mad beta-air The method used to calculate the air dose at the critical location due to noble gas is described by the following equations. The limits are provided in CONTROL 3.11.2.2 for air dose Off Site due to gamma and beta radiations from effluent noble gas.

2.4.1.1 AIR DOSE METHOD a: For Gamma Radiation:

n x Dose r : A "Vi -vQvi + A ySiQsi Q

b: For Beta Radiation Dose#=Z JAeQei where:

Dose y =the gamma dose during any specified time period (mrem).

Dose P=the beta dose during any specified time period (mrad).

AyVi = the air dose factor due to ground level gamma emissions for each identified noble gas radionuclide, i; (mrad/yr per ttCi/m 3). Table A-2 AySi = the factor for air dose at ground level due to irradiation for an airborne plume resulting from a Stack release (mrad per jiCi), Table A-3.

AN31 = the air dose factor due to beta emissions for each identified noble gas radionuclide, i (mrad/yr per jtCi/m 3). Table A-3 Xor =the annual average relative concentration for areas at or beyond the site Q Q boundary for releases from either the Stack or ground vent at the critical location (sec/m 3), Table 2.4.1.1-1 Qvi = amount of radionuclide i released from vents (gCi).

Qsi = amount of radionuclide i released from the Stack (jiCi).

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CY-OC-170-301 Revision 4 Page 90 of 140 Qei = amount of radionuclide i released from release point e(pCi).

Noble gases may be released from the ground level vents and stack. The quantity of noble gas radionuclides released will be determined from the continuous noble gas monitors and periodic isotopic analyses. The maximum Off Site gamma radiation dose rate to air from noble gases discharged from either the stack or from building vents occurs at 522 meters SE of the OCGS. Values of AySi depend upon the meteorological conditions and the location of exposure and are calculated using the NRC RABFIN code or similar one in accordance with Reg. Guide 1.109, Appendix B, Section 1. AyVi and ABi are derived from Reg. Guide 1.109, Table B-1 for a semi-infinite cloud, independent of meteorology or location. Values of AySi, AyVi and ABi used to calculate the noble gas radiation dose to air at 522 meters SE of the OCGS are in Table A-3. Reference atmospheric dispersion from the OCGS to 522 meters SE is given in Table 2.4.1.1-1.

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CY-OC-170-301 Revision 4 Page 91 of 140 TABLE 2.4.1.1-1 RECEPTOR LOCATIONS AND DISPERSION FOR AIR DOSE Discharge Point Receptor Location Atm. Dispersion Sector Distance (i) (sec/m3)

Ground Level SE 522 1.36 E-5 or Vent Stack SE 522 1.09 E-8 Alternatively, an approved computer code (e.g., "SEEDS" or "Open EMS") that implements the requirements of Reg. Guide 1.109 may be used.

2.4.1.2 EXAMPLE NOBLE GAS AIR DOSE CALCULATION Calculate the gamma air dose from a release of 1 Ci per hour of Xe133 for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months from a ground level release and 100Ci per hour for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months from an elevated release:

Dosey = A Wi-v Qvi + A )SiQsi H

3 Q

A*,Vi = 1.12E-5 mrad - m /iCi - sec X/Q = 1.36 E-5 Qvi = 1E7 ptCi [lCi/hr*10hrs*1E6 ltCi/Ci]

AySi = 5.45E-1 3 mrad / iCi Qsi = 1E9 lCi [100Ci/hr*10hrs*1E6 lCi/Ci]

n Dosey= 1.1 2E-5*1.36E-5*lE7+5.45E-1 3*lE9 i=1 n

Dosey= Z1.63E-3-3+5.45E-4 i=1 Dosey = 2.2E - 3mrad Note how the ground level portion has a higher dose contribution per unit activity than the elevated portion.

2.4.1.3 INDIVIDUAL PLUME DOSE METHOD The method for dose to an individual from noble gases is essentially identical with the air dose method except that different dose factors apply. Also, since dose to the skin combines the contribution from gamma and beta emissions, the gamma dose must be added to the beta dose to obtain a total skin dose.

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CY-OC-170-301 Revision 4 Page 92 of 140 a: For Total Body:

Doset= ZP 1=l Ai-vQvi+

Q PTSiQsi b: For Skin Doses = S/Ji- eQei + Doset e j=1 Q where:

Doset the total body dose during any specified time period (mrem).

Doses = the skin dose during any specified time period (mrad).

PyV 1 = the plume dose factor due to ground level gamma emissions for each identified noble gas radionuclide, i; (mrad/yr per gtCi/m 3 ). Table A-5 PyS' = the factor for plume dose at ground level due to irradiation for an airborne plume resulting from a Stack release (mrad per jiCi), Table A-5.

Spi3 = the skin dose factor due to beta emissions for each identified noble gas radionuclide, i (mrad/yr per gCi/m 3 ) from Table A-5.

Xor = the annual average relative concentration for areas at or beyond the site Q Q boundary for releases from either the Stack or ground vent at the critical location (sec/mr3 ) from Table 2.5.1.

Qvi = amount of radionuclide i released from vents (gCi).

Qsi = amount of radionuclide i released from the Stack (jiCi).

Q=i = amount of radionuclide i released from release point e (lRCi).

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CY-OC-170-301 Revision 4 Page 93 of 140 2.5 RADIOIODINE, PARTICULATE AND OTHER RADIONUCLIDES DOSE CALCULATIONS -

10 CFR 50 Doses resulting from the release of radioiodines and particulates must be calculated to show compliance with Appendix I of 10CFR50. Calculations will be performed at least monthly for all gaseous effluents as stated in SURVEILLANCE REQUIREMENT 4.11.2.2 and SURVEILLANCE REQUIREMENT 4.11.2.3 to verify that the dose to air is kept below the limits specified in CONTROL 3.11.2.2 and the dose to MEMBERS OF THE PUBLIC is maintained below the limits specified in CONTROL 3.11.2.3.

The maximum dose to an individual from radioiodines, tritium, and radioactive particulates with half-lives of greater than eight days in gaseous effluents released to unrestricted areas is determined as described in Reg. Guide 1.109. Environmental pathways that radioiodine, tritium, and particulates in airborne effluent follow to the maximally exposed MEMBER OF THE PUBLIC as determined by the annual land use survey and reference meteorology will be evaluated. The seasonality of exposure pathways may be considered.

For instance, if the most exposed receptor has a garden, fresh and stored vegetables are assumed to be harvested and eaten during April through October. Fresh vegetables need not be considered as an exposure pathway during November through March. To assess compliance with CONTROL 3.11.2.3, the dose due to radioactive iodine, tritium, and particulates in airborne effluent is calculated to a person residing 966 meters SE of the OCGS. Reference atmospheric dispersion and deposition factors are given in Table 2.5-1.

TABLE 2.5-1 DISPERSION FOR 10CFR50 DOSES Discharge Point Dispersion Deposition X/Q (sec/m3 ) D/Q(I/m 2 )

Ground Level or Vent 4.86 E-6 1.41 E-8 Stack 1.19 E-8 1.74 E-9 The environmental pathways of exposure to be evaluated are: inhalation, irradiation from ground deposition, and ingestion of milk (cow and goat are treated separately), meat, and vegetables. Eight organs are considered: Bone, Liver, Total Body, Thyroid, Kidney, Lung, GI-LLI (Gastro-Intestinal tract / Lower Large Intestine), and Skin. Four different age groups are considered: Infants, Children, Teens, and Adults. Doses are calculated to a 'receptor'- a person who inhales the airborne activity and resides in a location with ground deposition, and eats and drinks the foodstuffs produced. The maximally exposed individual is conservatively assumed to reside at the location of the highest sum of the inhalation and ground plane doses, while eating and drinking foodstuffs transported from the locations that are highest for those pathways. Receptor locations are provided in Table A-4.

Alternatively, an approved computer code (e.g., "SEEDS" or "Open EMS") that implements the requirements of Reg Guide 1.109 may be used.

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CY-OC-170-301 Revision 4 Page 94 of 140 2.5.1 INHALATION OF RADIOIODINES, TRITIUM, PARTICULATES, AND OTHER RADIONUCLIDES.

Dose from the inhalation pathway is generally in the form:

Da=RaTZ!-QiDFAijaExp(-AiTr) iQ Where:

Dja = the dose to the organ j (of eight) of age group a (of four)

Ra = the respiration rate for age group a from Table B-1 T = the duration of the release in fraction of a year X = The atmospheric dispersion to the point of interest (the 'receptor') in sec/r 3 from Table 2.5-1 Qi = The release rate of radionuclide i (pCifsec)

DFAija= The inhalation dose conversion factor (mrem per pCi) for radionuclide i to organ j of age group a from Reg. Guide 1.109 Appendix E.

Xi = decay constant of isotope i: 0.693/ Half life in years Tr = plume transit time from release to receptor in years Xi and Tr may be in any time units as long as they are the same Note that a 'depleted X/Q' (dX/Q) is applicable to particulates only, which accounts for the natural settling and lack of surface reflection of particulates to estimate the downwind concentration accounting for these removal processes. Depleted X/Q will be slightly smaller than the X/Q. This is not used in the ODCM for simplicity. Using the X/Q is therefore slightly conservative compared to the dX/Q.

0929 I

CY-OC-170-301 Revision 4 Page 95 of 140 2.5.2 EXAMPLE CALCULATION - INHALATION OF RADIOIODINES, TRITIUM, PARTICULATES, AND OTHER RADIONUCLIDES Calculate the dose to child lung from inhalation from a ground level release of 100 kCi of Co-60 in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Plume transit decay time is ignored (exp(-l-iTr)=1).

Dja = 3700* 0.00114*1.53E - 5

2.78E3

1.91E - 3 Dja = 3.4E - 4mrem Dja = RaT> QiDFAija Dja = the dose to the organ j (of eight) of age group a (of four)

Ra = 3700 m 3lyr T = 0.00.114 yrs [10 hrs / 8760 hrs / yr]

X- = 1.53 E-5 sec/m3 Qi = 2.78E3 pCi/sec [100 DCi

1E6 pCi/iCi / (10 hrs*3600 sec / hr)]

DFAija=1.91E-3 mrem / pCi 2.5.3 INGESTION OF RADIOIODINES, PARTICULATES AND OTHER RADIONUCLIDES Dose from the ingestion pathways is more complex and is broken out here into multiple steps:

2.5.3.1 CONCENTRATION OF THE RADIONUCLIDE INANIMAL FORAGE AND VEGETATION - OTHER THAN TRITIUM The concentration of a radionuclide in a foodstuff (other than tritium - see section 2.5.3.3 for tritium) is dependent on the atmospheric deposition, the biological uptake into the food, various decay times (plume travel, harvest to table, etc.) and is generally of the form:

Where:

Civ = the concentration (pCi/kg) of radionuclide i in vegetation Qi = the release rate of isotope i in pCi/hr D - The atmospheric deposition to the point of interest (the 'receptor') in I/m2 from Q

Table 2.5-1.

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CY-OC-1 70-301 Revision 4 Page 96 of 140 Civ =D Qi r(I - EXP(-2EiTe)) + Biv(1 - EXP(-2iTb)) EXP(-AiTh)EXP(-AiTr) r= the retention coefficient for deposition onto vegetation surfaces (1.0 for iodines, 0.2 for particulates)

Ji = the decay constant of radionuclide i; 0.693/half life in hours XEi = the effective removal constant which is the sum of Xi + Xw where Xw is the weathering constant, 0.0021/hr Te = duration of crop exposure during the growing season in hours. This is not the entire duration of the growing season, and is different for leafy vegetable and fruit/grain/vegetables. Provided in Table E-15 of Reg. Guide 1.109 or Table B-1.

2 2 Yv = agricultural yield Kg of vegetation per m , typically 0.7 kg/M Biv = soil uptake concentration factor for transfer of the radionuclide i from the soil to the vegetation through normal root uptake processes in pCi/kg in vegetation per pCi/Kg in soil. Values are provided in Reg. Guide 1.109 Table E-1.

Tb = the length of time the soil is exposed to contaminated inputs - nominally 30 years (2.63E5 hr) 2 2 P = effective soil density in kg/mr normally 240 kg/m Th = holdup time, the time the foodstuff is in transit between harvest and consumption in hours Tr = plume transit time from release to receptor in hours 2.5.3.2 EXAMPLE CALCULATION OF CONCENTRATION OF THE RADIONUCLIDE IN ANIMAL FORAGE AND VEGETATION - OTHER THAN TRITIUM.

Calculate the forage and vegetation concentration from a ground level release of 100 RCi of Co-60 in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (plume transit time is ignored Tr=O, EXP(-XiTr)=I ):

2 D/Q = 1.41 E-8m Qi = 1E7 pCi/hr [100ltCi

1E6 pCi/gCi /10 hr]

r = 0.2 Xi = 1.5E-5/hr [0.693 / (5.27yr

8760 hr/yr)]

XEi = 2.1E-3/hr [1.5E-5 + 0.0021]

Te = 720 hr [grass-cow-milk-man pathway value]

Yv = 0.7 kg/m2 Biv = 9.4E-3 0929

CY-OC-170-301 Revision 4 Page 97 of 140 Civ = D Qi{r(1 - EXP(-2EiTe))

Yv cEi P+ }

Biv(l - EXP(-bTh)) EXP(-AiTh)EXP(-2AbTr) 0.2 * (1- EXP(-2. IE- 3

720)) +

Civ=1.41E-8*1E7 0.7*2.5E-3 5E5) (-1.5E-5'0) 240*1.5E-5 0.2 * (1- EXP(-1.52)) +

Civ =1.41E - 8 *1E7 1.47E3EXP(-O) 9.4E - 3* (1- EXP(-3.95))

3.6E - 3 06.

Civ =1.41E - 1 2.56 Civ = 15.3 pCi / Kg Tb = 2.63E5 hr P = 240 kg/m2 Th = 0 hr (consumption of pasture grass directly by animals) 2.5.3.3 CONCENTRATION OF TRITIUM INANIMAL. FORAGE AND VEGETATION Since tritium is assumed to be released as tritiated water (HTO), the concentration of tritium in a foodstuff is dependent on atmospheric dispersion like a gas, rather than particulate deposition as for other. radionuclides for foodstuff uptake. Further, the concentration of tritium in food is assumed to be based on equilibrium between the concentration of the tritium in the atmospheric water and the concentration of tritium in the water in the food. Concentration of tritium in vegetation can be calculated generally in the form (a plume transit decay term: EXP(-EliTr) is ignored since plume travel times are very short compared to the half life):

X 0.5 Ctv = 1000Qt X* 0.75

Q H Where:

Ctv = the concentration (pCi/kg) of tritium in vegetation 0929 I

CY-OC-170-301 Revision 4 Page 98 of 140 1000= g per kg Qt = the release rate of the tritium in pCi/ sec X/Q = the atmospheric dispersion at the vegetation point, sec/m 3 from Table 2.5-1 0.75 = the fraction of vegetation that is water 0.5 = the effective ratio between the atmospheric water concentration and the vegetation concentration H = the absolute humidity g/m 3 . Absolute humidity is seasonally dependent, varying from as little as 1 in the winter to as much as 20 in the summer. Monthly average values derived from historical data are provided in Table B-2.

2.5.3.4 EXAMPLE CALCULATION OF CONCENTRATION OF TRITIUM INANIMAL FORAGE AND VEGETATION.

Calculate the forage and vegetation concentration from a ground level release of 100 jICi of H-3 in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Plume transit decay time is ignored (exp(-XiTr)=1):

X 0.5 Ctv= 10000tX.0.75.

Q H Qt = 2778 pCi/sec 3

[100uCi

1E6 pCi/uCi I (10hrs*3600sec/hr)]

X/Q = 4.86E-6 sec/m H = 5 g/m 3 (assumed for this example) 0.5 Ctv = 2778

1000

4.86E - 6

0.75

5 Ctv= 1.OpCi/kg 2.5.3.5 CONCENTRATION OF THE RADIONUCLIDE INMILK AND MEAT Meat and milk animals are assumed to eat both pasture grass and stored feed.

During a fraction of the year, they may be assumed to be exclusively on stored feed, outside of the growing season. If using annual average release, the fraction of stored and fresh feed must be accounted for with fractions, otherwise (as in this ODCM), the fresh pasture pathway is turned on or off depending on the growing season.

The concentration of a radionuclide in the animal feed is calculated as follows:

Civ = FpCis+ (1- Fp)Cis(1- Fs)+ CipFs(1 - Fp) 0929 I

CY-OC-1 70-301 Revision 4 Page 99 of 140 Where:

Fp = the growing season pasture factor: 1 if not growing season, 0 if in growing season Fs = the fraction of the daily feed from fresh pasture from Table B-1 or Exhibit E-15 from Reg. Guide 1.109.

Cip = the. concentration in the fresh pasture feed (Civ from section 2.5.3.2 with Th = 0 for immediate consumption)

Cis = the concentration in stored feed (Civ from section 2.5.3.2 with Th = 90 days)

The concentration in the milk is then based on this feed concentration:

Cim = FmCivQJEXP(-1IiTf)

Where; Cim = the concentration in milk pCi/I Fm = the transfer coefficient of intake to concentration in the milk (d/i) from Reg. Guide 1.109 Table E-1.

Qf = feed intake rate Kg/d from Reg. Guide 1.109 Table E-3.

Xi = radionuclide i decay constant in 1/days Tf = transport time from milk production to consumption (2 days for milk)

The Goat milk pathway may be similarly evaluated:

Cim = FgCivQJfXP (-2iTf)

Where Fg = the transfer coefficient of intake to concentration in the milk (d/I) for goats from Reg. Guide 1.109 Table E-2.

And for meat:

Of= FfCivQJEXP(-AiTs)

Where:

Ff = the transfer coefficient of intake to concentration in the meat d/kg from Reg.

Guide 1.109 Table E-1.

Ts = The transport time from slaughter to consumption (20 days) 0929 1

CY-OC-170-301 Revision 4 Page 100 of 140 2.5.3.6 EXAMPLE CALCULATION OF CONCENTRATION OF THE RADIONUCLIDE IN MILK AND MEAT Calculate the concentration in cow milk from a ground level release of 100 RiCi of Co-60 in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Plume transit decay time is ignored (exp(-XiTr)=l):

Civ = FpCis + (1- Fp)Cis(1 - Fs) + CipFs(1 - Fp)

Assume animals are on pasture and receive half of their food from stored feed.

Cip = 18.4 pCi/kg as previously calculated in section 2.5.3.2 Fp =0 Fs = 0.5 Cis is calculated by applying a 90 day decay term to the Cip value previously calculated, since the previous decay correction was for 0 time as shown in 2.5.3.2.

Cis - 18.4 * (exp(-0.693

90/(5.27

365.25)))

Cis = 17.8pCi / kg Civ is then:

Civ =0*17.8+ (1-0.5)17.8* (1-0)+ 18.4 *.5* (1- 0)

Civ = 18.1pCi/ kg The concentration in milk is given by:

Cim = FmCivQJEXP(-2iTf)

Fm =1.OE-3 d/I Qf = 50 Kg/d xi = 3.6E-4/d [0.693 / (5.27yrs*365.25days/yr)]

Cim = 1.0E- 3

18.1 *50 *EXP(-3.6E- 4 2)

Cim = 0.90pCi / l The concentration in meat given by:

Cif = FfCivQfEXP (-,iTf)

Ff =1.3E-2 d/kg Qf = 50 Kg/d Ai = 3.6E-4/d Cif = 1.3E- 2 18.1

50

EXI(-3.6E- 4 20)

Cif = 11.7pCilkg 0929 1

CY-OC-170-301 Revision 4 Page 101 of 140 2.5.3.7 DOSE FROM CONSUMPTION OF MILK, MEAT, AND VEGETABLES The environmental pathway ingestion dose is the sum of the milk, meat, and vegetation ingestion pathways. There are two separate pathways for vegetation: fresh leafy vegetables and a combination of fruits, non-leafy vegetables, and grains. These differ only in the decay and buildup processes applied to account for the environmental exposure, and transportation delay decay represented by Te and Th as shown in section 2.5.3.1. For long half-life isotopes (e.g. Co-60) the decay differences have little impact on the dose.

Dose from the environmental ingestion pathways is generally of the form:

Dja = T> DFIj'a[UavFgCiv+UamCim + UafCif + UalFICil]

Where:

Dja = the dose to organ j of age group a - mrem T = fraction of year of release duration DFlija = the ingestion dose factor for isotope i to organ j for age group a - mrem/pCi from Reg.

Guide 1.109 Appendix E Uav = Ingestion rate (usage factor) for non-leafy vegetables, grains, and fruits for age group a from Reg. Guide 1.109 Table E-5 or Table B-I.

Fg = the fraction of vegetables, grains, and fruits from the location of interest : 0.76 in Reg.

Guide 1.109.

Civ = the concentration of isotope i in the vegetables, fruits, and grains calculated from section 2.5.3.2.

Uam = Ingestion rate (usage factor) for milk for age group a: from Table B-1 or Reg. Guide 1.109 Table E-5.

Cim = the concentration of isotope i in milk calculated from section 2.5.3.5.

Uaf = the ingestion rate for meat for age group a: from Table B-1 or Reg. Guide 1.109 Table E-5.

Cif = the concentration of isotope i in meat calculated from section 2.5.3.2.

Ual = the ingestion rate for leafy vegetables for age group a: from Table B-1 or Reg.

Guide 1.109 Table E-5.

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CY-OC-170-301 Revision 4 Page 102 of 140 FI = the fraction of annual leafy vegetable ingestion from the location of interest : 1.0 in Reg. Guide 1.109.

Cil = concentration of isotope i in the leafy vegetables for direct human consumption:

Civ calculated from section 2.5.3.2 with Th=0.

2.5.3.8 EXAMPLE CALCULATION - DOSE FROM CONSUMPTION OF MILK, MEAT, AND VEGETABLES Calculate the ingestion dose to child whole body from a ground level release of 100 p.Ci of Co-60 in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Plume transit decay time is ignored (exp(-kiTr)=1):

Dja = Tj DFIYja[UavFgCiv+UamCim+ UafCif+ UalFlCi[]

i T = .00114 [10hrs / 8760 hrs/yr)

DFlija= 1.56E-5 rnrem/pCi Uav = 520 Fg = 0.76 Civ = 18.0 [18.4*EXP(-k*60) using 60 day delay for ingestion]

Uam = 330 Cim = 0.9 Uaf = 41 Cif = 11.7 Ual = 26 FI =1 Cil = 18.4 Dja=.00114Y1.56E-5[520*0.76*18+ 330*0.9+41*11.7 + 26*l*18.4]

Dja =.001 14X1.56E- 5[7114+ 297 +480+478]

Dja = 1.5E - 4mrem: child: wholebody 2.5.4 GROUND PLANE DEPOSITION IRRADIATION Dose from ground plane deposition is estimated by determining the surface activity resulting from the release.

2.5.4.1 GROUND PLANE CONCENTRATION The ground surface activity is estimated as:

=Qi Cig =-'-(-1.- EXP(-AiTb))

Where:

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CY-OC-170-301 Revision 4 Page 103 of 140 2

Cig = ground plane concentration of radionuclide i in pCi/mr D_-D local atmospheric release deposition factor in 1/M2 from Table 2.5-1 Q

Qi = release rate in pCi/sec i = radiological decay constant in 1/sec Tb = long term buildup time 30 years (9.46E8 sec)

Note: Qi, Xi and Tb can utilize any time units as long as they are all the same 2.5.4.2 EXAMPLE GROUND PLANE CONCENTRATION CALCULATION Calculate the ground plane concentration from a 100 igCi release of Co-60 over 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months from a ground level release point.

Cig D Qi (1- EXP(-AiTb))

D A

- = 1.41 E-8/m 2 Q

Qi- = 2778 pCi/sec [1001tCi/10hrs/3600sec/hr]

Xi = 4.17E-9/sec [0.693/(5.27yr*8760hr/yr*3600sec/hr)]

Tb = 9.46E8 sec Cig = 1.41E - 8 2778(1 - EXP(-4.17E - 9

9.46E8))

4.17E-9 Cigzl41E-82778 Cig = 1.41E _ 48 4.17E- 9 (7- EXP(-4.17E - 9

9.46E8))

Cig = 1.1E4pCi/rm2 2.5.4.3 GROUND PLANE DOSE Annual dose from the ground plane deposition is of the form:

Djg = 8760

T

SfE CigDFGUi Where:

Djg the annual dose (mrem) from ground plane pathway (g) to the total body or skin ()

8760 = hours in a year 0929

CY-OC-170-301 Revision 4 Page 104 of 140 T = fraction of year release is in progress

,Sf = shielding factor accounting for shielding from dwelling from Table B-1 DFGij = Ground plane dose factor for skin or total body () for radionuclide i from Table E-6 of Reg. Guide 1.109 in mrem/hr / pCi/m 2 .

2.5.4.4 EXAMPLE GROUND PLANE DOSE Calculate the ground plane Total Body dose from a 100 IaCi release of Co-60 over 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months from a ground level release point.

Given: Djg = 8760

T

Sf , CigDFGij T = 0.00114 [10/8760]

Sf = 0.7 DFGij = 1.7E-8 Djg = 8760*0.001 14*0.7-9729*1.7E-8 Dig =1.1 5E -3mremTotalBody 2.6 PROJECTED DOSES - GASEOUS The projected doses in a 31 day period are equal to the calculated doses from the current 31 day period.

3.0 TOTAL DOSE TO MEMBERS OF THE PUBLIC - 40 CFR 190 The Radiological Environmental Monitoring Report (REMP) submitted by May 1st of each year shall include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from effluents and direct radiation from on-site sources). For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Oyster Creek, the sources of exposure need only consider the Oyster Creek Generating Station. No other fuel cycle facilities would contribute significantly to the MEMBER OF THE PUBLIC dose for the Oyster Creek vicinity, however, both plant operation and ISFSI sources must be included in the dose assessment.

To assess compliance with CONTROL 3.11.4, calculated organ and total body doses from effluents from liquid pathways and atmospheric releases as well as any dose from direct radiation will be summed.

As appropriate for demonstrating/evaluating compliance with the limits of CONTROL 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for providing data on actual measured levels of radiation and / or radioactive 0929 1

CY-OC-170-301 Revision 4 Page 105 of 140 material and resultant dose to the MEMBER OF THE PUBLIC in the actual pathways of exposure.

3.1 EFFLUENT DOSE CALCULATIONS For purposes of implementing the surveillance requirements of CONTROL 3/4.11.4 and the reporting requirements of Technical Specification 6.9.1.d (RERR), dose calculations for the Oyster Creek Generating Station may be performed using the calculation methods contained within the ODCM; the conservative controlling pathways and locations from the ODCM or the actual pathways and locations as identified by the land use census (CONTROL 3/4.12.1) may be used. Average annual meteorological dispersion parameters provided herein or meteorological conditions concurrent with the release period under evaluation may be used.

3.2 DIRECT EXPOSURE DOSE DETERMINATION Any potentially significant direct exposure contribution to off-site individual doses may be evaluated based on the results of environmental measurements (e.g., TLD) and/or by the use of radiation transport and shielding calculation methodologies.

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of CONTROL 3.12.1. The objectives of the program are:

- To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the vicinity of Oyster Creek

- To determine if the operation of the Oyster Creek Generating Station has resulted in any increase in the inventory of long lived radionuclides in the environment;

- To detect any changes in the ambient gamma radiation levels; and

- To verify that OCGS operations have no detrimental effects on the health and safety of the public or on the environment.

The REMP sample locations are presented in Appendix E.

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CY-OC-170-301 Revision 4 Page 106 of 140 APPENDIX A - DERIVED DOSE FACTORS AND RECEPTOR LOCATIONS 0929 1

CY-OC-170-301 Revision 4 Page 107 of 140 Table A-1 Dose Conversion Factors for Deriving Radioactive Noble Gas Radionuclide-to-Dose Equivalent Rate Factors*

Radionuclide Factor DFSi for Factor DFVi for Stack Release* Ground-level or Split-Wake Release**

mrem-sec mrem-m gCi-year giCi-year

Based on reference meteorology applicable at 522 meters SE of stack.

- For exposure to a semi-infinite cloud of noble gas.

0929 1

CY-OC-170-301 Revision 4 Page 108 of 140 I Table A-2 Noble Gas Radionuclide-to-Dose Equivalent Rate Factors*

Radionuclide PySi** PVi*** AVi*** 3 SBi***

mrem mrem-m 3 mrad-m mrem-m 3 ltCi gCi-sec (Ki) gCi-sec (M) ltCi-sec (Li)

Kr83m 4.66E-17 2.40E-09 6.13E-07 Kr85m 2.91 E-12 3.71E-05 3.90E-05 4.63E-05 Kr85 4.66E-14 5.11E-07 5.46E-07 4.25E-05 Kr87 1.52E-11 1.88E-04 1.96E-04 3.09E-04 Kr88 3.73E-1 I 4.67E-04 4.83E-04 7.52E-05 Kr89 3.70E-11 5.27E-04 5.49E-04 3.21 E-04 Kr90 4.95E-04 5.17E-04 2.31E-04 Xel31 m 6.65E-13 2.90E-06 4.95E-06 1.51 E-05 Xe133m 5.20E-13 7.97E-06 1.04E-05 3.16E-05 Xe133 4.97E-13 9.33E-06 1.12E-05 9.71E-06 Xe135m 8.78E-12 9.90E-05 1.07E-04 2.26E-05 Xe135 4.78E-12 5.75E-05 6.10E-05 5.90E-05 Xe137 3.36E-12 4.51E-05 4.79E-05 3.87E-04 Xe138 2.42E- 11 2.80E-04 2.92E-04 1.31 E-04 Xe139 4.56E-12 - -

Ar4l 2.89E-11 2.81E-04 2.95E-04 8.54E-05

All of these dose factors apply out-of-doors.

- Based on reference meteorology at 522 meters SE of effluent stack.

- - Derived from Reg Guide 1.109, Revision 1, Table B-1.

0929 1

CY-OC-170-301 Revision 4 Page 109 of 140 Table A-3 Air Dose Conversion Factors for Effluent Noble Gas Radionuclide AySi** AT\i*** A;i***

mrad mrad-m3 mrad-m3 pCi 4Ci-sec(Mi) gCi-sec (N1)

Kr83m 9.35E-15 6.13E-07 9.14E-06 Kr85m 3.03E-12 3.90E-05 6.25E-05 Kr85 4.94E-14 5.46E-07 6.19E-05 Kr87 1.60E-1 1 1.96E-04 3.27E-04 Kr88 3.93E-1 1 4.83E-04 9.30E-05 Kr89 3.90E-1 1 5.49E-04 3.37E-04 Kr90 - 5.17E-04 2.49E-04 Xe131m 7.26E-13 4.95E-06 3.52E-05 Xel33m 5.86E-1 3 1.04E-05 4.70E-05 Xe133 5.45E-13 1.12E-05 3.33E-05 Xel35m 9.32E-12 1.07E-04 2.35E-05 Xe135 6.18E-12 6.10E-05 7.81E-05 Xe137 3.55E-12 4.79E-05 4.03E-04 Xe138 2.54E-11 2.92E-04 1.51 E-04 Xel 39 4.82E-12 -

Ar4l 3.03E-11 2.95E-04 1.04E-04

- Based on reference meteorology at 522 meters SE of effluent stack.

Derived from Reg Guide 1.109, Revision 1, Table B-1.

0929 I

CY-OC-170-301 Revision 4 Page 110 of 140 Table A-4 Locations Associated with Maximum Exposure of a Member of the Public*

Effluent Location Distance Direction (meters) (to)

Liquid U.S. Route 9 Bridge at Discharge Canal Airborne Iodine and Particulates 966 SE Tritium 966 SE Noble Gases 966 SE Irradiation by OCGS Site Boundary All Noble Gas g Air Dose Site Boundary All Noble Gas B Air Dose 966 SE Note: the nearby resident experiencing the maximum exposure to airborne effluent and to gamma radiation directly from the Station is located 966 meters SE of the OCGS. The most exposed member of the public is assumed to be exposed by irradiation from the OCGS, by inhaling airborne effluent, by irradiation by the airborne effluent, by irradiation by the airborne plume of the noble gas, by radionuclides deposited onto the ground, by irradiation by shoreline deposits, and by eating fish and shellfish caught in the discharge canal.

The age group of the most exposed member of the public is based on Reg. Guide 1.109, Revision 1.

0929 I

CY-OC-170-301 Revision 4 Page 111 of 140 Table A-5 Critical Receptor Noble Gas Dose Conversion Factors*

Radionuclide PYSi** PyVi*** AYVi*** AySi** SBi***

mrem mrem-m 3 mrad-m 3 mrad mrem-m 3 ACi p*Ci-sec(K*) gCi-sec(Mi) ACi ACi-sec(LI)

Kr83m 3.76E-17 2.40E-09 6.13E-07 9.66E-15 Kr85m 1.68E-12 3.71E-05 3.90E-05 1.75E-12 4.63E-05 Kr85 2.60E-14 5.11E-07 5.46E-07 2.75E-14 4.25E-05 Kr87 8.37E-12 1.88E-04 1.96E-04 8.81E-12 3.09E-04 Kr88 2.08E-11 4.67E-04 4.83E-04 2.18E-11 7.52E-05 Kr89 1.83E-11 5.27E-04 5.49E-04 1.93E-11 3.21E-04 Kr90 - 4.95E-04 5.17E-04 - 2.31E-04 Xe131m 3.99E-13 2.90E-06 4.95E-06 4.44 E-13 1.51 E-05 Xe133m 3.10E-13 7.97E-06 1.04E-05 3.58E-13 3.16E-05 Xe133 3.11E-13 9.33E-06 1.12E-05 3.42E-13 9.71E-06 Xe135m 4.71E-12 9.90E-05 1.07E-04 5.01E-12 2.26E-05 Xe135 2.73E-12 5.75E-05 6.1OE-05 2.87E-12 5.90E-05 Xe137 1.65E-12 4.51E-05 4.79E-05 1.75E-12 3.87E-04 Xe138 1.33E-11 2.80E-04 2.92E-04 1.40E-11 1.31 E-04 Xe139 - - 1.61E-12 Ar4l 1.58E-11 2.81E-04 2.95E-04 1.66E-11 8.54E-05

All of these dose factors apply out-of-doors.

- Based on reference meteorology at 522 meters SE of effluent stack.

Derived from Reg Guide 1.109, Revision 1, Table B-1 0929 1

CY-OC-170-301 Revision 4 Page 112 of 140 APPENDIX B - MODELING PARAMETERS 0929 1

CY-OC-170-301 Revision 4 Page 113 of 140 Table B-i- OCGS Usage Factors For Individual Dose Assessment Effluent Ingestion Parameters Usage Factor Fraction Of Produce From Local Garden 7.6E-1 Soil Density In Plow Layer (Kg/m2) 2.4E+2 Fraction Of Deposited Activity Retained On Vegetation 2.5E-1 Shielding Factor For Residential Structures 7.OE-1 Period Of Buildup Of Activity In Soil (hr) 1.31 E+5 Period of Pasture Grass Exposure to Activity (hr) 7.2E+2 Period Of Crop Exposure to Activity (hr) 1.44E+3 Delay Time For Ingestion Of Stored Feed By Animals (hr) 2.16E+3 Delay Time For Ingestion Of Leafy Vegetables By Man (hr) 2.4E+1 Delay Time For Ingestion Of Other Vegetables By Man (hr) 1.44E+3 Transport Time Milk-Man (hr) 4.8E+1 Time Between Slaughter and Consumption of Meat Animal (hr) 4.8E+2 Grass Yield Wet Weight (Kg/m 2 ) 7.OE-1 Other Vegetation Yield Wet-Weight (Kg/m2) 2.0 Weathering Rate Constant For Activity on Veg. (hr- 1 ) 2.1E-3 Milk Cow Feed Consumption Rate (Kg/day) 5.0E+1 Goat Feed Consumption Rate (Kg/day) 6.0 Beef Cattle Feed Consumption Rate (Kg/day) 5.0E+1 Milk Cow Water Consumption Rate (L/day) 6.0E+1 Goat Water Consumption Rate (L/day) 8.0 Beef Cattle Water Consumption Rate (L/day) 5.0E+1 Environmental Transit Time For Water Ingestion (hr) 1.2E+1 Environmental Transit Time For Fish Ingestion (hr) 2.4E+1 Environmental Transit Time For Shore Exposure (hr) 0 Environmental Transit Time For Invertebrate Ingestion (hr) 2.4E+1 0929 I

CY-OC-170-301 Revision 4 Page 114 of 140 Table B-1 (Continued)

OCGS Usage Factors For Individual Dose Assessment Effluent Ingestion Parameters Usage Factor Water Ingestion (L/yr)

a. Adult 7.3E+2

b. Teen 5.1E+2

c. Child 5.1E+2

d. Infant 3.3E+2 Shore Exposure (hr/yr)

a. Adult 1.2E+1

b. Teen 6.7E+1

c. Child 1.4E+1

d. Infant 0 Salt Water Sport Fish Ingestion (Kg/yr)

a. Adult 2.1E+1

b. Teen 1.6E+1

c. Child 6.9

d. Infant 0 Salt Water Commercial Fish Ingestion (Kg/yr)

a. Adult 2.1E+1

b. Teen 1.6E+1

c. Child 6.9

d. Infant 0 Salt Water Invertebrate Ingestion (Kg/yr)

a. Adult 5.0

b. Teen 3.8

c. Child 1.7

d. Infant 0 Irrigated Leafy Vegetable Ingestion (Kg/yr)

a. Adult 6.4E+1

b. Teen 4.2E+1

c. Child 2.6E+1

d. Infant 0 0929 I

CY-OC-170-301 Revision 4 Page 115 of 140 Table B-1 (Continued)

OCGS Usage Factors For Individual Dose Assessment Effluent Ingestion Parameters Usaqe Factor Irrigated Other Vegetable Ingestion (Kg/yr)

a. Adult 5.2E+2

b. Teen 6.3E+2

c. Child 5.2E+2

d. Infant 0 Irrigated Root Vegetable Ingestion (Kglyr)

a. Adult 5.2E+2

b. Teen 6.3E+2

c. Child 5.2E+2

d. Infant 0 Irrigated Cow and Goat Milk Ingestion (L/yr)

a. Adult 3.1E+2

b. Teen 4.OE+2

c. Child 3.3E+2

d. Infant 3.3E+2 Irrigated Beef Ingestion (Kg/yr)

a. Adult 1.1E+2

b. Teen 6.5E+1

c. Child 4.1E+1

d. Infant 0 Inhalation (m 3/yr)

a. Adult 8.OE+3

b. Teen 8.OE+3

c. Child 3.7E+3

d. Infant 1.4E+3 Cow and Goat Milk Ingestion (LUyr)

a. Adult 3.1E+2

b. Teen 4.OE+2

c. Child 3.3E+2

d. Infant 3.3E+2 Meat Ingestion (Kg/yr)

a. Adult 1.1E+2

b. Teen 6.5E+1

c. Child 4.1E+1

d. Infant 0 0929 1

CY-OC-170-301 Revision 4 Page 116 of 140 Table B-1 (Continued)

OCGS Usage Factors For Individual Dose Assessment Effluent Ingestion Parameters Usage Factor Leafy Vegetable Ingestion (Kg/yr)

a. Adult 6.4E+1

b. Teen 4.2E+1

c. Child 2.6E+1

d. Infant 0 Fruits, Grains, & Other Vegetable Ingestion (Kg/yr)

a. Adult 5.2E+2

b. Teen 6.3E+2

c. Child 5.2E+2

d. Infant 0 0929 I

CY-OC-170-301 Revision 4 Page 117 of 140 Table B-2 Monthly Average Absolute Humidity g/m 3 (derived from historical climatological data)

Average Absolute Month Humidity (./m 3)

January 3.3 February 3.3 March 4.5 April 6.1 May 9.4 June 12.8 July 15.2 August 15.6 September 12.4 October 7.9 November 5.9 December 3.8 0929 1

CY-OC-170-301 Revision 4 Page 118 of 140 APPENDIX C - REFERENCES 0929 I

CY-OC-170-301 Revision 4 Page 119 of 140 Table C REFERENCES

1) Oyster Creek Updated Final Safety Analysis Report

2) Oyster Creek Facility Description and Safety Analysis Report

3) Oyster Creek Operating License and Technical Specifications

4) NUREG 1302 "Off Site Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Boiling Water Reactors" - Generic Letter 89-10, Supplement No.

1,April 1991

5) Reg Guide 1.21 "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of radioactive materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants" Rev.1, June 1974

6) Reg Guide 1.23

7) Reg Guide 1.97

8) Reg Guide 1.109 "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR 50, Appendix I",Rev 1, October, 1977

9) Reg Guide 1.111 "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water-Cooled Reactors", Rev.1, July, 1977

10) Reg Guide 4.8" Environmental Technical Specifications for Nuclear Power Plants"

11) NRC Radiological Assessment Branch Technical Position, Rev 1, November 1979 (Appendix A to NUREG1302)

12) NUREG-0016

13) NUREG-0133

14) Licensing Application, Amendment 13, Meteorological Radiological Evaluation for the Oyster Creek Nuclear Power Station Site.

15) Licensing Application, Amendment 11, Question IV-8.

16) Evaluation of the Oyster Creek Nuclear Generating Station to Demonstrate Conformance to the Design Objectives of 10CFR50, Appendix I, May, 1976, Tables 3-10 0929

CY-OC-170-301 Revision 4 Page 120 of 140

17) XOQDOQ Output Files for Oyster Creek Meteorology, Murray and Trettle, Inc.

18) Hydrological Information and Liquid Dilution Factors Determination to Conform with Appendix I Requirements: Oyster Creek, correspondence from T. Potter, Pickard, Lowe and Garrick, Inc. to Oyster Creek, July, 1976.

19) Carpenter, J. J. "Recirculation and Effluent Distribution for Oyster Creek Site", Pritchard-Carpenter Consultants, Baltimore, Maryland, 1964.

20) Nuclear Regulatory Commission, Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section and Relocation of the Procedural Details of RETS to the ODCM or PCP", January, 1989.

21) GroundWater Monitoring System (Final Report), Woodward-Clyde Consultants, March, 1984.

22) Meteorology and Atomic Energy, Department of Energy, 1981.

23) SEEDS Code Documentation through V & V of Version 98.8F (Radiological Engineering Calculation No. 2820-99-005, Dated 3/23/99)

24) Lynch, Giuliano, and Associates, Inc., Drawing Entitled, "Minor Subdivision, Lots 4 and 4.01 Block 1001", signed 13 Sep 99.

25) Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements".

26) NUREG/CR-4007 (September 1984).

27) HASL Procedures Manual, HASL-300 (revised annually).

28) Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purposes of Implementing Appendix I," April 1977

29) Reg. Guide 4.13

30) 10CFR20, Appendix B, Table 2, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage 0929 1

CY-OC-170-301 Revision 4 Page 121 of 140

31) Conestoga Rovers and Associates, Hydrogeologic Investigation Report, Fleet wide Assessment, Oyster Creek Generating Station, Forked River, New Jersey, Ref. No.

045136(18), September, 2006.

0929 1

CY-OC-170-301 Revision 4 Page 122 of 140 APPENDIX D - SYSTEM DRAWINGS 0929 I

CY-OC-170-301 Revision 4 Page 123 of 140 FIGURE D-1-la: LIQUID RADWASTE TREATMENT CHEM WASTE AND FLOOR DRAIN SYSTEM o

00 "rI',

.* V

.2 cý 0

VL)

V--"*g

- *S cl) 0 o 09291

CY-OC-1 70-301 Revision 4 Page 124 of 140 FIGURE D-1-lb: LIQUID RADWASTE TREATMENT - HIGH PURITY AND EQUIPMENT DRAIN SYSTEM I-w (n 0 zJ a) oo Z0

, 0

<* XE *m T-co8 W*

- - o U)

Cc*

0929 I

CY-OC-1 70-301 Revision 4 Page 125 of 140 FIGURE D=1-2: SOLID RADWASTE PROCESSING SYSTEM FIGURE D- 1-2: SOLID RADWASTE PROCESSING SYSTEM Filter Sludge 0929

CY-OC-1 70-301 Revision 4 Page 126 of 140 FIGURE D-2-1: GASEOUS RADWASTE TREATMENT - AUGMENTED OFF GAS SYSTEM 0929 1

CY-OC-170-301 Revision 4 Page 127 of 140 FIGURE D-2-2: VENTILATION SYSTEM 0929 1

CY-OC-1 70-301 Revision 4 Page 128 of 140 Figure D-2-3 AOG Ventilation System HV-S-1O 0929 I

CY-OC-170-301 Revision 4 Page 129 of 140 APPENDIX E - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM - SAMPLE TYPE AND LOCATION All sampling locations and specific information about the individual locations are given in Table E-1. Figures E-1, E-2 and E-3 show the locations of sampling stations with respect to the site. Figure E-4 shows the site layout.

0929 I

CY-OC-170-301 Revision 4 Page 130 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (1)

Sample Station Distance Azimuth Medium Code (miles) (degreesl Description TLD 1 0.4 219 SW of site at OCGS Fire Pond, Forked River, NJ (Inner Ring)

DW 1 0.1 209 On-site southern domestic well at OCGS, Forked River, NJ 0.2 349 On-site northern domestic well at OCGS, Forked River, NJ Either the southern or northern well is sampled.

APT, AIO, TLD 3 6.0 97 East of site, near old Coast Guard Station, Island Beach State Park TLD - Special Interest Area TLD 4 4.6 213 SSW of Site, Garden State Parkway and Route 554, Barnegat, NJ (Outer Ring)

TLD 5 4.2 353 North of Site, Garden State Parkway Rest Area, Forked River, NJ (Outer Ring)

TLD 6 2.1 13 NNE of site, Lane Place, behind St. Pius Church, Forked River, NJ (Outer Ring)

TLD 8 2.3 177 South of site, Route 9 at the Waretown Substation, Waretown, NJ (Outer Ring)

TLD 9 2.0 230 SW of site, where Route 532 and the Garden State Parkway meet, Waretown, NJ (Outer Ring)

APT, AIO, TLD C 24.7 313 NW of site, JCP&L office in rear parking lot, Cookstown, NJ (Background Station)

TLD 11 8.2 152 SSE of site, 8 0 th and Anchor Streets, Harvey Cedars, NJ (Special Interest Area)

TLD 14 20.8 2 North of site, Larrabee Substation on Randolph Road, Lakewood, NJ (Background Station)

APT, AIO 20 0.7 95 East of site, on Finninger Farm on south side of access road, Forked River, NJ TLD 22 1.6 145 SE of site, on Long John Silver Way, Skippers Cove, Waretown, NJ (Outer Ring)

SWA, CLAM, 23 3.6 64 ENE of site, Barnegat Bay off Stouts Creek, 0929 1

CY-OC-1 70-301 Revision 4 Page 131 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (1)

Sample Station Distance Azimuth Medium Code (miles) (degrees) Description AQS approximately 400 yards SE of "Flashing Light 1" SWA, CLAM, 24 2.1 101 East of site, Barnegat Bay, approximately AQS 250 yards SE of "Flashing Light 3" SWA, AQS, 33 0.4 123 ESE of site, east of Route 9 Bridge in FISH OCGS Discharge Canal VEG 35 0.4 111 ESE of site, east of Route 9 and north of the OCGS Discharge Canal, Forked River, NJ VEG 36 23.1 319 NW of site, at "U-Pick" Farm, New Egypt, NJ (Background Station)

DW 37 2.2 18 NNE of Site, off Boox Road at Lacey MUA Pumping Station, Forked River, NJ (Background Station)

DW 38 1.6 197 SSW of Site, on Route 532, at Ocean Township MUA Pumping Station, Waretown, NJ DW 39 3.5 353 N of Site, Trenton Ave. off Lacey Road Lacey Twp., MUA Pump Station, Forked River, NJ (Background Station)

TLD 46 5.6 323 NW of Site, on Lacey Road adjacent to Utility Pole BT 259 65 (Outer Ring)

TLD 47 4.6 26 NNE of Site, Route 9 and Harbor Inn Road, Berkeley Township, NJ (Outer Ring)

TLD 48 4.5 189 South of Site, Intersection of Brook and School Streets, Barnegat, NJ (Outer Ring)

TLD 51 0.4 358 North of site, on the access road to Forked River Site, Forked River, NJ (Inner Ring)

TLD 52 0.3 333 NNW of site, on the access road to Forked River Site, Forked River, NJ (Inner Ring)

TLD 53 0.3 309 NW of site, at sewage lift station on the access road to the Forked River Site, 0929 I

CY-OC-170-301 Revision 4 Page 132 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (1)

Sample Station Distance Azimuth Medium Code (miles) (degrees) Description Forked River, NJ (Inner Ring)

TLD 54 0.3 288 WNW of site, on the access road to Forked River Site, Forked River, NJ (Inner Ring)

TLD 55 0.3 263 West of site, on Southern Area Stores security fence, west of OCGS Switchyard, Forked River, NJ (Inner Ring)

TLD 56 0.3 249 WSW of site, on utility pole east of Southern Area Stores, west of the OCGS Switchyard, Forked River, NJ (Inner Ring)

TLD 57 0.2 206 SSW of site, on Southern Area Stores access road, Forked River, NJ (TLD - ODCM Required - Inner Ring)

TLD 58 0.2 188 South of site, on Southern Area Stores access road, Forked River, NJ (Inner Ring)

TLD 59 0.3 166 SSE of site, on Southern Area Stores access road, Waretown, NJ (Inner Ring)

TLD 61 0.3 104 ESE of site, on Route 9 south of OCGS Main Entrance, Forked River, NJ (Inner Ring)

TLD 62 0.2 83 East of site, on Route 9 at access road to OCGS Main Gate, Forked River, NJ (Inner Ring)

TLD 63 0.2 70 ENE of site, on Route 9, between main gate and OCGS North Gate access road, Forked River, NJ (Inner Ring)

TLD 64 0.3 42 NE of site, on Route 9 North at entrance to Finninger Farm, Forked River, NJ (Inner Ring) 0929 1

CY-OC-1 70-301 Revision 4 Page 133 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (Continued)

Sample Station Distance Azimuth Medium Code (miles) (degrees) Description TLD 65 0.4 19 NNE of site, on Route 9 at Intake Canal Bridge, Forked River, NJ (Inner Ring)

APT, AIO, 66 0.4 133 SE of site, east of Route 9 and south of TLD, the OCGS Discharge Canal, inside VEG fence, Waretown, NJ (TLD - Inner Ring)

TLD 68 1.3 266 West of site, on Garden State Parkway North at mile marker 71.7, Lacey Township, NJ (Outer Ring)

APT, AIO, TLD 71 1.6 164 SSE of site, on Route 532 at the Waretown Municipal Building, Waretown, NJ (TLD - Special Interest Area)

APT, AIO, TLD 72 1.9 25 NNE of site, on Lacey Road at Knights of Columbus Hall, Forked River, NJ (TLD - Special Interest Area)

APT, AIO, TLD 73 1.8 108 ESE of site, on Bay Parkway, Sands Point Harbor, Waretown, NJ (TLD - Outer Ring)

TLD 74 1.8 88 East of site, Orlando Drive and Penguin Court, Forked River, NJ (Outer Ring)

TLD 75 2.0 71 ENE of site, Beach Blvd. and Maui Drive, Forked River, NJ (Outer Ring)

TLD 78 1.8 2 North of site, 1514 Arient Road, Forked River, NJ (Outer Ring)

TLD 79 2.9 160 SSE of site, Hightide Drive and Bonita Drive, Waretown, NJ (Outer Ring)

TLD 81 3.5 201 SSW of site, on Rose Hill Road at intersection with Barnegat Boulevard, Barnegat, NJ( Special Interest Area) 0929 I

CY-OC-170-301 Revision 4 Page 134 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (Continued)

Sample Station Distance Azimuth Medium Code (miles) (degrees) Description TLD 82 4.4 36 NE of site, Bay Way and Clairmore Avenue, Lanoka Harbor, NJ (Outer Ring)

TLD 84 4.4 332 NNW of site, on Lacey Road, 1.3 miles west of the Garden State Parkway on siren pole, Lacey Township, NJ (Outer Ring)

TLD 85 3.9 250 WSW of site, on Route 532, just east of Wells Mills Park, Waretown, NJ (Outer Ring)

TLD 86 5.0 224 SW of site, on Route 554, 1 mile west of the Garden State Parkway, Barnegat, NJ (Outer Ring)

TLD 88 6.6 125 SE of site, eastern end of 3 rd Street, Barnegat Light, NJ (Special Interest Area)

TLD 89 6.1 108 ESE of site, Job Francis residence, Island Beach State Park (Special Interest Area)

TLD 90 6.3 75 ENE of site, parking lot A-5, Island Beach State Park (Special Interest Area)

TLD 92 9.0 46 NE of site, at Guard Shack/Toll Booth, Island Beach State Park (Special Interest Area)

FISH, CRAB 93 0.1 242 WSW of site, OCGS Discharge Canal between Pump Discharges and Route 9, Forked River, NJ SWA, AQS, 94 20.0 198 SSW of site, in Great Bay/Little Egg CLAM, Harbor FISH (Background Station)

TLD 98 1.6 318 NW of site, on Garden State Parkway at mile marker 73.0, Lacey Township, NJ (Outer Ring) 0929 1

CY-OC-170-301 Revision 4 Page 135 of 140 TABLE E-1: REMP SAMPLE LOCATIONS (Continued)

Sample Station Distance Azimuth Medium Code (miles) (dea rees) Description TLD 99 1.5 310 NW of site, on Garden State Parkway at mile marker 72.8, Lacey Township, NJ (Outer Ring)

TLD 100 1.4 43 NE of site, Yacht Basin Plaza South off Lakdeside Dr., Lacey Township, NJ (Outer Ring)

TLD 101 1.7 49 NE of site, end of Lacey Rd., East, Lacey Township, NJ (Outer Ring)

TLD 102 1.6 344 NNW of site, end of Sheffield Dr.,

Barnegat Pines, Lacey Township, NJ (Outer Ring)

TLD 103 2.4 337 NNW of site, Llewellyn Parkway, Barnegat Pines, Lacey Township, NJ (Outer Ring)

TLD 104 1.8 221 SW of site, Rt. 532 West, before Garden State Parkway, Ocean Township, NJ (Outer Ring)

TLD 105 2.8 222 SW of site, Garden State Parkway North, beside mile marker 69.6, Ocean Township, NJ (Outer Ring)

TLD 106 1.2 288 NW of site, Garden State Parkway North, beside mile marker 72.2 Lacey Township, NJ (Outer Ring)

TLD 107 1.3 301 NW of Site, Garden State Parkway North, beside mile marker 72.5, Lacey Township, NJ (Outer Ring)

TLD 109 1.2 141 SE of site, Lighthouse Dr., Waretown, Ocean Township, NJ (Outer Ring)

TLD 110 1.5 127 SE of site, Tiller Drive and Admiral Way, Waretown, Ocean Township, NJ (Outer Ring)

APT, AIO 111 0.3 64 ENE of site, Finninger Farm property along access road, Lacey Township, NJ TLD 112 0.2 178 S of site, along Southern access road, Lacey Township, NJ (Inner Ring)

TLD 113 0.3 90 E of site, along Rt. 9 North, Lacey Township, NJ (Inner Ring)

TLD T1 0.4 219 SW of site, at OCGS Fire Pond, Lacey Township, NJ (Inner Ring)

GW W-3C 0.4 112 ESE of site on Finninger Farm adjacent to Station 35, Lacey Township, NJ 0929

CY-OC-170-301 Revision 4 Page 136 of 140 GW MW- 0.8 97 E of site on Finninger Farm on South 24-3A side of access road, Lacey Township, NJ SAMPLE MEDIUM IDENTIFICATION KEY APT = Air Particulate SWA = Surface Water TLD = Thermoluminescent Dosimeter A1O = Air Iodine AQS = Aquatic Sediment FISH = Fish CLAM = Clams CRAB =Crab VEG = Vegetables DW = Drinking Water GW = Ground Water (1) Samples may not be collected from some locations listed in this table, as long as the minimum number of samples listed in Table 3.12.1-1 is collected.

0929.1

CY-OC-1 70-301 Revision 4 Page 137 of 140 FIGURE E-1 Figure E-1.

Locations of REMP Stations within a 1-mile radius of the Oyster Creek Generating Station 0929 I

CY-OC-1 70-301 Revision 4 Page 138 of 140 FIGURE E-2

.I- , '--.-. - *f.

7

."' 4.:,,- : o3 , .

i'*_

" "74' a 82 LoationsatREMPStaionswithina to.- 5 -mile'r

.~ ~ n 7.9*-"_:.:

NW ..."SW -. sI SE41/2"'Ž'

..- -, ,; . r" ENEx' 7 jai..

of .w the. OytrCee.e

.. '54L Figur E-2 Location. ,of .,.,M,Sttin - erdu ,~hnaIt of th OseCreGeeaigStio 0929 I

CY-OC-170-301 Revision 4 Page 139 of 140 HIGURE E-3 Figure E-3.

Locations of REMP Stations greater than 5 miles from the Oyster Creek Generating Station 0929 1

CY-OC-170-301 Revision 4 Page 140 of 140 FMGURE E-4 AREA PLOT PLAN OF SITE SITE MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 0929 I

Procedure Approval Form AD-AA-101-F-01 Page 1 of 1 Revision 0 Document Number: RW-AA-100 Revision: 6

Title:

Process Control Program for Radioactive Wastes Q New [I Cancel 0 Cancel 0 Revision EC#: PCRN: _ PPIS#:

Document Revision El Editorial 0 Batch ER#: AR#: __#:

[C Supersede corporate document(s) List:

Revision Summary: Changed Step 4.2.10 for Peach Bottom and Limerick due to DHEC approval for Attach add'i descript, itreq'd several HIC designs resulting from the acquisition of Nukem and since Certificate of Compliance are not require for disposal at the Clive disposal site.

Impact on Operating [ N/A and Design Margins:

Attach add'] descript, it req'd CONFIRM that no commitments (i.e.. those steps annotated with CM-X) have been changed or deleteo, upless evaluated via completion of LS-AA-1 10 commitment change/deletion form and INITIAL [Preparer]:

Preparer: Miguel Azar/ 03/24/08 Cantera/3204 Print Date Location/Ext Applicable BR 0 Marcia Morris DR 0 Sandy Livecchi 0 QC Terry Barber Site Contacts BY 0 Norma Jean Gordon LA 0 Lynn Kofoid-Durdan CL 0 Lindsay Green Check box and PB 0 George Tharpe OC 0 Michael Steloff LG 0 Linda Knapp provide name TMI0 Jessica Spagnuolo ZN [] Other 0[

SA C3 HC [I Other 0_

Validation Req'd: 0 No [] Yes (attach) Common Training Req'd: 0 No [ Yes (Validation requirement see AD-AA-101) /-*' Print/Signature Site Specific Training Req'd: 0 No C] Yes Change Management: 17 HU-AA- 1101 Changd'Checklist Attached [] Document Traveler N None Required Level of Use: El Level 1 - Continq_, UseU , Lev 12 - Reference Use 0 Level 3 - Information Use Approval Miguel Azar . 03/24/08 Cantera/3200 CFAM (Standard Procadu,,e)/ Print/Sign Date Location/Ex:

Approval 0 c ,. Site Document(s) to be superseded:

Location: I Use additional sheets as necessary. Assure that all pending changes are dispositioned.

[I Temp. Change El Interim Change Temp or Interim Change #: Interim Change expiration:

10CFR50.59 Applicable: E] No 1 Yes 0 Excluded (Or applicable regulatory process reviews) t - V0 *- S " Of?t

0) per P' (A 10CFR72.48 Applicable: [$ No El Yes Tracking Number PORC Required: E] No g Yes PORC Number (after PORC Approved): 0 0 It speredig cotaiingcomitmnts noifythe Comrmitment Tracking Coordinator per LS-AA-110 so the CTD can be updated as appropriate.

adocmen Cross Discipline Reviews: (list below) illance Coordinator Review Req'd8 No C3 Yes, list below Print Signature Date Ciscipline or Org."

Print Signature Date Oisopline or Org.

Print Signature Date Discileine or Org.

Attach acdtdional it rea'd Temp Change Uri Authorization u Only SRO Print/Sign/Date SOR Print/Sign/Date Impl. Date Exp. Date SOR Approval indicates that all required Cross-Disciplinary reviews have been performed an the reviewers have signed this form. This procedure is technically and functionally accurate tot all functional areas.

Site Authorization: \Ll*rlaI**l. AAr1`1 M16Oh* -

LS-AA-106 Revision 3 Page 20 of 22.

ATTACHMENT 3 Nuclear Safety Significance Assessment Form Page 1 of I Item

Title:

/7C ESS c6&7* 0T4OL_ G4'4A* Foe /*?4P1e4WT/V v 4 SSTE-S Item Number: /I?, - ,4, -/00 Revision No. 6 Does the proposed change: YES NO 1 Result in a change to a procedure affecting Technical Specifications, ECCS, X ESF, or PRA risk significant equipment or systems?

2 Result in a modification or change to a ECCS, ESF, or PRA risk significant X system?

3 Consist of a major change to the facility and/or a major test or experiment? X 4 Consist of a major change to a plant process? X 5 Change the qualification or operational characteristics of installed components X or systems classified as safety related.

6 Change the nuclear safety response of the plant to normal evolutions, X anticipated operational occurrences, or design basis accidents?

7 Have the potential to reduce the ability of the operator to assess or control the X nuclear safety status of the plant?

8 Result from investigations of significant operational abnormalities including X accidental unplanned or uncontrolled radioactive releases?

9 Increase the potential for a plant trip or present a challenge to safety systems? X 10 Require NRC approval prior to implementation, e.g., TS, Security Plan, X Emergency Plan?

11 10 CFR 50.59/10 CFR 72.48 written evaluation be prepared? X If any answer to the above questions is "Yes", a full PORC review is required. If all questions are answered "No", a full PORC review is not required. Assumptions must be documented in the Comments Section below.

COMMENTS: There are no safety related aspects of this procedure revision Circle as applicable THIS ITEM DOES91 EQUIRE FULL PORC REVIEW.

PORC Members Name ,?/c,4e:' ,VfeoS Department: A4S'4,I/ci 4&*?*'

PORC Member's Date: 6 -e -O' Signature Department: Q A._-..t ,0 Date: 4'- Z-/c PORC Member's Name Department:

Print PORC Member's Date:

Signature

LS-AA-106 Revision 3 Page 20 of 22 ATTACHMENT 3 Nuclear Safety Significance Assessment Form Page 1 of I Item

Title:

PIFaeES5 24,vTAo- '0LP?3C4,.4A F-oe 1p4cy r71 V LVA4STIFs Item Number: /?'L4/-4-4 -/00 Revision No. 6 Does the proposed change: YES NO 1 Result in a change to a procedure affecting Technical Specifications, ECCS, x ESF, or PRA risk significant equipment or systems?

2 Result in a modification or change to a ECCS, ESF, or PRA risk significant X system?

3 Consist of a major change to the facility and/or a major test or experiment? x 4 Consist of a major change to a plant process? x 5 Change the qualification or operational characteristics of installed components X or systems classified as safety related.

6 Change the nuclear safety response of the plant to normal evolutions, X anticipated operational occurrences, or design basis accidents?

7 Have the potential to reduce the ability of the operator to assess or control the X nuclear safety status of the plant?

8 Result from investigations of significant operational abnormalities including X accidental unplanned or uncontrolled radioactive releases?

9 Increase the potential for a plant trip or present a challenge to safety systems? x 10 Require NRC approval prior to implementation, e.g., TS, Security Plan, x Emergency Plan?

11 10 CFR 50.59/10 CFR 72.48 written evaluation be prepared? - x If any answer to the above questions is "Yes", a full PORC review is required. If all questions are answered "No", a full PORC review is not required. Assumptions must be documented in the Comments Section below.

COMMENTS: There are no safety related aspects of this procedure revision Circle as applicable THIS ITEM DOESOýýEQUIRE FULL PORC REVIEW.

PORC Member's Name ,?iCi'i:>,.IoS Department: 1?'F6- ,45"'S4,eIAMA'(C PORC Members /I Date: 6 -"4 -e-Signature PORC Member's Name -1bOhl/ V. M'44I*/L. Department: Cg,'*"

PORC Member's Dae:ture Signature PORC Member's Name __Department:

Print PORC Member's Date:

Signature

PORC ITEM

SUMMARY

COVER SHEET

SUBJECT:

Revioinn 6 RW-AA-100 Prnoces Cnntrnl Prngro m fnr Radrinnrtive Wastes PREPARED BY: James Richwine IS A 50.59/72.48 SAFETY EVALUATION REQUIRED? [ Yes [X ] No ISSUE

SUMMARY

RW-AA-100 is being revised to change step 4.2.10 for Peach Bottom and Limerick due to DHEC approval for several HIC designs resulting from the acquisition of Nukem and since Certificates of Compliance are not required for disposal at the Clive disposal site.

This activity is to revise RW-AA-1 00 "Process Control Program for Radioactive Wastes" Step 4.2.10.1 requires vendors to supply a Certificate of Compliance for HIC disposal at the Barnwell disposal site.

Step 4.2.10.2 requires vendors to supply a Certificate of Conformance for HIC disposal at-The Clive disposal site.

Step 4.2.10.3 added the words Certificate of Conformance.

SAFETY IMPACT:

The major change in this procedure is to ensure there is a Certificate of Compliance for Disposal of HIC's at Barnwell and a certificate of Conformance for disposal of HIC's at Clive.

This activity does not have any safety impact on Plant Operations The Presentation Material is Ready for PORC Review

7 0(?

,resenter's Signature Date upervisor's S(6nature Date

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 2 Page I of I Station/Unit(s): Oyster Creek Activity/Document Number: Revision RW-AA- 100 Revision Number: 6

Title:

Process Control Program for Radioactive Wastes NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

This activity is to revise RW-AA-100 "Process Control Program for Radioactive Wastes" Step 4.2.10.1 requires vendors to supply a Certificate of Compliance for HIC disposal at the Barnwell disposal site.

Step 4.2.10.2 requires vendors to supply a Certificate of Conformance for HIC disposal at the Clive disposal site.

Step 4.2.10.3 added the words Certificate of Conformance.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

This activity is to ensure there is a certificate of compliance for disposal of HIC's at Barnwell, SC and a certificate of conformance for disposal of HIC's at Clive, UT.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

This activity is administrative and will not affect plant operations, design bases or safety analyses. The solid waste management system is described in the UFSAR section 11.4 and does not describe Certificates of Compliance or Certificates of Conformance for Oyster Creek Station.

Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The Screening demonstrates a 50.59 Evaluation is not required. The proposed changes are necessary to ensure a clear understanding of which certificate is required for the disposal sites of Barnwell, SC and Clive, UT.

The procedure may be implemented without prior NRC approval.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

(NOTE: if both a Screening and Evaluation are completed, no Screening No. is required.)

Forms Attached: (Check all that apply.)

H

Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.

50.59 Evaluation No.

OC-2008-S-0082 Rev.

Rev.

0

50.59 APPLICABILITY REVIEW FORM LS-AA- 104-1002 Revision 3 Page I of I Activity/Document Number: RW-AA-100 Revision Number: .6 Address the questions below for all aspects of the Activity. If the answer is yes for any portion of the Activity, apply the identified process(es) to that portion of the Activity. Note that it is not unusual to have more than one process apply to a given Activity.

See Section 4 of the Resource Manual (RM) for additional guidance.

I. Does the proposed Activity involve a change:

I. Technical Specifications or Facility Operating License (I0CFR50.90)? 0 NO El YES See Section 4.2.1.1 of the RM

2. Conditions of License Quality Assurance program (I OCFR50.54(a))? 0 NO El YES Security Plan (IOCFR50.54(p))? 0 NO El YES See Section 4.2.1.2 of the RM Emergency Plan (I 0CFR50.54(q))? 0 NO El YES

3. Codes and Standards IST Program Plan (IOCFR50.55a(f))? 0 NO El YES ISI Program Plan (IOCFR50.55a(g))? Z NO El YES

4. ECCS Acceptance Criteria (IOCFR50.46)? [ NO 0l YES See Section 4.2.1.4 of the RM

5. Specific Exemptions (IOCFR50.12)? 0 NO 0l YES See Section 4.2.1.5 of the RM

6. Radiation Protection Program (I OCFR20)? Z NO El YES See Section 4.2.1.6 of the RM

7. Fire Protection Program (applicable UFSAR or operating license S NO El YES See Section 4.2.1.7 of the RM condition)?

8. Programs controlled by the Operating License or the Technical ED NO Dl YES See Section 4.2.1.7 of the RM Specifications (such as the ODCM).

9. Environmental Protection Program 0 NO El YES See Section 4.2.1.7 of the RM

10. Other programs controlled by other regulations. E NO 0l YES See Section 4.2.1 of the RM II. Does the proposed Activity involve maintenance which restores SSCs to their original condition or involve a temporary alteration supporting E NO E] YES See Section 4.2.2 of the RM maintenance that will be in effect during at-power operations for 90 days or less?

III. Does the proposed Activity involve a change to the:

1. UFSAR (including documents incorporated by reference) that is excluded from the requirement to perform a 50.59 Review by NEI 96-07 Z NO E] YES See Section 4.2.3 of the RM or NEI 98-03?

2. Managerial or administrative procedures governing the conduct of E NO El YES See Section 4.2.4 of the RM facility operations (subject to the control of IOCFR50, Appendix B)

3. Procedures for performing maintenance activities (subject to I0CFR50, El NO El YES See Section 4.2.4 of the RM Appendix B)?

4. Regulatory commitment not covered by another regulation based change [ NO El YES See Section 4.2.3/4.2.4 of the RM process (see NEI 99-04)?

IV. Does the proposed Activity involve a change to the Independent Spent Fuel [ NO El YES See Section 4.2.6 of the RM

_ Storage Installation (ISFSI) (subject to control by 10 CFR 72.48) 1 Check one of the following:

E] If all aspects of the Activity are controlled by one or more of the above processes, then a 50.59 Screening is not required and the Activity may be implemented in accordance with its governing procedure.

[ If any portion of the Activity is not controlled by one or more of the above processes, then process a 50.59 Screening for the portion not covered by any of the above processes. The remaining portion of the activity should be implemented in accordance with its governing procedure.

Si gnoff:

ý50.59Screeiier 0.59 Evaluator: James Richwine Sign:

gn::/Date: S,/'7.Co/O (Circle One) (Print name) (Signature)

50.59 SCREENING FORM LS-AA- 104-1003 Revision I Page I of 2 50.59 Screening No. OC-2008-S-0082 Rev. No. 0-Activity/Document Number: Revision RW-AA-I(N0 Revision Number:6 I. 50.59 Screening Questions (Check correct response and provide separate written response providing the basis for the answer to each question)(See Section 5 of the Resource Manual (RM) for additional guidance):

I. Does the proposed Activity involve a change to an SSC that adversely affects an UFSAR YES x NO described design function? (See Section 5.2.2.1 of the RM)

2. Does the proposed Activity involve a change to a procedure that adversely affects how UFSAR YES x NO described SSC design functions are performed or controlled? (See Section 5.2.2.2 of the RM)

3. Does the proposed Activity involve an adverse change to an element of a UFSAR described YES x NO evaluation methodology, or use of an alternative evaluation methodology, that is used in establishing the design bases or used in the safety analyses? (See Section 5.2.2.3 of the RM)

4. Does the proposed Activity involve a test or experiment not described in the UFSAR, where an YES x NO SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or is inconsistent with analyses or descriptions in the UFSAR? (See Section 5.2.2.4 of the RM)

5. Does the proposed Activity require a change in the Technical Specifications or Operating YES x NO License? (See Section 5.2.2.5 of the RM)

!1. List the documents (e.g., UFSAR. Technical Specifications, other licensing basis, technical, commitments, etc.) reviewed, including sections numbers where relevant information was found (if not identified in the response to each question).

See References section on page 2 I11. Select the appropriate conditions:

If all questions are answered NO, then complete the 50.59 Screening and implement the Activity per the applicable X governing procedure.

If question 1, 2, 3, or 4 is answered YES and question 5 is answered NO, then a 50.59 Evaluation shall be performed.

If questions I, 2, 3. and 4 are answered NO and question 5 is answered YES, then a License Amendment is required prior to implementation of the Activity.

If question 5 is answered YES for any portion of an Activity, then a License Amendment is required prior to implementation of that portion of the Activity. In addition, if question I, 2, 3, or 4 is answered YES for the remaining portions of the Activity, then a 50.59 Evaluation shall be performed for the remaining portions of the Activity.

IV. Screening Signoffs:

50.59 Screener: '4 e ,i t'ci,* Sign: 1 ,--I Date: _____lo_

(Print name) / - (Signature) 50.59 Reviewer: Sign: Date* V 3, 4 ,,/ 0a (Print name)

50.59 SCREENING FORM LS-AA-1104-1003 Revision I Page 2 of 2 50.59 Screening No. OC-2008-S-0082 Rev. No. 0(

Activity/I)ocument Number: Revision RW-AA-100 Revision Number:6 I. The proposed activity implements revision 6 of the Process Control Program for Radioactive Wastes (RW-AA- 100). The procedure revision does not change any system, structure or component. Therefore the proposed procedure revision does not involve a change that adversely affects a UFSAR described design function of any SSC.

2. The proposed activity, revision 6 to procedure RW-AA-100, does not involve a change to a procedure that adversely affects how UFSAR described SSC design functions are performed or controlled. The changes to the Process Control Program are administrative and involve no changes to plant equipment or processing methods in the Radwaste Building.

3. The proposed procedure implementation does not impact, revise or replace a method of UFSAR evaluation methodology. The proposed implementation of revision 6 to procedure RW-AA- 100 ensures that the proper Certificate of Compliance or Certificate of Conformance is provided to the station.. Therefore, no change is made to any evaluation methodology as stated in the UFSAR.

4. The proposed activity is neither an experiment nor a test as described in the UFSAR where the SSC is utilized or controlled in a manner that is outside the reference bounds of SSC design or is inconsistent with analyses or descriptions in the UFSAR.

5. The proposed activity is not discussed in the Technical Specifications This activity in no way alters or changes any license requirement. Therefore, this activity will not require a change to either the Operating License or Technical Specifications.

References:

UFSAR Sections 1I. I and 11.2: Radioactive Waste Management Section 11.4: Solid Waste Management Oyster Creek ODCM:

4. 1.1.1 Radioactive Liquid Waste Surveillance Requirements Table 4.11. 1.1- 1 Liquid Waste Sample and Analysis Program.

4.11 .1.3 Liquid Waste Treatment System 4.11 .1 Liquid (Waste) Effluents 4.11 .1.3 Liquid Radioactive Treatment Part 11: Liquid Effluents Oyster Creek Technical Specifications Section 6.8.1 - Written procedures shall be established.., for (Process Control Plan Implementation.)

Section 6.18 - Licensee initiated changes to the PCP Corporate Procedure

_LS-AA-106 Section 3.6.2 PORC Review Responsibilities

RM-AA-102-1001 Revision 1 Page 11 of 11 Attachment D - Record Location of Corporate Documents Record Copy of Corporate Procedure/T&RM is located .in NCS Records.

RW-AA-100 Revision 6 Page 1 of 9 Nuclear Level 3 - Information Use PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTES 1, PURPOSE 1.1. The purpose of the Process Control Program (PCP) is to:

1.1.1. Establish the process and boundary conditions for the preparation of specific procedures for processing, sampling, analysis, packaging, storage, and shipment of solid radwaste in accordance with local, state, and federal requirements. (CM-1) 1.1.2. Establish parameters which will provide reasonable assurance that all Low Level Radioactive Wastes (LLRW), processed by the in-plant waste process systems on-site OR by on-site vendor supplied waste processing systems, meet the acceptance criteria to a Licensed Burial Facility, as required by 10CFR Part 20, 10CFR Part 61, 10CFR Part 71, 49CFR Parts 171-172, "Technical Position on Waste Form (Revision 1)" [1/91], "Low-Level Waste Licensing Branch Technical Position on Radioactive Waste Classification" [5/83], and the Station Technical Specifications, as applicable.

1.1.3. Provide reasonable assurance that waste placed in "on-site storage" meets the requirements as addressed within the Safety Analysis Reports for the low level radwaste storage facilities for dry and/or processed wet waste.

2. TERMS AND DEFINITIONS 2.1. Process Control Program (PCP): The program which contains the current formulas, sampling, analysis, tests, and determinations to be made to ensure that processing and packaging of solid radioactive waste based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure the waste meets the stabilization criteria specified in 10CFR Parts 20, 61 and 71, state regulations, and burial site requirements.

2.2. Solidification

Liquid waste processed to either an unstable or stable form per 10CFR61 requirements. Waste solidified does not have to meet the 300-year free standing monolith criteria. Approved formulas, samples and tests do not have to meet NRC approval for wastes solidified in a container meeting stability (e.g. High Integrity Container).

2.3. Stabilization

Liquid waste processed to a "stable state" per 10CFR61 Requirements. Established formulas, samples, and tests shall be approved by the NRC in order to meet solidification "stabilization" criteria. This processing method is currently not available, because the NRC recognizes that waste packed in a High Integrity Container meets the 300-year stabilization criteria. In the event that this processing method becomes an acceptable method, then the NRC shall approve the stabilization formulas, samples, tests, etc.

RW-AA-100 Revision 6 Page 2 of 9 2.4. Solidification Media: An approved media (e.g. Barnwell - vinyl ester styrene, cement, bitumen) when waste containing greater than 5-year half lives is solidified in a container when the activity is greater than 1 micro curie/cc. Waste solidified in a HIC is approved by the commission meeting the 10CFR61 stabilization criteria, including 1% free standing liquids by volume when the waste is packaged to a "stable" form and < 0.5% when waste is packaged to an "unstable" form. The formulas, sampling, analysis, and test do not require NRC approval, because the HIC meets the stability criteria.

2.4.1. Solidification to an unstable or stable state are performed by vendors, when applicable. Liquid waste solidified to meet stabilization criteria (10CFR61 and 01-91 Branch Technical Requirements) must have documentation available that shows that the process is approved by the NRC or disposal facility.

2.5. Dewatering

The process of removing fluids from liquid waste streams to produce a waste form that meets the requirements of 10CFR Part 61 and applicable burial site criteria, <0.5% by volume when the waste is packaged to an "unstable" state, or

<1% by volume when the waste is packaged to a "stable" form.

2.6. High Integrity Container (HIC): A disposable container that is approved to the Requirements of 10CFR61. The use of HIC's is an alternative to solidification or encapsulation in a steel container to meet burial stability. HIC's are used to package dewatered liquid wastes, (e.g. filter cartridges, filter media, resin, sludges, etc), or dry active waste.

2.7. Encapsulation

The process of placing a component (e.g. cartridge filters or mechanical components) into a special purpose disposable container and then completely surrounding the waste material with an approved stabilization media, such as cement.

2.8. Liquid Waste Processing Systems: In-plant or vendor supplied processing systems consisting of equipment utilized for evaporation, filtration, demineralization, dewatering, compression dewatering, solidification, or reverse osmosis (RO) for the treatment of liquid wastes (such as Floor Drains, Chemical Drains and Equipment Drain inputs).

2.9. Incineration, RVR, and/or Glass Vitrification of Liquid or Solid: Dry or wet waste processed via incineration and/or thermal processing where the volume is reduced by thermal means meets 10CFR61 requirements.

2.10. Compaction: When dry wastes such as paper, wood, plastic, cardboard, incinerator ash, and etc. are volume reduced through the use of a compactor.

2.11. Waste Streams: Consist of but are not limited to

- Filter media (powdered, bead resin and fiber),

- Filter cartridges,

- Pre-coat body feed material,

- Contaminated charcoal,

RW-AA-100 Revision 6 Page 3 of 9

- Fuel pool activated hardware,

- Oil Dry absorbent material added to a container to absorb liquids

- Fuel Pool Crud

- Sump and tank sludges,

- High activity filter cartridges,

- Concentrated liquids,

- Contaminated waste oil,

- Dried sewage or wastewater plant waste,

- Dry Active Waste (DAW): Waste such as filters, air filters, low activity cartridge filters, paper, wood, glass, plastic, cardboard, hoses, cloth, and metals, etc, which have become contaminated as a consequence of normal operating, housekeeping and maintenance activities.

Other radioactive waste generated from cleanup of inadvertent contamination.

3. RESPONSIBILITIES 3.1. Implementation of this Process Control Program (PCP) is described in procedures at each station and is the responsibility of the each site to implement.

4. MAIN BODY 4.1. Process Control Program Requirements 4.1.1. A change to this PCP (Radioactive Waste Treatment Systems) may be made provided that the change is reported as part of the annual radioactive effluent release report, Regulatory Guide 1.21, and is approved by the Plant Operations Review Committee (PORC).

4.1.2. Changes become effective upon acceptance per station requirements.

4.1.3. Records of reviews performed shall be retained for the duration of the unit operating license. This documentation shall contain:

1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change, and

2. A determination which documents that the change will maintain the overall conformance of waste products to Federal (10CFR61 and the Branch Technical Position), State, or other applicable requirements, including applicable burial site criteria.

4.1.4. A solidification media, approved by the burial site, MAY BE REQUIRED when liquid radwaste is solidified to a stable/unstable state.

RW-AA-100 Revision 6 Page 4 of 9 4.1.5. When processing liquid radwaste to meet solidification stability using a vendor supplied solidification system:

1. If the vendor has its own Quality Assurance (QA) Program, then the vendor SHALL ADHERE to its own QA Program and SHALL HAVE SUBMITTED its process system topical report to the NRC or agreement state.,

2. If the vendor DOES NOT HAVE its own Quality Assurance Program, then the vendor SHALL ADHERE to an approved Quality Assurance Topical Report standard belonging to the Station or to another vendor.

4.1.6. The vendor processing system(s) is/are controlled per the following:

1. A commercial vendor supplied processing system(s) MAY BE USED for the processing of LLRW streams.

2. Vendors that process liquid LLRW at the sites must meet applicable QA Topical Report and Augmented Quality Requirements.

4.1.7. Vendor processing system(s) operated at the site WILL BE OPERATED and CONTROLLED in accordance with vendor approved procedures or station procedures based upon vendor approved documents.

4.1.8. All waste streams processed for burial or long term on-site storage SHALL MEET the waste classification and characteristics specified in 10CFR Part 61.55, Part 61.56, the 5-83 Branch Technical Position for waste classification, and the applicable burial site acceptance criteria (for any burial site operating at the time the waste was processed).

4.2. General Waste Processing Requirements 4.2.1. On-site resin processing involves tank mixing and settling, transferring to the station or vendor processing system via resin water slurry or vacuuming into approved waste containers, and, when applicable, dewatering for burial.

4.2.2. Vendor resin beds MAY BE USED for decontamination of plant systems, such as, Spent Fuel Pool, RWCU (reactor water cleanup), and SDC (Shut Down Cooling).

These resins ARE then PROCESSED via the station or vendor processing system.

4.2.3. Various drains and sump discharges WILL BE COLLECTED in tanks or suitable containers for processing treatment. Water from these tanks MAY BE SENT through a filter, demineralizer, concentrator or vendor supplied processing systems.

4.2.4. Process waste (e.g. filter media, sludges, resin, etc) WILL BE periodically DISCHARGED to the station or vendor processing system for onsite waste treatment or PACKAGED in containers for shipment to offsite vendor for volume reduction processing.

4.2.5. Process water (e.g. chemical, floor, equipment drain, etc.) MAY BE SENT to either the site waste process systems or vendor waste processing systems for further filtration, demineralization for plant re-use, or discharge.

RW-AA-100 Revision 6 Page 5 of 9 4.2.6. All dewatering and solidification/stabilization WILL BE PERFORMED by either utility site personnel or by on-site vendors or WILL BE PACKAGED and SHIPPED to an off-site vendor low-level radwaste processing facility.

4.2.7. Dry Active Waste (DAW) WILL BE HANDLED and PROCESSED per the following:

1. DAW WILL BE COLLECTED and SURVEYED and MAY BE SORTED for compactable and non-compactable wastes.

2. "DAW may be packaged in containers to facilitate on-site pre-compaction and/or off-site vendor contract requirements

3. DAW items MAY BE SURVEYED for release onsite or offsite when applicable. ,

4. Contaminated filter cartridges WILL BE PLACED into a HIC or WILL BE ENCAPSULATED in an in-situ liner for disposal or SHIPPED to an offsite waste processor in drums, boxes or steel liners per the vendor site criteria for processing and disposal.

4.2.8. Filtering devices using pre-coat media MAY BE USED for the removal of suspended solids from liquid waste streams. The pre-coat material or cartridges from these devices MAY BE routinely REMOVED from the filter vessel and discharged to a Filter Sludge Tank or Liner/HIC. Periodically, the filter sludge MAY BE DISCHARGED to the vendor processing system for waste treatment onsite or PACKAGED in containers for shipment to offsite vendor for volume reduction processing.

4.2.9. Activated hardware stored in the Spent Fuel Pools WILL BE PROCESSED periodically using remote handling equipment and MAY then BE PUT into a container for shipment or storage 4.2.10. High Integrity Containers (HIC):

1. For Barnwell disposal vendors who supply HIC's to the station MUST PROVIDE a copy of the HIC Certificate of Compliance, which details specific limitations on use of the HIC.

2. For Disposal at Clive vendors who supply HIC's to the station MUST PROVIDE a copy of the HIC Certificate of Conformance, which details specific limitations on use of the HIC.

3. Vendors who supply HIC's to the station MUST PROVIDE a handling procedure, which establishes guidelines for the utilization of the HIC. These guidelines serve to protect the integrity of the HIC and ensure the HIC is handled in accordance with the requirements of the Certificate of Compliance or Certificate of Conformance.

4.2.11. Lubricants and oils contaminated as a consequence of normal operating and maintenance activities MAY BE PROCESSED on-site (by incineration, for oils meeting 10CFR20.2004 and applicable state requirements, or by an approved vendor process) or SHIPPED offsite (for incineration or other acceptable processing method).

RW-AA-100 Revision 6 Page 6 of 9 4.2.12. Former in-plant systems GE or Stock Drum Transfer Cart and Drum Storage Areas MAY BE USED for higher dose DAW storage at Clinton, Dresden, Quad Cities, Braidwood and Byron.

4.2.13 Certain waste, including flowable solids from holding pond, oily waste separator, cooling tower basin and emergency spray pond, may be disposed of onsite under the provisions of 10CFR20.2002 permit. Specific requirements associated with the disposal shall be incorporated into station implementing procedures. (CM-2) 4.3. Burial Site Requirements 4.3.1. Waste sent directly to burial WILL COMPLY with the applicable parts of 49CFR, 10CFR61, and 10CFR71, and the acceptance criteria for the applicable burial site.

4.4. Shipping and Inspection Requirements 4.4.1. All shipping/storage containers WILL BE INSPECTED, as required by station procedures, for compliance with applicable requirements (Department Of Transportation (DOT), Nuclear Regulatory Commission (NRC), station, on-site storage, and/or burial site requirements) prior to use.

4.4.2. Containers of solidified liquid waste WILL BE INSPECTED for solidification quality and/or dewatering requirements per the burial site, offsite vendor acceptance, or station acceptance criteria, as applicable.

4.4.3. Shipments sent to an off site processor WILL BE INSPECTED to ensure that the applicable processor's waste acceptance criteria are being met.

4.5. Inspection and Corrective Action 4.5.1. Inspection results that indicate non-compliance with applicable NRC, State, vendor, or site requirements WILL BE IDENTIFIED and TRACKED through the Corrective Action Program.

4.5.2. Administrative controls for preventing unsatisfactory waste forms from being released for shipment are described in applicable station procedures. If the provisions of the Process Control Program are not satisfied, then SUSPEND shipments of defectively packaged radioactive waste from the site. (CM-1) 4.5.3. If freestanding water or solidification not meeting program requirements is observed, then samples of the particular series of batches WILL BE TAKEN to determine the cause. Additional samples WILL BE TAKEN, as warranted, to ensure that no freestanding water is present and solidification requirements are maintained.

4.6. Procedure and Process Reviews 4.6.1. The Exelon Nuclear Process Control Program and changes to it (other than editorial/minor changes) SHALL BE REVIEWED and APPROVED in accordance with the station procedures, plant-specific Technical Specifications (Tech Spec),

Technical Requirements Manual (T&RM), Operation Requirements Manual (ORM),

as applicable, for the respective station and LS-AA-106. Changes to the Licensees Controlled Documents, UFSAR, ORM, or TRM are controlled by the provisions of 10CFR 50.59.

RW-AA-1 00 Revision 6 Page 7 of.9 4.6.2. Any changes to the PCP shall be reviewed to determine if reportability is required in the Annual Radiological Effluent Release Report (ARERR). The Radwaste Specialist shall ensure correct information is submitted to the ODCM program owner prior to submittal of the ARERR.

4.6.3. Station processes, cask manual procedures as applicable to your station, or other vendor waste processing/operating procedures shall be approved per RM-AA-102-1006. Procedures related to waste manifests, shipment inspections, and container activity determination are CONTROLLED by Radiation Protection Standard Procedures (RP-AA-600 Series).

1. Site waste processing IS CONTROLLED by site operating procedures.

2. Liquid processed by vendor equipment WILL BE DONE in accordance with vendor procedures.

4.7. Waste Types, Point of Generation, and Processing Method Methods of processing and individual vendors MAY CHANGE due to changing financial and regulatory options. The table below is a representative sample. It is not intended be all encompassing.

Waste Stream POINTS OF GENERATION \ AVAILABLE WASTE PROCESSING METHODS Bead Resin Systems - Fuel Pool, Condensate, Dewatering, solidification to an Reactor Water Cleanup, Blowdown, unstable/stable state Equipment Drain, Chemical and Thermal Processing Volume Control Systems, Floor Drain, Maximum Recycle, Blowdown, Free Release to a Land Fill Boric Acid Recycling System, Vendor Supplied Processing Systems, and Portable Demin System Powdered Resin Systems - (Condensate System, Dewatering, solidification to an Floor Drain/Equipment Drain unstable/stable state filtration, Fuel Pool) Thermal Processing Concentrated Waste Waste generated from Site Solidification to an unstable/stable state Evaporators resulting typically from Thermal Processing the Floor Drain and Equipment Drain Systems Sludge Sedimentation resulting from various Dewatering, solidification to an sumps, condensers, tanks, cooling unstable/stable state tower, emergency spray pond, Thermal Processing holding pond, and oily waste separators.. Evaporation on-site or at an offsite processor On-site disposal per 10CFR20.2002 permit

RW-AA-100 Revision 6 Page 8 of 9 Waste Stream POINTS OF GENERATION AVAILABLE WASTE PROCESSING METHODS Filter cartridges Systems - Floor/Equipment Drains, Dewatering, solidification to an Fuel Pool; cartridge filters are unstable/stable state typically generated from clean up Processed by a vendor for volume activities within the fuel pool, torus, reduction etc.

Dry Active Waste Paper, wood, plastic, rubber, glass, Decon/Sorting for Free Release, metal, and etc. resulting from daily Compaction/Super-compaction plant activities.

Thermal Processing by Incineration or glass vitrification Sorting for Free Release Metal melting to an ingot Contaminated Oil Oil contaminated with radioactive Solidification unstable state materials from any in-plant system. Thermal Processing by Incineration Free Release for recycling Drying Bed Sludge Sewage Treatment and Waste Water Free release to a landfill or burial Treatment Facilities Metals See DAW See DAW Irradiated Hardware Fuel Pool, Reactor Components Volume Reduction for packaging efficiencies

5. DOCUMENTATION - None

6. REFERENCES 6.1. Technical Specifications:

6.1.1. The details contained in Current Tech Specs (CTS) or Improved Technical Specifications (ITS), as applicable, in regard to the Process Control Program (PCP),

are to be relocated to the Licensee Controlled Documents. Some facilities have elected to relocate these details into the Operational Requirements Manual (ORM).

Relocation of the description of the PCP from the CTS or ITS does not affect the safe operation of the facility. Therefore, the relocation details are not required to be in the CTS or the ITS to provide adequate protection of the public health and safety.

6.2. Source Documents:

6.2.1. Code Of Federal Regulations: 10 CFR Part 20, Part 61, Part 71, 49 CFR Parts 171-172 6.2.2. Low Level Waste Licensing Branch Technical Position On Radioactive Waste Classification, May 1983

RW-AA-100 Revision 6 Page 9 of 9 6.2.3. Technical Position on Waste Form (Revision 1), January 1991 6.2.4. Branch Technical Position on Concentration Averaging and Encapsulation, January 1995 6.2.5. Regulatory Guide 1.21 6.2.6. I.E. Circular 80.18, 10CFR 50.59 Safety Evaluation for Changes to Radioactive Waste Treatment Systems 6.2.7. Quality Assurance Program 6.2.8. LS-AA-106 6.2.9. RM-AA-1 02-1006 6.2.10. RP-AA-600 Series 6.3. Station Commitments:

6.3.1. Peach Bottom CM-1, T03819, Letter from G.A. Hunger, Jr., dated Sept. 29,94, transmitting TSCR 93-16 (Improved Technical Specifications).

6.3.2. Limerick CM-2, T03896, 10CFR20.2002 permit granted to Limerick via letter dated July 10, 1996.

7. ATTACHMENTS - None

Procedure Approval Form AD-AA-1 01 -F-01 Page 1 of 1 Revision 1 Document Number: RW-AA-100 Revision: 7

Title:

Process Control Program for Radioactive Wastes El New 0 Cancel El Cancel 0 Revision EC#: _ PCR#: PPIS#:

Document Revision 99 Editorial 0l Batch ERR#: AR#:

o Supersede corporate document(s) List:

Revision Summary: This revision has been prepared to correct and standardize the use of "shall", "should",

Attach add'i descript, if req'd "may", and "will" in accordance with AD-AA-101-1002, moves records retention requirements to section 5, corrects reference at 4.3.1. Several other format corrections were also made.

Impact on Operating Z N/A and Design Margins:

Attach addl descript, if req'd CONFIRM that n3 commitments (i.e., those steps annotated with CM-X) have been changed or deleted unless evaluated via completion of LS-AA-1 10 commitment change/deletion form and INITIAL [Prepareri: RMC Preparer: RM Claes 10/02/2008 CAN/3214 Print Date Location/Ext Applicable BR 0 Marcia Morris DR 0 Sandy Llveechl QC 0 Terry Barber Site Contacts BY 0 Norma Jean Gordon LA 0 Lynn Kofold-Durdan CL 0 Robert Campbell Check box and PB (M George Tharpe 0 0] Michael Seeloff LG (R Linda Knapp provide name TMIO1 Jesaica Spagnuolo ZN [] Other [3 Validation Req'd: 0 No E3 Yes (attach) Common Training Req'd: 0 No 0 Yes (Validation requirement see AD-AA-101) Print/Signature Site Specific Training Req'd: 0 No [3 Yes Change Management: 0 HU.AA-1 101 Chan o C ecklist Attached [ Document Traveler 0 None Required Level of Use: D Level 1 - Contin SU e vel 2- Re once Use 0 Level 3 - Information Use Approval Mliuel Azarl/ 10/03/06 CAN/3200 CFAM (Standard Procedugs f Print/Sign Date Location/Ex:

Approval Site Document(s) to be superseded:

S W/A Location: Op, 0A{l)

NIUse additional sheets as necessary. Assure that all pending changes are dlapositlonead o Temp. Change [] Interim Change Temp or Interim Change #: Interim Change expiration:

10CFR50.59 Applicable: r No$ Yes Tracking Number: ( '- (

10CFR72.48 Applicable: ; No C3 Yes Other Regulatory Process Applicable: .fNo 0 Yes Other Regulatory Process Number:

PORC Required: C3 No;"%Yes PORC Number (after PORC Approved): '( la Environmental Review Required:$ No 0 Yes If "Yes" then attach completed EN-AA-103 Attachment 1.

13 Itsuperceding a document containing commitments, notify the Commitment Tracking Coordinator per LS-AA.1 10 so the CTO can be updated as appropriate.

Cross Discipline Reviews: (list below) ur ell !ce "ordInator Review Req'd 0l No 0 Yes, list below

-FAL~WARA I R&A13 4A1 j_> _ /11-8 frP Print ;S gnat re Dale - inp -oer-0g.7 Sgature D"iscipine or Org.

snPrint I

Print Signature Dale Oisdcline or Org.

Attach additional it reg'd Tamp Change I' Authorization Only SRO PrInt/Sign/Date SOR Print/Sign/Date Impl. Date Exp. Date SOR Approval indicates that all required Cross-Disciplinary reviews have been performed and the reviewers have signed this form. This procedure is technicaily and functionally accurate for allfunctional rea (Refer to AD-AA-102)

SOR Approval: , e Site Authorization: _ ,.rmej

,7 , ./

'"plant nage r ll (O npg relquireftlirroedure) , Dbte(

'U*"~"

Date:

Subject:

PORC Chair Subcommittee Approval The topic listed below is assigned to a subcommittee to assess the potential effect on nuclear safety.

TOPIC: Uw)-F /0 0, ecv. 7 09occss coAvTOoi, 1ZO&RA,,M FOR RAblolfcrnv c 4 SinS OWNER: 7" 6,e ITT I Item APPROVED for Subcommittee Review Item NOT APPROVED for Subcommittee Review PORC Chair: -"aV/oI Date 11_

_ _168 SUBCOMMITTEE DISCIPLINES:

PORC Chair: ,5 4.TIU_'? 7"-aylov Date:

To Subcommittee Member(s):

The subcommittee is to determine safety significance of the item. Some items may have a minimal effect on nuclear safety and may not warrant full committee review.

Therefore, the PORC Chair assigns items to a subcommittee to screen items submitted for review based on safety significance.

The subcommittee shall complete the Nuclear Safety Significance Assessment form (Attachment 3) for each item and forward the conclusions to the PORC Coordinator, in 5 working days. Due Date: I 7t,-0 9OtIf the subcommittee concludes that the item warrants a full PORC review based on the criteria in Attachment 3, the PORC Coordinator will include the item in the PORC agenda for full PORC review. If the subcommittee concludes that the item does not warrant full PORC review based on criteria in Attachment 3, the PORC Coordinator informs the PORC during the next available meeting and documents those items that were not reviewed by full PORC in the minutes of the meeting.

Return Attachment 3 to the PORC Coordinator- V1t M' SCyR,?_

Attachments: Attachment 3 of LS-AA-106 - complete one Attachment 3 per item.

PORC #08- 0 73

LS-AA-106 Revision 4 Page 20 of 23 ATTACHMENT 3 Nuclear Safety Significance Assessment Form Page 1 of I Item

Title:

Ju0/* P_ *W 09

'- 7.;

Item Number: 4?-R/91 -/10 Revision No. 7 C-x 0 C665 L0oU73*oL t-&OOKA11% Foie /rAýA/O090IV- W.,I/40 S Does the proposed change: YES NO 1 Result in a change to a procedure affecting Technical Specifications, ECCS, ESF, or PRA risk significant equipment or systems?

- 7 2 Result in a modification or change to a ECCS, ESF, or PRA risk significant -

V/

system?

3 Consist of a major change to the facility and/or a.major test or experiment? -

4 Consist of a major change to a plant process?

5 Change the qualification or operational characteristics of installed components 7 or systems classified as safety related.

6 Change the nuclear safety response of the plant to normal evolutions, V/

anticipated operational occurrences, or design basis accidents?

7 Have the potential to reduce the ability of the operator to assess or control the -

nuclear safety status of the plant?

8 Result from investigations of significant operational abnormalities including - 7 accidental unplanned or uncontrolled radioactive releases?

9 Increase the potential for a plant trip or present a challenge to safety systems? -

10 Require NRC approval prior to implementation, e.g., TS, Security Plan, - 7 Emergency Plan?

11 10 CFR 50.59/10 CFR 72.48 written evaluation be prepared? vd IT any answer iu mne aove quesilons 1s - Yes ,eaull rrI"U6 review is requlrea. Ivall questions are answered "No", a full PORC review is not required. Assumptions must be documented in the Comments Section below.

COMMENTS (Use additional pages as necessary)

Circle as applicabl THIS ITEM DOE DOES NO E FULL PORC REVIEW.

PORO Member's Name " Department: ..

POR embe'sD :nt_

PORC Member's * . ,, )Date: */"10 K :. Si~nature--

LS-AA-106 Revision 4 Page 20 of 23 ATTACHMENT 3 Nuclear Safety Significance Assessment Form Page 1 of I Item

Title:

POX/C -w- Of - 0 7; Item Number: j-?4)-f. ~-/00 Revision No. 7 P90~cC.SS C o jne 0o1 Pk~ooetl role 1Pnb)oo~cryC ti 5m Does the proposed changle, YES NO 1 Result in a change to a procedure affecting Technical Specifications, ECCS, ESF, or PRA risk significant equipment or systems? /

2 Result in a modification or change to a ECCS, ESF, or PRA risk significant system?

3 Consist of a major change to the facility and/or a major test or experiment?

4 Consist of a major change to a plant process?

5 Change the qualification or operational characteristics of installed components or systems classified as safety related.

6 Change the nuclear safety response of the plant to normal evolutions, anticipated operational occurrences, or design basis accidents?

7 Have the potential to reduce the ability of the operator to assess or control the nuclear safety status of the plant?

8 Result from investigations of significant operational abnormalities including accidental unplanned or uncontrolled radioactive releases?

9 Increase the potential for a plant trip or present a challenge to safety systems?

10 Require NRC approval prior to implementation, e.g., TS, Security Plan, Emergency Plan?

11 10 CFR 50.59/10 CFR 72.48 written evaluation be prepared?

If any answer to the above questions is res, a TuII P-uu review is requireo. HTall questions are answered "No", a full PORC review is not required. Assumptions must be documented in the Comments Section below.

COMMENTS (Use additional pages as necessary)

Circle as applicable ----.4 THIS ITEM DOESOES N REQUIRE FULL PORC REVIEW.

PORC Member's Name f.H f J*-'.*...- Department: FRP Print PORC Member's c(TA*. Date: 00l10 (0 Signature

LS-AA-1 06 Revision 4 Page 20 of 23 ATTACHMENT 3 Nuclear Safety Significance Assessment Form Page 1 of I item

Title:

POX* C t 0 9' - 0 7,,*

Item Number: ._ O-i -/oo Revision No. "7 AeoccSS C0U7-9oL- R10Jm~ FOR 1PAhoQACz'1VC eJTFSm Does the proposed change: YES NO 1 Result in a change to a procedure affecting Technical Specifications, ECCS, ESF, or PRA risk significant equipment or systems?

2 Result in a modification or change to a ECCS, ESF, or PRA risk significant system?

3 Consist of a major change to the facility and/or a major test or experiment?

4 Consist of a major change to a plant process?

5 Change the qualification or operational characteristics of installed components v/

or systems classified as safety related.

6 Change the nuclear safety response of the plant to normal evolutions, - 7 anticipated operational occurrences, or design basis accidents?

Have the potential to reduce the ability of the operator to assess or control the V/

nuclear safety status of the plant?

8 Result from investigations of significant operational abnormalities including V/7 accidental unplanned or uncontrolled radioactive releases?

9 Increase the potential for a plant trip or present a challenge to safety systems? V 10 Require NRC approval prior to implementation, e.g., TS, Security Plan, Emergency Plan?

11 aS ..

10 CFR 50.59/10

.. .... AL, CFR 72.48 written

-I.. ..... . ... A!....---

evaluation

&a.

be prepared?

.a,lN/-- -- S..ll nrt'%~rm P .... A* ... .-- .. ... .--J SI -I IT any answer vo ne aoove questions is Tes , a lull ruPRt review is requireo. iTaii questions are answered "No", a full PORC review is not required. Assumptions must be documented in the Comments Section below.

COMMENTS (Use additional pages as necessary)

Circle as applicable THIS ITEM DOE =ýOE, SIOE REQUIRE FULL PORC REVIEW.

PORC Member's Name idW044.. , Department: ýIARI Print PORC Member's 1 ",1111 Date:__________

6,'Signature (0

RW-AA-100 Revision 6 Changes Procedure Procedure Rev 6 Changes Rev 7 Step Rev 6 Step Rev 7 ________________

2.4.1 2.4.1 must shall 4.1.3. Step 5.0 4.1.3: was moved The entire step was removed and moved to Ste 5. DOCUMENTATION 4.14 4.13 MAY BE REQUIRED may be REQUIRED 4.1.5.1 4.1.4.1 SHALL ADHERE and SHALL HAVE shall ADHERE" shall have SUBMITTED SUBMITTED",

4.1.5.2 4.1.4.2 SHALL ADHERE and SHALL HAVE shall ADHERE" shall have SUBMITTED SUBMITTED" 4.1.6.1 4.1.5.1 MAY BE USED may be USED 4.1.6.2 4.1.5.2 must MUST 4.1.7 4.1.6 WILL BE OPEATED and Shall be OPERATED and CONTROLLED CONTROLLED 4.1.8 4.1.7 SHALL MEET shall MEET 4.2.1 4.2 Note Deleted and made into a note A note was added explain the process of getting waste from station tanks to the

____________________________ processing liners.

4.2.2 4.2.1 MAY BE USED, ARE then may be USED, are then PROCESSED PROCESSED 4.2.3 4.2.2 WILL BE COLLECT, MAY BE SENT will be COLLECTED Will be SENT 4.2.4 4.2.3 WILL BE, or PACKAGED will be, or PACKAGED 4.2.5 4.2.4 MAY BE SENT may be SENT 4.2.6 4.2.5 WELL BE PERFORMED, or WILL BE Will be PERFORMED, or will be S_______PACKAGED and SHIPPED PACKAGED and SHIPPED 4.2.7 4.2.6 WILL BE HANDLED and Will be HANDLED and PROCESSED

________PROCESSED

Procedure Procedure Rev 6 Changes Rev 7 Step Rev 6 Step Rev 7 4.2.7.1 4.2.6.1 WILL BE COLLECTED and Will COLLECTED and SURVEYED SURVEYED and MAY BE SORTED and may be SORTED 4.2.7.3 4.2.6.3 MAY BE SURVEYED may be SURVEYED 4.2.7.4 4.2.6.4 WILL BE PLACED, or WILL BE will be PLACED, or will be ENCASULATED ENCASULATED 4.2.8 4.2.7 MAY BE may be 4.2.9 4.2.8 WILL BE, MAY then BE will be, may then be 4.2.10.1, 2, 3 4.2.9 1,2,3 MUST PROVIDE must PROVIDE 4.2.11 4.2.10 MAY BE PROCESSED may be PROCESSED 4.2.12 4.2.11 MAY BE USED may be USED 4.2.13 4.2.12 WILL COMPLY shall COMPLY 4.4.1,4.4.2, 4.4.1,4.4.2, WILL BE will be INSPECTED 4.4.3 4.4.3 4.5.1 4.5.1 WILL BE shall be 14.5.3 4.5.3 WILL BE shall be 4.6.1 4.6.1 SHALL BE shall be 4.6.2 4.6.2 submitted SUBMITTED 4.6.3.1 4.6.3.2 WILL BE shall be 4.7 4.7 MAY may 6.3 6.3 Station Commitments User References 6.3.5 CY-AA- 170-2000, Annual Radiological Effluent Release Report 1 .1.

Screening l pe. Procdur Screening Number ]Exempt322rtxeNOtus+/-

Screening Preparer.: Edward J. RowlS Screening RwiewUYJr RK. Review Date: i L2 Oa '

Screening For Dec ID: e 7 Syustems:

Tle.- Proce. Control Program for Radioactive Wastes ___

Dcrption:

This Activity is a revision to RW-A-100: Process Control Program for Radioactive Wastes'. The procedure is being revised

. from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure Iitmeets: (1) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage.

The changes made to this corporate procedure are mostly format related. Specifically, the use of the terms "will be", 'may be' and "shall be' as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writers guide. In addition, instruction on 'records' was moved from the 4. 'Main Body' to 5. 'Documentation'.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 2 Page I of I Dresden Station Units 1. 2 and 3 Station/Unit(s):

Activity/Document Number: RW-AA- 100 Rev. No.: 07.

Title:

Process Control Program for Radioactive Wastes NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

This Activity is a revision to RW-AA-100: "Process Control Program for Radioactive Wastes". The procedure is being revised from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure it meets: (I) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

The changes made to this corporate procedure are mostly formal related. Specifically, the use of the terms "will be", "may be" and "shall be" as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writer's guide. In addition, instruction on 'records' was moved from the 4. "Main Body" to 5. "Documentation".

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

The effect of this Activity is to revise the procedure to be consistent with the current revision of the writer's guide.

Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

This Activity does not involve changes to plant Safety Systems, Structures and/or Components and will not effect SSC design features and/or functions as described in the UFSAR. This procedure change will not change the probability; frequency or severity of design basis accidents or create the potential for a new type of accident not previously considered in the UFSAR. Therefore, this Activity does not increase the risk to the health and safety of the public and this 10 CFR 50.59 screening concludes that a 10 CFR 50.59 Safety Evaluation is not required.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

(NOTE: if both a Screening and Evaluation are completed, no Screening No. is required.)

Forms Attached: (Check all that apply.)

- Applicability Review x 50.59 Screening 50.59 Screening No. 2008-0322 Rev. 00 50.59 Evaluation 50.59 Evaluation No. Rev.

50.59 SCREENING FORM LS-AA-104-1003 Revision I Page l of 3

. Rev. No.: 00.

50.59 Screening No.: 2008-0322 Activity/Document No.: RW-AA-100: "Process Control Program for Radiological Wastes" Rev. No.: 07.

1. 50.59 Screening Questions (Check correct response and provide separate written response providing the basis for the answer to each question)(See Section 5 of the Resource Manual (RM) for additional guidance):

I Does the proposed Activity involve a change to an SSC that adversely affects an UFSAR _ YES X NO described design function? (See Section 5.2.2.1 of the RM)

This Activity is a revision to RW-AA-100: "Process Control Program for Radioactive Wastes". The procedure is being revised from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure it meets: (1) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage.

The changes made to this corporate procedure are mostly format related. Specifically, the use of the terms "will be", "may be" and "shall be" as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writer's guide. In addition, instruction on 'records' was moved from the 4. "Main Body" to 5. "Documentation".

The procedure change does not involve changes to Safety Related Systems, Structures or Components that adversely affects a UFSAR described functions.

2. Does the proposed Activity involve a change to a procedure that adversely affects how UFSAR ___. YES X NO described SSC design functions are performed or controlled? (See Section 5.2.2.2 of the RM)

This Activity is a revision to RW-AA- 100: "Process Control Program for Radioactive Wastes". The procedure is being revised from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure it meets: (I) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage. The changes made to this corporate procedure are mostly format related. Specifically, the use of the terms "will be",

"may be" and "shall be" as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writer's guide. In addition, instruction on 'records' was moved from the 4. "Main Body" to 5. "Documentation".

This change does not adversely affect how UFSAR described SSC design functions are performed or controlled.

3. Does the proposed Activity involve an adverse change to an element of a UFSAR described _ YES ..X.. NO evaluation methodology, or use of an alternative evaluation methodology, that is used in establishing the design bases or used in the safety analyses? (See Section 5.2.2.3 of the RM)

This Activity is a revision to RW-AA- 100: "Process Control Program for Radioactive Wastes". The procedure is being revised from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure it meets: (i) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage. The changes made to this corporate procedure are mostly format related. Specifically, the use of the terms "will be",

"may be" and "shall be" as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writer's guide. In addition. instruction on 'records' was moved from the 4. "Main Body" to 5. "Documentation".

The procedure revision does not impact UFSAR evaluation methodology that is used in either establishing the design basis or the safety analysis.

50.59 SCREENING FORM LS-AA- 104-1003 Revision I Page 2 of 3 50.59 Screening No.: 2008-0322 Rev. No.: 00.

Activity/Document No.: RW-AA-100: "Process Control Program for Radiological Wastes" Rev. No.: 07.

4. Does the proposed Activity involve a test or experiment not described in the UFSAR, where an _ YES X NO SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or is inconsistent with analyses or descriptions in the UFSAR? (See Section 5.2.2.4 of the RM)

This Activity is a revision to RW-AA-100: "Process Control Program for Radioactive Wastes".

The procedure is being revised from Revision 6 to Revision 7. This corporate standard procedure describes the process for managing solid waste to assure it meets: (I) local, state and federal requirements, (2) the acceptance criteria of a Licensed Burial Facility, and, (3) Safety Analysis Report requirements for on-site storage.

The changes made to this corporate procedure are mostly format related. Specifically, the use of the terms "will be", "may be" and "shall be" as well as words that were bold and/or capitalized were revised to be consistent with the current revision of the writer's guide. In addition, instruction on 'records' was moved from the 4. "Main Body" to 5. "Documentation".

The Process Control Program for radioactive wastes does not involve tests or experiments not described in the UFSAR, where an SSC is utilized or controlled in a manner outside the bounds of the design for that SSC or are inconsistent with analyses or descriptions in the UFSAR.

5. Does the proposed Activity require a change in the Technical Specifications or Operating _ YES X NO License? (See Section 5.2.2.5 of the RM)

The procedure change does not require a change to the Technical Specifications or the Operating License.

This corporate procedure revision does not involve changes to Safety Systems, Structures and/or Components and will not effect SSC design features and/or functions as described in the UFSAR. This procedure is associated with radwaste processing and does not change the design basis limit fortfission barriers. This procedure change will not change the probability; frequency or severity of design basis accidents or create the potential for a new type of accident not previously considered in the UFSAR. Therefore, this Activity does not increase the risk to the health and safety of the public and this 10 CFR 50.59 screening concludes that a 10 CFR 50.59 Safety Evaluation is not required.

II. List the documents (e.g., UFSAR, Technical Specifications, other licensing basis, technical, commitments, etc.) reviewed, including sections numbers where relevant information was found (if not identified in the response to each question).

Technical Specifications:

1. Tech Spec 5.5. 1: "Offsite Dose Calculation Manual"

2. Tech Spec 5.6.2: "Annual Radiological Environmental Operating Report"

3. Tech Spec 5.7: "High Radiation Areas"

4. Tech Spec 5.7.2: "Extremely High Radiation Areas" Technical Requirements Manual (TRM) Sections:

1. TRM 3.7.d: "Liquid Holdup Tanks"

2. TRM 3.7.g: "Sealed source Contamination"

3. TRM 5.5.1: "Offsite Dose Calculation Manual"

50.59 SCREENING FORM LS-AA- 104-1003 Revision I Page 3 of 3 50.59 Screening No.: 2008-0322 Rev. No.: 00_

Activity/Document No.: RW-AA-100: "Process Control Program for Radiological Wastes" Rev. No.: 07.

UFSAR Sections I. UFSAR Section 1.2.1.6: "Radioactive Waste Disposal"

2. UFSAR Section 1.2.2.11: "Shielding Access Control, and Radiation Protection Procedures"

3. UFSARTable 1.2-14 & 15: "General Arrangement, Radwaste Solidification Building (M-0OC, Sheets 1 & 2)

4. UFSAR Section 9.3.2.10: "Radwaste Building Sampling System"

5. UFSAR Section 9.4.3: "Radwaste Facility Ventilation System"

6. UFSAR Table 9.4-6: "Diagram of Radwaste Solidification Building HVAC System (Drawing M-851 & 852)

7. UFSAR Section 11.0: "Radioactive Waste Management"

8. UFSAR Section 11.4: "Solid Waste Management System"

9. UFSAR Section 13.7.5: "Radiological and Chemical Records" III. Select the appropriate conditions:

If all questions are answered NO, then complete the 50.59 Screening and implement the Activity per the applicable X governing procedure.

If question 1, 2, 3, or 4 is answered YES and question 5 is answered NO, then a 50.59 Evaluation shall be performed.

If questions I, 2, 3, and 4 are answered NO and question 5 is answered YES, then a License Amendment is required prior to implementation of the Activity.

If question 5 is answered YES for any portion of an Activity, then a License Amendment is required prior to implementation of that portion of the Activity. In addition, if question I, 2, 3, or 4 is answered YES for the remaining portions of the Activity, then a 50.59 Evaluation shall be performed for the remaining portions of the Activity.

IV. Screening Signoffs:

50.59 Screener: Edward J. Rowley Date: . / _

(Print name) 50.59 Reviewer: J. Randy Kalb Date:

(Print name)

RM-AA-102-1001 Revision 1 Page 11 of 11 Attachment D - Record Location of Corporate Documents Record Copy of Corporate Procedure/T&RM is located in NCS Records.

SRW-AA-100 x l Revision 7 Page 1 of 9 NucleaT Level 3 - Information Use PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTES

1. RPURPOSE 1.1. The purpose of the Process Control Program (PCP) is to:

1.1.1. Establish the process and boundary conditions for the preparation of specific procedures for processing, sampling, analysis, packaging, storage, and shipment of solid radwaste in accordance with local, state, and federal requirements. (CM-1) 1.1.2. Establish parameters which will provide reasonable assurance that all Low Level Radioactive Wastes (LLRW), processed by the in-plant waste process systems on-site OR by on-site vendor supplied waste processing systems, meet the acceptance criteria to a Licensed Burial Facility, as required by 10CFR Part 20, 10CFR Part 61, 10CFR Part 71, 49CFR Parts 171-172, "Technical Position on Waste Form (Revision 1)" [1/91], "Low-Level Waste Licensing Branch Technical Position on Radioactive Waste Classification" [5/83], and the Station Technical Specifications, as applicable.

1.1.3. Provide reasonable assurance that waste placed in "on-site storage" meets the requirements as addressed within the Safety Analysis Reports for the low level radwaste storage facilities for dry and/or processed wet waste.

2. TERMS AND DEFINITIONS 2.1. Process Control Program (PCP): The program which contains the current formulas, sampling, analysis, tests, and determinations to be made to ensure that processing and packaging of solid radioactive waste based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure the waste meets the stabilization criteria specified in 10CFR Parts 20, 61 and 71, state regulations, and burial site requirements.

2.2. Solidification

Liquid waste processed to either an unstable or stable form per 10CFR61 requirements. Waste solidified does not have to meet the 300-year free standing monolith criteria. Approved formulas, samples and tests do not have to meet NRC approval for wastes solidified in a container meeting stability (e.g. High Integrity Container).

2.3. Stabilization

Liquid waste processed to a "stable state" per 10CFR61 Requirements. Established formulas, samples, and tests shall be approved by the NRC in order to meet solidification "stabilization" criteria. This processing method is currently not available, because the NRC recognizes that waste packed in a High Integrity Container meets the 300-year stabilization criteria. In the event that this processing method becomes an acceptable method, then the NRC shall approve the stabilization formulas, samples, tests, etc.

RW-AA-100 Revision 7 Page 2 of 9 2.4. Solidification Media: An approved media (e.g. Barnwell - vinyl ester styrene, cement, bitumen) when waste containing greater than 5-year half lives is solidified in a container when the activity is greater than 1 micro curie/cc. Waste solidified in a HIC is approved by the commission meeting the 10CFR61 stabilization criteria, including 1% free standing liquids by volume when the waste is packaged to a "stable" form and < 0.5% when waste is packaged to an "unstable" form. The formulas, sampling, analysis, and test do not require NRC approval, because the HIC meets the stability criteria.

2.4.1. Solidification to an unstable or stable state are performed by vendors, when applicable. Liquid waste solidified to meet stabilization criteria (10CFR61 and 01-91 Branch Technical Requirements) shall have documentation available that shows that the process is approved by the NRC or disposal facility.

2.5. Dewatering

The process of removing fluids from liquid waste streams to produce a waste form that meets the requirements of 10CFR Part 61 and applicable burial site criteria, <0.5% by volume when the waste is packaged to an "unstable" state, or

<1% by volume when the waste is packaged to a "stable" form.

2.6. High Integrity Container (HIC): A disposable container that is approved to the Requirements of 10CFR61. The use of HIC's is an alternative to solidification or encapsulation in a steel container to meet burial stability. HIC's are used to package dewatered liquid wastes, (e.g. filter cartridges, filter media, resin, sludges, etc), or dry active waste.

2.7. Encapsulation

The process of placing a component (e.g. cartridge filters or mechanical components) into a special purpose disposable container and then completely surrounding the waste material with an approved stabilization media, such as cement.

2.8. Liquid Waste Processing Systems: In-plant or vendor supplied processing systems consisting of equipment utilized for evaporation, filtration, demineralization, dewatering, compression dewatering, solidification, or reverse osmosis (RO) for the treatment of liquid wastes (such as Floor Drains, Chemical Drains and Equipment Drain inputs).

2.9. Incineration, RVR, and/or Glass Vitrification of Liquid or Solid: Dry or wet waste processed via incineration and/or thermal processing where the volume is reduced by thermal means meets 10CFR61 requirements.

2.10. Compaction: When dry wastes such as paper, wood, plastic, cardboard, incinerator ash, and etc. are volume reduced through the use of a compactor.

2.11. Waste Streams: Consist of but are not limited to

- Filter media (powdered, bead resin and fiber),

- Filter cartridges,

- Pre-coat body feed material,

- Contaminated charcoal,

RW-AA-100 Revision 7 Page 3 of 9

- Fuel pool activated hardware,

- Oil Dry absorbent material added to a container to absorb liquids

- Fuel Pool Crud

- Sump and tank sludges,

- High activity filter cartridges,

- Concentrated liquids,

- Contaminated waste oil,

- Dried sewage or wastewater plant waste,

- Dry Active Waste (DAW): Waste such as filters, air filters, low activity cartridge filters, paper, wood, glass, plastic, cardboard, hoses, cloth, and metals, etc, which have become contaminated as a consequence of normal operating, housekeeping and maintenance activities.

Other radioactive waste generated from cleanup of inadvertent contamination.

3. RESPONSIBILITIES 3.1. Implementation of this Process Control Program (PCP) is described in procedures at each station and is the responsibility of the each site to implement.

4. MAIN BODY 4.1. Process Control Program Requirements 4.1.1. A change to this PCP (Radioactive Waste Treatment Systems) may be made provided that the change is reported as part of the annual radioactive effluent release report, Regulatory Guide 1.21, and is approved by the Plant Operations Review Committee (PORC).

4.1.2. Changes become effective upon acceptance per station requirements.

4.1.3. A solidification media, approved by the burial site, may be REQUIRED when liquid radwaste is solidified to a stable/unstable state.

4.1.4. When processing liquid radwaste to meet solidification stability using a vendor supplied solidification system:

1. If the vendor has its own Quality Assurance (QA) Program, then the vendor shall ADHERE to its own QA Program and shall have SUBMITTED its process system topical report to the NRC or agreement state.

2. If the vendor does not HAVE its own Quality Assurance Program, then the vendor shall ADHERE to an approved Quality Assurance Topical Report standard belonging to the Station or to another vendor.

RW-AA-100 Revision 7 Page 4 of 9 4.1.5. The vendor processing system(s) is/are controlled per the following:

1. A commercial vendor supplied processing system(s) may be USED for the processing of LLRW streams.

2. Vendors that process liquid LLRW at the sites shall MEET applicable QA Topical Report and Augmented Quality Requirements.

4.1.6. Vendor processing system(s) operated at the site shall be OPERATED and CONTROLLED in accordance with vendor approved procedures or station procedures based upon vendor approved documents.

4.1.7. All waste streams processed for burial or long term on-site storage shall MEET the waste classification and characteristics specified in 10CFR Part 61.55, Part 61.56, the 5-83 Branch Technical Position for waste classification, and the applicable burial site acceptance criteria (for any burial site operating at the time the waste was processed).

4.2. General Waste Processing Requirements NOTE: On-site resin processing involves tank mixing and settling, transferring to the station or vendor processing system via resin water slurry or vacuuming into approved waste containers, and, when applicable, dewatering for burial.

4.2.1. Vendor resin beds may be USED for decontamination of plant systems, such as, Spent Fuel Pool, RWCU (reactor water cleanup), and SDC (Shut Down Cooling).

These resins are then PROCESSED via the station or vendor processing system.

4.2.2. Various drains and sump discharges will be COLLECTED in tanks or suitable containers for processing treatment. Water from these tanks may be SENT through a filter, demineralizer, concentrator or vendor supplied processing systems.

4.2.3. Process waste (e.g. filter media, sludges, resin, etc) will be periodically DISCHARGED to the station or vendor processing system for onsite waste treatment or PACKAGED in containers for shipment to offsite vendor for volume reduction processing.

4.2.4. Process water (e.g. chemical, floor, equipment drain, etc.) may be SENT to either the site waste process systems or vendor waste processing systems for further filtration, demineralization for plant re-use, or discharge.

4.2.5. All dewatering and solidification/stabilization will be PERFORMED by either utility site personnel or by on-site vendors or will be PACKAGED and SHIPPED to an off-site vendor low-level radwaste processing facility.

RW-AA-100 Revision 7 Page 5 of 9 4.2.6. Dry Active Waste (DAW) will be HANDLED and PROCESSED per the following:

1. DAW will be COLLECTED and SURVEYED and may be SORTED for compactable and non-compactable wastes.

2. "DAW may be packaged in containers to facilitate on-site pre-compaction and/or off-site vendor contract requirements

3. DAW items may be SURVEYED for release onsite or offsite when applicable.

4. Contaminated filter cartridges will be PLACED into a HIC or will be ENCAPSULATED in an in-situ liner for disposal or SHIPPED to an offsite waste processor in drums, boxes or steel liners per the vendor site criteria for processing and disposal.

4.2.7. Filtering devices using pre-coat media may be USED for the removal of suspended solids from liquid waste streams. The pre-coat material or cartridges from these devices may be routinely REMOVED from the filter vessel and discharged to a Filter Sludge Tank or Liner/HIC. Periodically, the filter sludge may be DISCHARGED to the vendor processing system for waste treatment onsite or PACKAGED in containers for shipment to offsite vendor for volume reduction processing.

4.2.8. Activated hardware stored in the Spent Fuel Pools will be PROCESSED periodically using remote handling equipment and may then be PUT into a container for shipment or storage 4.2.9. High Integrity Containers (HIC):

1. For Barnwell disposal vendors who supply HIC's to the station shall PROVIDE a copy of the HIC Certificate of Compliance, which details specific limitations on use of the HIC.

2. For Disposal at Clive vendors who supply HIC's to the station shall PROVIDE a copy of the HIC Certificate of Conformance, which details specific limitations on use of the HIC.

3. Vendors who supply HIC's to the station shall PROVIDE a handling procedure, which establishes guidelines for the utilization of the HIC. These guidelines serve to protect the integrity of the HIC and ensure the HIC is handled in accordance with the requirements of the Certificate of Compliance or Certificate of Conformance.

4.2.10. Lubricants and oils contaminated as a consequence of normal operating and maintenance activities may be PROCESSED on-site (by incineration, for oils meeting 10CFR20.2004 and applicable state requirements, or by an approved vendor process) or SHIPPED offsite (for incineration or other acceptable processing method).

4.2.11. Former in-plant systems GE or Stock Drum Transfer Cart and Drum Storage Areas may be USED for higher dose DAW storage at Clinton, Dresden, Quad Cities, Braidwood and Byron.

RW-AA-100 Revision 7 Page 6 of 9 4.2.13 Certain waste, including flowable solids from holding pond, oily waste separator, cooling tower basin and emergency spray pond, may be disposed of onsite under the provisions of 10CFR20.2002 permit. Specific requirements associated with the disposal shall be incorporated into station implementing procedures. (CM-2) 4.3. Burial Site Requirements 4.3.1. Waste sent directly to burial shall COMPLY with the applicable parts of 49CFR171-172, 10CFR61, 10CFR71, and the acceptance criteria for the applicable burial site.

4.4. Shipping and Inspection Requirements 4.4.1. All shipping/storage containers shall be INSPECTED, as required by station procedures, for compliance with applicable requirements (Department Of Transportation (DOT), Nuclear Regulatory Commission (NRC), station, on-site storage, and/or burial site requirements) prior to use.

4.4.2. Containers of solidified liquid waste shall be INSPECTED for solidification quality and/or dewatering requirements per the burial site, offsite vendor acceptance, or station acceptance criteria, as applicable.

4.4.3. Shipments sent to an off site processor shall be INSPECTED to ensure that the applicable processor's waste acceptance criteria are being met.

4.5. Inspection and Corrective Action 4.5.1. Inspection results that indicate non-compliance with applicable NRC, State, vendor, or site requirements shall be IDENTIFIED and TRACKED through the Corrective Action Program.

4.5.2. Administrative controls for preventing unsatisfactory waste forms from being released for shipment are described in applicable station procedures. If the provisions of the Process Control Program are not satisfied, then SUSPEND shipments of defectively packaged radioactive waste from the site. (CM-1) 4.5.3. If freestanding water or solidification not meeting program requirements is observed, then samples of the particular series of batches shall be TAKEN to determine the cause. Additional samples shall be TAKEN, as warranted, to ensure that no freestanding water is present and solidification requirements are maintained.

4.6. Procedure and Process Reviews 4.6.1. The Exelon Nuclear Process Control Program and changes to it (other than editorial/minor changes) shall be REVIEWED and APPROVED in accordance with the station procedures, plant-specific Technical Specifications (Tech Spec),

Technical Requirements Manual (T&RM), Operation Requirements Manual (ORM),

as applicable, for the respective station and LS-AA-106. Changes to the Licensees Controlled Documents, UFSAR, ORM, or TRM are controlled by the provisions of 10CFR 50.59.

RW-AA-100 Revision 7 Page 7 of 9 4.6.2. Any changes to the PCP shall be reviewed to determine if reportability is required in the Annual Radiological Effluent Release Report (ARERR). The Radwaste Specialist shall ensure correct information is SUBMITTED to the ODCM program owner prior to submittal of the ARERR.

4.6.3. Station processes, cask manual procedures as applicable to your station, or other vendor waste processing/operating procedures shall be approved per RM-AA-1 02-1006. Procedures related to waste manifests, shipment inspections, and container activity determination are CONTROLLED by Radiation Protection Standard Procedures (RP-AA-600 Series).

1. Site waste processing IS CONTROLLED by site operating procedures.

2. Liquid processed by vendor equipment shall be DONE in accordance with vendor procedures.

4.7. Waste Tvyes. Point of Generation. and Processina Method Methods of processing and individual vendors may CHANGE due to changing financial and regulatory options. The table below is a representative sample. It is not intended be all encompassing.

AVAILABLE WASTE WASTE STREAM POINTS OF GENERATION POEING M E PROCESSING METHODS Bead Resin Systems - Fuel Pool, Condensate,. Dewatering, solidification to an Reactor Water Cleanup, Blowdown, unstable/stable state Equipment Drain, Chemical and Volume Control Systems, Floor Drain, Maximum Recycle, Blowdown, Boric Free Release to a Land Fill Acid Recycling System, Vendor Supplied Processing Systems, and Portable Demin System Powdered Resin Systems - (Condensate System, Floor Dewatering, solidification to an Drain/Equipment Drain filtration, Fuel unstable/stable state Pool) Thermal Processing Concentrated Waste Waste generated from Site Solidification to an unstable/stable Evaporators resulting typically from the state Floor Drain and Equipment Drain Thermal Processing Systems ThermalProcessing Sludge Sedimentation resulting from various Dewatering, solidification to an sumps, condensers, tanks, cooling unstable/stable state tower, emergency spray pond, holding Thermal Processing pond, and oily waste separators..

Evaporation on-site or at an offsite processor On-site disposal per 10CFR20.2002 I permit

RW-AA-100 Revision 7 Page 8 of 9 AVAILABLE WASTE WASTE STREAM POINTS OF GENERATION POEING ME PROCESSING METHODS Filter cartridges Systems - Floor/Equipment Drains, Dewatering, solidification to an Fuel Pool; cartridge filters are typically unstable/stable state generated from clean up activities Processed by a vendor for volume within the fuel pool, torus, etc. reduction Dry Active Waste Paper, wood, plastic, rubber, glass, Decon/Sorting for Free Release, metal, and etc. resulting from daily Compaction/Super-compaction plant activities.

Thermal Processing by Incineration or glass vitrification Sorting for Free Release Metal melting to an ingot Contaminated Oil Oil contaminated with radioactive Solidification unstable state materials from any in-plant system. Thermal Processing by Incineration Free Release for recycling Drying Bed Sludge Sewage Treatment and Waste Water Free release to a landfill or burial Treatment Facilities Metals See DAW See DAW Irradiated Hardware Fuel Pool, Reactor Components Volume Reduction for packaging efficiencies

5. DOCUMENTATION 5.1.1. Records of reviews performed shall be retained for the duration of the unit operating license. This documentation shall contain:

1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change, and

2. A determination which documents that the change will maintain the overall conformance of waste products to Federal (10CFR61 and the Branch Technical Position), State, or other applicable requirements, including applicable burial site criteria.

6. REFERENCES 6.1. Technical Specifications:

6.1.1. The details contained in Current Tech Specs (CTS) or Improved Technical Specifications (ITS), as applicable, in regard to the Process Control Program (PCP),

are to be relocated to the Licensee Controlled Documents. Some facilities have elected to relocate these details into the Operational Requirements Manual (ORM).

Relocation of the description of the PCP from the CTS or ITS does not affect the safe operation of the facility. Therefore, the relocation details are not required to be in the CTS or the ITS to provide adequate protection of the public health and safety.

RW-AA-100 Revision 7 Page 9 of 9 6.2. Writers'

References:

6.2.1. Code Of Federal Regulations: 10 CFR Part 20, Part 61, Part 71,49 CFR Parts 171-172 6.2.2. Low Level Waste Licensing Branch Technical Position On Radioactive Waste Classification, May 1983 6.2.3. Technical Position on Waste Form (Revision 1), January 1991 6.2.4. Branch Technical Position on Concentration Averaging and Encapsulation, January 1995 6.2.5. Regulatory Guide 1.21 6.2.6. I.E. Circular 80.18, 10CFR 50.59 Safety Evaluation for Changes to Radioactive Waste Treatment Systems 6.3. User References 6.3.1. Quality Assurance Program 6.3.2. LS-AA-106 6.3.3. RM-AA-102-1006 6.3.4. RP-AA-600 Series 6.3.5. CY-AA-170-2000, Annual Radioactive Effluent Release Report 6.4. Station Commitments:

6.4.1. Peach Bottom CM-1, T03819, Letter from G.A. Hunger, Jr., dated Sept. 29,94, transmitting TSCR 93-16 (Improved Technical Specifications).

6.4.2. Limerick CM-2, T03896, 10CFR20.2002 permit granted to Limerick via letter dated July 10, 1996.

7. ATTACHMENTS - None

RA-10-002, CY-OC-170-301, Revision 4, Offsite Dose Calculation Manual

Contents

Text

SUMMARY

SUBJECT:

SUMMARY

Subject:

1.2.1. APPLICABILITY

6.3 RESPONSIBILITIES

Title:

Title:

Title:

SUMMARY

SUBJECT:

SUMMARY

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References:

2.2. Solidification

2.3. Stabilization

2.5. Dewatering

2.7. Encapsulation

Title:

Subject:

Title:

Title:

Title:

Title:

2.2. Solidification

2.3. Stabilization

2.5. Dewatering

2.7. Encapsulation

References:

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RA-10-002, CY-OC-170-301, Revision 4, Offsite Dose Calculation Manual

Text

SUMMARY

SUBJECT:

SUMMARY

Subject:

1.2.1. APPLICABILITY

6.3 RESPONSIBILITIES

Title:

Title:

Title:

SUMMARY

SUBJECT:

SUMMARY

Title:

References:

2.2. Solidification

2.3. Stabilization

2.5. Dewatering

2.7. Encapsulation

Title:

Subject:

Title:

Title:

Title:

Title:

2.2. Solidification

2.3. Stabilization

2.5. Dewatering

2.7. Encapsulation

References:

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