ML20235U372
| ML20235U372 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/1988 |
| From: | Cartwright W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 89-108, NUDOCS 8903090173 | |
| Download: ML20235U372 (99) | |
Text
. _ _ - _ - _ _ _ _ _ _ _ - _
RADI0 ACTIVE EFFLUENT RELEASE REPORT FOR THE NORTH ANNA POWER STATION
( JULY 01, 1988 TO DECEMBER 31, 1988 )
INDEX SUBJECT PAGE SECTION NO.
1 PURPOSE AND SCOPE . . . . . . . . . . . . . . . 1 2 DISCUSSION .................. 2 SUPPLEMENTAL INFORMATION ........... 4 3
5 Attachment 1 Effluent Release Data .................
Attachment 2 Annual and Quarterly Doses . . . . . . . . . . . . . . . 6 7
Attachment 3 Revisions to Offsite Dose Calculation Manual (0DCM) ..
Attachment 4 Revisions to Process Control Program (PCP) . . . . . . . 8 Attachment 5 Vajor changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems ............. 9 Attachment 6 Inoperability of Radioactive Liquid and Gaseous Effluent Instrumentation. . . . . . . . . . . . . . . . . . . . . 10 Attachment 7 Unpl anned Rel eases . . . . . . . . . . . . . . . . . . . 11 Attachment 8 Changes Required By The Land Use Census Evaluation . . . 12 Attachment 9 Lower Limits of Detection (LLD) for Effluent Sample Analysis . . . . . . . . . . . . . . . . . . . . . . . . 13/14 Attachment 10 Unavailability of Milk or Leafy Vegetation Samples with Replacement Sampling Locations. . . . . . . . . . . 15 1
8903090173 0812 7 -n38 py
Page 1 1.0 PURPOSE AND SCOPE The Radioactive Effluent Release Report includes in Attachment 1, a summary of the quantities of radioactive liquid and gaseous effluents and solid waste 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 of Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Tables 1, 2 and 3 of Appendix B thereof. The report submitted within 60 days after January 1 of each year includes an assessment of radiation doses i i
to the maximum exposed member of the public due to radioactive liquid and ;
gaseous effluents released from the site during the previous calendar year.
The report submitted within 60 days after July 1 of each year has the same sections except for the assessment of radiation doses. The report also includes a list of unplanned releases during the reporting period, in Attachment 7.
(
As required by Technical Specification 6.15.2 changes to the ODCM for the j time period covered by this report are included in Attachment 3. Information is provided to support the changes along with a package of those pages of the )
ODCM changed.
This report includes changes to the PCP with information and documentation necessary to support the rationale for the changes as required by Technical I
Specification 6.14.1, in Attachment 4.
Major changes to radioactive liquid, gaseous and solid waste treatment systems are reported in Attachment 5, as required by Technical Specification 6.16. Information to support the reason (s) for the change (s) and a summary
g . -
- 'Page 2' of the '.10lCFR 50.59 evaluation are included. In lieu of reporting major changes in this report, major _ changes to'the radioactive waste treatment systems may be submitted'as-part'of the annual FSAR update.
As required by Technical Specification 3.3.3.10.b and 3.3.3.11.b, a-list and explanation fcr the inoperability of radioactive liquid and/or_ gaseous effluent monitors are provided in this report, in Attachment 6.
2.0 DISCUSSION The basis for the calculation of the percent of technical specification for the critical ~ organ in Table 1 A of Attachment 1, is Technical Specification 3.11.2.1.b. Technical Specification 3.11.2.1.b requires that the dose rate for iodine-131, for ' tritium, and for all radionuclides in particulate form with. half-lives greater than 8 days shall be less than or equal to 1500 mrem /yr to the' critical organ at or beyond the. site boundary. The critical
-organ is the child's thyroid; inhalation pathway.
The basis for the calculation of percent of technical specification for the total body and skin in Table 1 A of Attachment 1, is Technical Specification 3.11. 2.1. a . Technical Specification 3.11.2.1.a requires that the dose rate for noble gases to areas at or beyond site boundary shall be less than or
~
equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.
The basis for the calculation of the percent of technical specification in Table 2A of Attachment 1, is Technical Specification 3.11.1.1. Technical Specification 3.11.1.1 states that the concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the L _ _ _ - _ _ _ - _- _ _ __ - _ __ _ __ - _ - _ - - __ _ __-_ - _ - _ _ _ _ _ _ _ _ _
H Page 3 ,
l
' concentrations specified~in 10 CFR 20, . Appendix B. Table II, Column 2 for
- r. .
radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be~ limited to 2.0E-4 i
microcuries/ml . !
I
, Percent of technical specification calculations are based on the total l gaseous or liquid effluents released for that respective quarter.
The annual and quarterly doses, as reported in Attachment 2, were calculated .!
according to the methodology presented in the ODCM. The beta and gamma air l
doses due to noble gases released from the site were calculated at site boundary. The maximum exposed member of the public from the releases of airborne iodine-131, tritium and all radionuclides in particulate form with half lives greater than 8 days, is defined as an infant, exposed through the o
grass-cow-milk pathway, with the critical organ being the thyroid. The f maximum exposed member of the public from radioactive materials in liquid effluents in unrestricted areas is defined as an adult, exposed by either the l r !
invertebrate or fish pathway, with the critical organ being the liver. The. I total body dose was also determined for this ind'ividual, f Unplanned releases presented in Attachment 7 are defined according to the criteria presented in 10 CFR 50.73, as those gaseous radioactive releases that exceed 2 times the applicable concentrations of the limits specified in Appendix B, Table II, of 10 CFR 20 in unrestricted areas, when averaged over a time period of one hour, and/or as those liquid radioactive releases that exceed 2 times the limiting combined Maximum Permissible Concentration (MPC) specified in Appendix B, Table II, of 10 CFR 20 in unrestricted areas for all J
i Page 4 radionuclides except tritium and dissolved noble gases, when averaged over a i time period of one hour.
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Technical Specifications 3.12.2 and 6.9.1.9 require the identification of new sample locations determined with the Land Use Census as yielding a calculated dose or dose commitment greater than the values currently being calculated in Technical Specification 4.11.2.3. These, if any, are identified in Attachment 8.
The typical Lower Limit of Detection ( LLD ) capabilities of the radioactive effluent analysis instrumentation are presented in Attachment 9. These Lower Limit of Detection values are based upon conservative conditions (i.e.,
minimum semple volume and maximum delay time prior to analysis). Actual Lower Limit of Detection values may be lower. If a radioisotope is not detected when analyzing effluent samples, then the activity of that radioisotope will be reported as Not Detectable (N/D) on Attachment 1 of this report. When all radioisotopes listed on Attachment 1 for a particular quarter and release mode are less than the Lower Limits of Detection, then the totals for this period will be designated as Not Applicable (N/A).
Technical Specification 3.12.1.c requires the identification of the cause of the unavailability of milk or leafy vegetation samples, required by Technical Specification Table 4.12-1, and identification of the new location (s) for obtaining replacement samples. This information is presented in Attachment 10.
3.0 SUPPLEMENTAL INFORMATION There are no inclusions for the time period covered by this report. '
Page 5 ATTACHMENT 1 EFFLUENT RELEASE DATA
( 07/88 - 12/88 )
This attachment includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste, as outlined in Regulatory Guide 1.21.
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TABLE 3 NORTH ANNA POWER STAT 10N r RADI0 ACTIVE EFFLUENT RELEASE REPORT SUMMATION OF SOLID RADI0 ACTIVE WASTE AND IRRADIATED FUEL SHIPMENTS FOR 07-01-88 THROUGH 12-31-88 Page 1 of 2 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) 6-MONTH ESTIMATED TOTAL
- 1. Type of Waste UNIT PERIOD PERCENT ERROR (%)
- a. Spent resins, filter sludges, m3 7.43E+1* 2.50 E+1 evaporator bottoms, etc. Ci 2.02E+0 2.50 E+1
- b. Dry compressible waste, m8 3.14E+1** 2.50 E+1 contaminated equipment, etc. Ci 5.71E-1 2.50 E+1
- c. Irradiated ~ components, control m3 0.00E0 0.00 E0 rods, etc. Ci 0.00E0 0.00 E0
- d. Other(describe) ms 6.25E+0*** 2.50 E+1 1 Contaminated Oil Ci 3.97E-2 2.50 E+1
- 2. Estimate of major nuclide composition (by type of waste)
- a. Iron - 55 % 1.99E+1 2.50 E+1 Antimony - 125 % 1.60E+1 2.50 E+1 Cesium - 137 % 1.60E+1 2.50 E+1 Cobalt - 60 % 1.45E+1 2.50 E+1 Cobalt - 58 % 9.22E+0 2.50 E+1 l Cesium - 134 % 8.77E+0 2.50 E+1 l Nickel - 63 % 7.96E+0 2.50 E+1 i Plutonium - 241 % 2.69E+0 2.50 E+1 Manganese - 54 % 2.20E+0 2.50 E+1 Silver - 110m % 1.14E+0 2.50 E+1
% . E . E
% . E . E
]
- b. Cobalt - 60 % 4.12E+1 2.50 E+1 Iron- 55 % 3.24E+1 2.50 E+1 l Nickel - 63 % 8.54E+0 2.50 E+1 Cobalt - 58 % 5.55E+0 2.50 E+1 l Cesium - 137 % 3.76E+0 2.50 E+1 Cesium - 134 % 2.44E+0 2.50 E+1 Silver - 110m % 2.20E+0 2.50 E+1 i Manganese - 54 % 2.18E+0 2.50 E+1
% . E . E
% . E . E l
- c. None % . E . E
% . E . E l
% . E . E l
% . E . E l 1
l l
l 1
l
TABLE 3 NORTH ANNA POWER STATION RADI0 ACTIVE EFFLUENT RELEASE REPORT SUMMATION OF SOLID RADIOACTIVE WASTE AND IRRADIATED FUEL SHIPMENTS FOR 07-01-88 THROUGH 12-31-88 Page 2 of 2
- 2. Estimated of major nuclide composition (by type of 6-MONTH ESTIMATED TOTAL waste) (cont.) UNIT PERIOD PERCENT ERROR (%)
55 % 1.22E+1 2.50 E+1 Cerium - 144 % 1.10E+ 1 2.50 E+1 Cobalt - 60 % 1.50E+0 2.50 E+1
% . E . E
- 3. Solid Waste Disposition NUMBER OF SHIPMENTS MODE OF TRANSPORTATION DESTINATION 6 Truck Barrwell, SC 3 Truck Oak Ridge, TN B. IRRADIATED FUEL SHIPMENTS (Disposition)
NUMBER OF SHIPMENTS MODE OF TRANSPORTATION DESTINATION 0 N/A N/A
- 1 shipment of Powdex Resin was shipped from North Anna to a licensed waste processor for volume reduction. Therefore the volume as listed for this waste type is not representative of the actual volume buried. The total volume buried for the reporting period was 56.8 m3
- 1 shipment of Dry Compressible Waste was shipped from North Anna to a licensed waste processor for volume reduction. Therefore the volume as listed for this waste type is not representative of the actual volume buried. The total volume buried for the reporting period was 45.6 m3
- 1 shipment of Contaminated Oil was shipped from North Anna to a licensed waste processor for incineration. Therefore the volume as listed is not representative of the actual volume buried. The total volume buried for the reporting period was 0 m3
L..
Pega 6 i
ATTACEMENT 2 ANNUAL AND QUARTERLY DOSES (7/88 - 12/88)
An assessment of radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents released from the site for each calendar quarter for the calendar year of this report along with an annual total of each effluent pathway will be made pursuant to Technical Specification 6.9.1.9.
Liquid Effluents:
1st 2nd 3rd 4th Annual Quarter Quarter Quarter Quarter Total Total Body Dose (mrem) 3.39E-1 2.35E-1 1.90E-1 4.36E-1 1.20E+0 Critical Organ Dose (mrem) 4.16E-1 2.76E-1 1.96E-1 5.53E-1 1.44E+0 Gaseous Effluents:
i lst 2nd 3rd 4th Annual ]
Quarter Quarter Quarter Quarter Total l l
Noble Gas Gamma Dose (mrad) 5.72E-3 2.27E-3 6.17E-4 2.30E-2 3.16E-2 Noble Gas Beta Dose (mrad) 1.70E-2 6.75E-3 1.94E-3 6.85E-2 9.42E-2 Critical Organ
' Dose for I-131, i
H-3, Particulate 3.70E-2 1.55E-2 1.47E-2 2.57E-2 9.29E-2 with Tk > 8 days i (mrem)
Page 7 ATTACHMENT 3 (07/88 - 12/881 REVISIONS TO OFFSITE DOSE CALCULATION MANUAL (ODCMJ As required by Technical Specification 6.15, revisions to the ODCM for the time period covered by this report are included. The revisions are synopsized below. A list of the supporting documentation and affected pages of the ODCM are provided on the next page.
12-29-88: Revisions were made to the ODCM Section 4 to correct typographical errors in example calculations, half-lives for Zn-65 and Sb-125, dilution factors and site related ingestion dose commitment factors dependent on these half-lives.
12-29-88: Revisions were made to the ODCM Section 9 to correct typographical errors, example calculations, and the Rypy I-131 dose factor in Table 9.0.
L-_ --_--_-- - __ - __ _ _ _ _ _ _ _ _ _ _ _ __ _ _ __ _ _ _ _ _
l 1
LIST OF SUPPORTING DOCUMENTATION FOR REVISIONS TO THE
)
1 0FFSITE DOSE CALCULATION MANUAL SNSOC Approved Technical Justification for Change to 0DCM-4 ;
1 SNS0C Approved Request to Change Procedure and Routing Form with the Procedure Review Checklist for ODCM-4 Deleted-version of Section 4 of ODCM (crossed-out pages)
Revised Section 4 of the ODCM SNS0C Approved Technical Justification for Change to 0DCM-9 SNSOC Approved Request to Change Procedure and Routing Form with the Procedure Checklist for ODCM-9 Deleted version of Section 9 of ODCM (crossed-out pages)
Revised Section 9 of the ODCM Other supporting documentation:
.- Section A1, Meteorological Analysis of ODCM
. Section A3, Gaseous Pathway Analysis of ODCM i
. Regulatory Guide 1.109, Calculation of Annual Doses to Man from i Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I (Appendix E, Table E-1, and Table E-14 only)
. Radionuclides Transformations, Energy and Intensity of Emissions by the International Commission on Radiological Protection (Table for 53 - Iodine - 131 only)
I I
l
~
V TECHNICAL JUSTIFICATION TO CHANGE ODCM-4 IAW TS 6.15.2 4 kg/
h/Mb
- a. It is requested to change the following items of ODCM-4 (see attached pages showing corrections and newly approved and dated pages for detail)
- 1. Front Page correct typographical error of page number
.. 2. Example on page 4 de correct numeric errors in example calculation. The exemple contains incorrect numbers caused by typographical errors and errors in transferring numeric factors from Table 4.0. Of the two examples the overall result changes only for one sample insignificant 1y from .289 mrem to .299 mrem.
- 3. Table 4.0 The individual Dilution Factors ( fi ) and the Site Related Liquid Pathway Ingestion Dose Factors ( Ai ) change for Zn-65 and Sb-125 based on wrong halflives used for their calculation. Also corrected I are Critical organ, A1 values for Rb-86 by 1.3% and I-132 by 6%
which contain typographical errors. In addition old Ai values (first column under Total Body Ai and Critical Organ A1) which erroneously remained on this table after the 10-23-86 revision, are removed.
- 4. Page 7 The halflives for.Zn-65 and Sb-125 are corrected and ( depending on their halflives) new corresponding values for the Effective Dilution Flow and the Individual Dilution Factor are introduced for these two nuclides.
- 5. Page 9 For Zn-65 and Sb-125 the site related Ingestion Dose Commitment Factors are changed based on the corrected halflives for these two nuclides. _
- b. Determination that the change does not reduce the accuracy or reliability .
I of dose calculation or set point determinations ( T.S. 6.15.2.1.b)
Changes described under a.1 and a.2 above have no influence on accuracy or reliability of dose calculation or setpoint determinations. Changes described under a.3, a.4 and a.5 above do increase the accuracy or reliability of dose calculation in that some site Related Ingestion Dose Commitment Factors have been corrected. However the extent of the changes
, are minor because only four of 42 considered nuclides are involved which are not major dose contributors.
- c. Documentation of the fact that the change has been reviewed and found acceptable by the SNSOC ( T.S. 6.15.2.1.c )
See attached ADM 5-4, Attachment 3 documenting SNSOC approval.
k j,s NO9TH ANNA ADM-5.4 REQUEST TO CHANGE PROCEDURE AND ROUTING FORM Attachment 2 Page 1 of 1 Date 12-09-88 Procedure No: // ^ # D M '/ 1 Unit No: //z 2 Rev. Date: /a-J N 3
Title:
sinr< tw at <<.o n o m n - sa r7~ v - M..;d wimr rw a:s a :
/
Changes Requested: /buJ h u Ji<d m 3 sh s<JA,um 6 n= 6 e w 4 W C s5 i An O M & ss-ar, (*Al ? ,d:4.46 ,Le/ls,5 L% seh/J a.w M J,s'em ;r
/ A < fru 2 '
,u--. aL-. + r., An/% Jn d? A-,, m 4 A "
References:
WD 6 v.
Reason: Jr c.~m- f -Im ~12sLm-.2/~s ,4 n /L, sa n s n 4 '
T
/
Requested By:
.wa .
4[tC4 N. / Af-E 8 Department: AAW /%e 9 Date: /> > 3 # 10 i
Safety ED NSO O Non Safety O Classification Change O 11 i New 0 Revision @ Deletion O Required Distribution Date: 3 3 W 'l Review Record (5.3) O Review Checklist (5.4) H This Section To Be Performed By The Cognizant Supervisor 12 Does This Change The Operagng Methods As Desenbod in The UFSAR? YES O NO @
Does This Change invoNe A Change To Tech. Specs? YES O NO @
Does This Change lmeNe A Possbie Unrevewed Safety Ouestion? YES O NO @
If All"NO' No ' Safety Analysis' Is Required. If Any 'YES*, A ' Safety Analysis' is Required.
(10CFR50.59) Approved Copy To Be Provided To Ucensing Coord. For inclusion in Annual Report.
ORDER TITLE INIT. DATE ORDER TITLE INIT. DATE 13 Sta. Procedures Cognizant Supv. d4f l2-23-8f r Supt. d jr-2r-88
&A ' 11211 a b b8 Return To:
14 Appr ed SNSOC 15 16 YES NO O Chairman Signature sl v Date ' fN Immediate Selective Control Distribution O 17 dtaglered implementation Date: /[/Nf18-Action Comoisted 19 l Date Initials Typing, First Time Affix Attachment 3 if entire procedure was retyped. J-07S k Proof Reading (1)
Typing Corrections (I)
Proof Reading (II)
Typing Corrections (II)
Proof Reading (111)
Typing Corrections (Ill)
Drafting Station Records, for Processing and Distribution Special Notes / Instructions 20
PROCEDURE ADM-5.4 REVIEW Attachment 3 CHECKLIST Page 1 of 1 12-09-88 (11 PROC. NO: N/'. c orm - 9 CURRENT REV: .u t DATE: /.:' .23 .n. !
PROCEDURE TITLE: c H.s.rt XJo n Cates.tA m ss w.os s - 1ss A s af. ./.y 'lal tssta.rst 4.u L A./1 FOR NEW, REVISED, OR PROGRAMMATIC UPGRADE:
(2 i (CHECK) l l
J V
Human Factors Review Criteria (Ref: Attachment 9) ]
Radiological Work Practices Criteria (Ref: Attachment 10)
General Procedure Review Criteria (Ref: Attachment 11)
FOR CHANGES:
(3 (CIECK)
/ Latest revision of existing procedure used Changes and location of their placement clear Deletions do not remove constitted material /information
' Additions clearly portray equipment, readings, data, etc.
- /1 Setpoints and/or acceptance criteria changed
- 43 Calculational basis provided or updated per ADM-17.15 U
Review Completed By: /r [ < hen Date: /./- 4 3 W #
Depart.nent : /d //4 [/w ,1
./
l l
=
_ - ____-m_ _ _ _ _ -
o-
\ H.P.-0DCM-4 P:ga 1 of 9
\ 10-23-86 NORTH ANNA POWER STATION OFFSITE DOSE CALCULATION MANUAL SECTION 4 LIQUID EFFLUENT DOSE LIMIT Part Subject Page 1
Technical pecificatio Requirement 2 2 Calculation 2 3 Example 3
~
4 Quarterly Com osite Analyses 4
{
m .. .. .
\
\. H.P.-0DCM-4
\ Page 2 of 9
> 10-23-86
- 1. TECHNICAL SPECIFICATION REQUIREMENT Technical Specification 3.11.1.2 requires that:
"The dose fr dose commitment to the maximum exposed MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:
- a. During any calendar quarter to less than or equal to .5 mrem to the (otal body and to less than or equal to 5 mrem o the critical ot an, and
- b. Duri any calendar year to less than or eq l to 3 mrem to the total ody and to less than or equal to 10 mrem to the critical organ".
S
- 2. ' CALCULATION 2.1 Dose contribu on shall be cale lated for all radionuclides identified in l(iquid effluents tel ased to UNRESTRICTED AREAS based on the following e ressions:
D=tFIf C f g i
Where:
D= the cumulative dpse co itment to the total body or critical organ, from the dquid ef uents for the time period t, in mrem; t= t over which C1 and F are averaged fo,he length r all liq idof releases, the time peri hours F= the near ield average dilution f ctor for C during g any liquid efflue release'. Defined as he ratio of the average
, undil d liquid waste flow during i lease to the average flow from he site discharge structure to STRICTED AREAS; i f g = t individual dilution factor to a ount for increases in c ncentration of longlived nuclides e to recirculation, isted in Table 4.0.
C = the average concentration of radionuclides, i, in undiluted liquid effluent during time period, t, .om any liquid releases, in pCi/ml; A the site related ingestion dose commitment factor t the total g = body or critical organ of an adult for each principal gamma and beta emmitter listed in Table w,.0, in 4entified mrem-ml per br-pCi; Ag = 1.14E+05 (730/D y + 21BFf /D,) DFg where:
1.14E+05 = 1E+06 pCi/pCi x IE+03 ml/kg + 8760 hr/yr, units conversion factor;
---vN.h-'
l %.
~\
l k.' \ .- H.P.-0DCM-4 h :
Paga 3 of 9 L h 10-23-86 1~
l' l 730 = adult water consumption,.kg/yr from NUREG-0133;
,'s .
l
' \,\C D" = dilution factor '. from .. theinear- field area within one quarter mile of the. release point to the potable water.'inta for the adult water consumption. Dw includes the' dilution c ntributions h from the North Anna ' Das to Doswell (0.73), ' th Waste Heat
\- Treatment Facility (C /C ), and Lake Anna ( / ) The potable.
h- water mixing ratio il halculated as 1/(C C /(C x .73) where C /C a are the resp tihe/C,) x .73 =
d' concentrations l, , f$rtEeconsiderednuElike.ndCin.theDischargeannel, Waste Heat-Treatment Facility (Lagoon) and the Res oir. Calculation.
is per Expressions 11.2 - 5, 11.2 - 6, and ' 11.2 - 8 UFSAR. ,
BF g= the bioaccumulation factor for nucJ de,- 1, in fish, pCi/kg p' r pCi/1, from Table A-1 Regulato Guide 1.109; D
- = dilu ion factor ' for the fish athway, calculated as C /C i where and C are the conc ntrations for the considkre8 nuclidek n the# Discharge Cha el and the Waste Heat Treat-ment F,aci y (Lagoon).- Cal lation is per Expressions 11.2 - 5, .2 - 6 UFSAR. !
DFg = the critical gan dose conversion factor for nuclide, i, for adults, in are pCi, from. Table E-11 of Regulatory Guide 1.109, Rev. 1.
- 3. EXAMPLE 3.1 Compilation of dat from rel ase records for a 31 day period provides the follo ing informatio :
Total Volume o Undiluted Liquid ' fluent Released = 2.00E+10 ml Total Volum of Dilution Water Used ring Period = 1.59E+14 ml Average -!
ncentration of Radionr.clides in ndiluted Liquid Effluent C -134 = 6.23E-08 pCi/ml s-137 = 2.13E-07 pCi/ml I-131 = 5.17E-07 pCi/ml Co-58 = 1.53E-07 pCi/ml Co-60 = 7.27E-07 pCi/ml H-3 = 4.62E-03 pCi/ml 3 31 Day Total Body Calculation:
D=tFIf C fAf g i
Obtain total body Agvalues from Table 4.0 t = 744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br /> '
H.P.-0DCM-4 Pega 4 of 9 10-23-86
.C A aree-al arem Nuclide i (pCi/ml) -X f g X i pCi-br = hr Cs-134 6.23E-08 X 12.4 X. 5.82E+5 = 4.50 01 Cs-137 2.13E-07 X 14.0 X 3.45E+5 = 1.4 +00 I-131 5.17E-07 X 1.2 'X 1.10E+02 = 6 2E-05 Co-58 1.53E-07 X 3.3 X 2.57E+02 = .30E-04 Co-60 7.27E-07 X 16.0 X 8.32E+02 = 9.68E-03
-3 4.62E-03 X 18.0 X- 6.25E+00 = 5.20E-01 IC f g Ag= 2.38E+00 i
F= 2' 0E+10 (ml)'
1.59 +14 (al) 2.
Therefore,' D = 744(hr) x ,5 )
x 2.38E+00 ""*r D 2.23E-01 mrem to otal Body.
3.3 31 Day Critical Orga Calculatio :
D=tFIf C gA g g 1
Obtain Critical Organ g value from Table 4.0.
t = 744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br />.
rem-al mrem A
Nuclide i Ci/c1) X fg X i bCi-hr = hr Cs-134' 6 23E-08 X 12.4 X 7.11E 5 = 5.49E-01 Cs-137 .13E-07 X 19.0 X 5.27E+0 = 2.11E+00 I-131 5.17E-07 X 1.2 X 1.91E+02 = 1.19E-04 Co-58 1.53E-07 X 3.3 X 1.14E+02 = 5.76E-05 Co-60 7.27E-07 X 16.0 X 3.37E+02 = 3.92E-03 H-3 4.62E-03 X 18.0 X 6.25E+00 = 5.20E-01 I Cg Ag = 3.18E+00 i
F= 2.00E+10 (ml) 1.59E+14 (ml) 2 Therefore, D = 744(hr) x $5 x3.18E+00"[*r D = 2.98E-01 mrem to critical organ.
T X.: ::
/
H.P.-0DCM-4 Page 5 of 9 10-23-86
- 4. QUARTERLY COMPOSITE ANALYSES For radionuclides not determined in each batch or weekly compo te, the dose contribution to the current monthly or calendar quarter ulative summation may be approximated by assuming an averag monthly concentration based on the previous monthly or quarter p composite a .alyses. However, for reporting purposes, the ca ulated dose coh ribution shall be based on the actual composite analys s.
/
O H.P.-0DCM-4 Page 6 of 9 10-23-86 l
TABLE 4.0 l
SITE RELATED LIQUID PATHWAY INGESTION DOSE FACTORS AND INDIVIDUAL DILUTION FACTORS NAPS UNIT 1 AND 2
\ Individual Total Body A g Critical Organ
- A g Radionuclides yil.Factorfi mres/hr pCi/ml mrem /hr pCi/ml N
H-3 \ 18.0 2.80E+00 6.25E400 2.80E+00 6.25E+00 Na-24 \ 1.0 3.11E+02 4.43t+01 3.11E+02 4.43E+01 C r.-51 \1.9 9.05E-01 Ivi5E+00 - -
Mn-54 &w3 5.71E+02 g.67E+02 2.99E+03 4.54E+03 Fe-55 1316 8.10E+01 /1.30E+02 3.47E+02 5.59E+02 Fe-59 2.$N 7.15E+02/ 9.79E+02 1.86E+03 2.55E+03 Co-58 3.3 \ 1.74E+02 2.57E+02 7.75E+01 1.14E+02 Co-60 16.0 \ 4.91Ef02 8.32E+02 2.23E+02 3.77E+02 Zn-65 1.0 \ 2.21E+04 1.07E+04 4.89E+04 2.37E+04 Rb-86 1.6 \ 3.J2E+04 3.71E+04 6.70E+04 7.86E+04 S r-89f C 7. 2.6 \ V.41E+02 9.10E+02 - -
Sr-90 19.0 X1.35E+05 2.40E+05 - -
Y-91 2.9 / 2w44E-01 3.57E-01 - -
Zr-95 3.1 / 2.01E-01 3.18E-01 2.97E-01 1.81E-01 Zr-97 1.0 / 4.73E-03 1.85E-04 1.03E-02 4.04E-04 Nb-95 1.1 / 8.79Et01 5.50E+01 1.64E+02 1.02E+02 ~
Mo-99 1.0 / 3.37E+01 8.87E+00 1.77E+02 4.66E+01 Ru-103 2.3 / 3.27E+00\ 4.40E+00 - -
Ru-106 9.# 1.43E+01 \ 2. 70E+01 - -
Ag-110m 7' 3 2.54E+00 5.08E+00 4.28E+00 8.56E+00 Sb-124 /2.9 2.98E+01 4.10E+01 1.42E+00 2.24E+00 Sb-125 / 1.3 1.14E+01 5.97E+00 5.37E-01 2.80E-01 Te-125m / 2.9 2.35E+02 3.30Et02
~
6.36E+02 8.92E+02 Te-127m / 4.3 5.41E+02 7.92E+02 1.59E+03 2.32E+03 Te-129m / 2.1 1.19E+03 1.57E+03\ 2.81E+03 3.69E+03 Te-131m / 1.0 4.62E+02 1.32E+02 \ 5.54E+02 1.59E+02 Te-132 / 1.1 1.00E+03 5.72E+02 \ 1.07E+03 6.09E+02 I-131 / 1.2 1.67E+02 1.10E+02 \ 2.91E+02 1.91E+02 I-133/ 1.0 9.29E+00 1.26E-01 \2.66E+01 3.39E-01 I-133 1.0 3.68E+01 4.16E+00 h21E+02 1.36E+01 Iy134 1.0 5.04E+00 2.61E-02 1 41E+01 7.30E-02 1-135 1.0 2.09E+01 7.97E-01 5.67E+01 2.16E+00
/ Cs-134 12.4 3.84E+05 5.82E+05 4.70ET05 7.11E+05
/ Cs-136 1.4 5.87E+04 6.39E+04 8.16E+04 8.88E+04
/ Cs-137 19.0 2.27E+05 3.45E+05 3.46E+05\ 5.27E+05 Cs-138 1.0 1.71E+02 1.12E+00 3.E46+02 \2.25E+00 Ba-140 1.4 4.20E+01 3.07E+01 8.06E-01 $s.89E-01 La-140 1.0 2.15E-02 5.26E-03 8.15E-02 1.'99E-02 Ce-141 2.1 1.93E-02 2.38E-02 1.70E-01 2 .1'OE-01 Ce-143 1.0 3.63E-03 1.76E-04 3.28E+01 1.59Et00 Ce-144 7.9 7.04E+01 1.45E+00 5.48E+00 1.13E+0J Np-239 1.0 2.65E-03 6.10E-4 4.80E-03 1.11E-03\
- Critical Organ is defined in'dP-0DCM-A2, Page 2 of 2. N g
\
-Q---. .:. .. . . . . . - - -
H.P.-0DCM-4 Page 7 of 9 10-23-86 NORTH ANNA 1ARE SPECIFIC DATA See UFSAR CM 11.25 and CDCM 4 for Nomenclature Volumes Reservoir VR = 1.C6E+10 cft Lagoon VL = 2.662+ cft F ow Rates : Reservoir to Lagoon RRL = 414 cfs/sec 3
Lagoon to Reservoir RLR = l Lake over Das ROR =
41 4.5 cfs/sec j Evap, ates : Reservoir RER =
20 cfs/sec Lagoon REL =
59 cfs/sec 21 cfs/sec
- NUCLID HALFLIFE BIOACCUM. EFF.DIL FLOW fi INDIVIDUAL (sec) FACTOR (c ) DIL. FACTOR 1 -H-3~ 3. 9E+08 0.9 230 18.0 2 Na-24 5.4 E+04 100.0 4127 1.0 3 Cr-51 2.39 +06 200.0 2194 4 Mn-54 2.70E 7 1.9 i 400.0 500 8.3 5 Fe-55 8.52E+ 100.0 305 13.6 6 Fe-59 3.86E+06 100.0 1687 2.5 7 Co-58 6.12E+06 50. 1267 3.3 8 Co-60 1.66E+08 50 259 16.0 9 Zn-65 2.11E+05 200 .0 3999 1.0 10 Rb-86 1.61E+06 0.0 2634 1.6 11 Sr-89 4.37E+06 0.0 1566 2.6
-12 Sr-90 9.02E+08 3'9. 0 218 19.0 13 Y-91 5.06E+06 25\.0 1430 2.9 14 Zr-95 5.53E+06 3.3 1351 3.1 15 -Zr-97 6.08E+04 3.3 4124- 1.0 16 Nb-95 3.12E+0 30000.0 3887 1.1 17 Mo-99 2.38E+ 10.0 3971 1.0 18 Ru-103 3.40E 6 10.0 1816 2.3 19 Ru-106 3.1 +07 10.0 458 9.0 20 Ag-110m 2. E+07 2.3 567 7.3 21 Sb-124 5 OE+06 1.0 405 2.9 22 Sb-125 .34E+05 1.0 3 91 1.3 23 Te-125m 5.01E+06 400.0 14h4 2.9 24 Te-127 9.42E+06 400.0 95d 4.3 25 Te-12 2.90E+06 400.0 1983 2.1 26 Te- la 1.08E+05 400.0 4094 1.0 27 Te 32 2.82E+05 400.0 3922 1.1 38 I 31 6.95E+05 15.0 3440 1.2 29 132 8.28E+03 15.0 4140 1.0 30 I-133 7.49E+04 15.0 4116 1.0 3 I-134 3.16E+03 15.0 4140 1.0 I-135 2.38E+04 15.0 4137 .0 33 Cs-134 6.50E+07 2000.0 334 1 4 34 Cs-136 1.14E+06 2000.0 3002 1.
35 Cs-137 9.51E+08 2000.0 218 19.0 36 Cs-138 1.93E+03 2000.0 4140 1.0 37 Ba-140 1.11E+06 4.0 3030 1.4 38 La-140 1.45E+05 25.0 4064 1.0 39 Co-141 2.81E+06 1.0 2019 2.1 40 Co-143 1.19E+05 1.0 4085 1.0 41 Co-144 2.46E+07 1.0 527 7.9 42 'Np-239 2.04E+05 10.0 4007 1.0
o H.P.-0DCM-4 Page 8 of 9 10-23-86 INGESTION DOSE FACTORS FOR ADULTS (mren/pci ingested) 4
- Nuclide BONE LIVER W. BODY THYROID KIDNE LUNG GI-LLI 1 H-3 -
g- -
1.05E-07 1.05E.07 1.05E-07 1.05 -07 1.05E-07 1.05E-O' 2 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.7 E-06 1.70E-06 1.70E-0(
Cr-51
- - -\- 2.66E-09 1.59E-09 5 6E-10 3.53E-09.6.69E-01 3 --
4 Mn-34 ---
s 4 .57E-06 8.72E-07 - - - - .36E - - - 1.40E-0!
5 Fe-55 2.75E-0 1.90E-06 4.43E-07 - - - - ---- 1.06E-06 1.09E-0(
6 Fe-59 4.34E-06 .02E-05 3.91E-06 - - - ----
7 Co-58 ----
7 45E-07 1.67E-06 - - - - ---- ----
1.51E-0!
8 Co-60 ---- 2. 4E-06 4.72E-06 - - - --- ----
4.02E-0!
9 Zn-65 4.84E-06 1.5 E-05 6.96E-06 - - - - 1.03E - - - 9. 7 0E-0<
10 Rb-86 ----
2.111 05 9.83E - - - ---- - - --
4.16E-0<
11 Sr-89 3.08E-04 - - - 8.84E-06 --- ---- - - --
4.94E-0 12 Sr-90 7.58E-09 - - - - 1.86E - - - ---- - - --
2.19E-0<
13 Y-91 1.41E-07 - - - - .77E- ---- ---- - - -- 7. 7 6E-0; 14 Zr-95 3.04E-08 3.75E-09 60E 09 - - - - 1.53E - - - 3.09E-0:
15 Zr-97 1.68E-09 3.39E-10 1. 5 - - - 5.12E - - - 1.05E-0 16 Nb-95 6.22E-09 3.46E-09 1. K - - - 3.42E - - - 2.10E-0
- 17. Mo-99 ----
4.31E-06 8 OE 07 - - - - 9.76E - - - 9.99E-@
18 Ru-103 1.85E-07 - - - - .97E - - - 7.06E - - - 2.16E-@;
19 Ru-106 2.75E-06 - - - - 3.48E-0 ---- 5.31E.06 - - - - 1.78E-@.
20 Ag-110m 1.60E-07 1.48E- 8.79E - - - 2.91E - - - 6.04E-@
21 Sb-124 2.80E-06 5.29E 8 1.11E-06 g79E - - - 2.18E-06 7.95E-@
Sb-125 22 1.79E-06 2.00 -08 4.26E-07 1.32E - - - 1.38E-06 1.97E-0 23 Te-125m 2.68E-06 9.7 E-07 3.59E-07 8.0' -07 1.09E - - - 1.07E-@'
24 Te-127m 6.77E-0.6 2 2E-06 8.25E-07 1.73 06 2.75E - - - 2.27E-@
25 Te-129m 1.15E-05 .29E-06 1.92E-06 3.95E- 6 4.80E - - - 5.79E-@
26 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-0 8.57E - - - 8.40E-0 27 Te-132 2.52E- 1.63E-06 1.53E-06 1.80E-a6_ .57E - - - 7.71E-0. .
28 I-131 4.16E 6 5.95E-06 3.41E-06 1.95E-03 1 02E - - - 1.57E-0 29 I-132 2.03 -07 5.43E-07 1.90E-07 1.90E-05 8. SE - - - 1.02E-O' 30 I-133 1.4, E-06 2.47E-06 7.53E-07 3.63E-04 4.3 - - - 2.22E-O' 31 I-134 1p6E-072.88E-071.03E-074.99E-064.58 07 - - - - 2.51E-li 32 I-135 .43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E - - - 1. 31E-0 <
33 Cs-134 6.22E-05 1.48E-04 1.21E - - - 4.79E-0 1.59E-05 2.59E-0(
34 Cs-136 6.51E-06 2.57E-05 1.85E - - - 1.43E-05 . 9 6E-06 2. 9 2E-0(
35 Cs-137 7.97E-05 1.09E-04 7.14E - - - 3.70E-05 23E-05 2.11E-0(
36 Cs- 5.52E-08 1.09E-07 5.40E - - - 8.01E-08 7. 1E-09 4.65E-1:
37 Ba 0 2.03E-05 2.55E-08 1.33E - - - 8.67E-09 1.4 -08 4.18E-0!
38 140 2.50E-09 1.26E-09 3.33E - - - ---- -- -
' 9.25E-0!
39 -141 9.36E-09 6.33E-09 7.18E - - - 2.94E - - 2.42E-0!
40 . e-143 1.65E-09 1.22E-06 1.35E - - - 5.37E - - - 4.56E-0{
41 Ce-144 4.88E-07 2.04E-07 2.62E - - - 1.21E - - - . 65E-0<
4 Np-239 1.19E-09 1.17E-10 6.45E - - - 3.65E - - - 40E-0{
~ - - - - -
H.P.-0DCM-4 Pa8e 9 of 9 10-23-86 SITE REIATED INGESTION DOSE COMMITMENT FACTORS FOR ADUL S (aram-al/uci-hr)
- Nuclide BONE LIVER W. BODY THYROID - KIDNEY LUNG GI-LLI 1 H-3 -- --
6.25E+00 6.25E+00 6.25E+00 6.25E 0 6.25E+00 6.25E+00 2 Na-24 4.4 E+01 4.43E+01 4.43E+01 4.43E+01 4.43 +01 4.43E+01 4.43E+01 3 Cr-51 -- -
-.- 1.15E+00 6.87E-01 2.5 E-01 1.52E+00 2.89E+02 4 Mn-54 ----
4.54E+03 8.67E+02 - - - - 1. 5E+03 - - - - 1.39E+04 5 Fe-55 8.09E+ 2 5.59E+02 *.30E+02 - - - - ---
3.12E+02 3.21E+02 6 Fe-59 1.09E+03 2.55E+03 9.79E+02 - - - - ----
7.14E+02 8.51E+03 7 Co-58 ----
.14E+02 2.57E+02'- - - - ---- ----
2.32E+03 8 Co-60' ----
- 3. 7E+02 8.32E+02 - - - ---- ----
7.08E+03 9 Zn-65 7.46E+03 2.3 E+04 1.07E+04 - - -
1.59E+04 - - - - 1.50E+04 10 Rb-86 ----
7.96 +04 3.71E+04 - - - - ---- ----
1.57E+04 11 Sr-89 3.17E+04 - - - 9.10E+02 - -- ---- ----
5.08E+03 12 Sr-90 9. 8 0E - - - 2.40E+05 --- ---- ---- 2.83E+04 13 Y-91 1.34E+01 - - - - 3.57E - - - ---- ----
7.35E+03 14 Zr-95 1.47E+00'1.81E-01 ----
7.38E-01 - - - - 1.49E+03 15 96 Zr-97 Nb-95 2.00E-03 4.04E-04 1k.18E-0 SE-A4 - - - -
1.84E+02 1.02E+02 5.5 E+01 - - - -
6.10E-04 - - - - 1.25E+02 1.01E+02 - - - - 6.21E+05 7 Mo-99 ----
4.66E+01 8.8 00 - - - - 1.06E+02 - - - - 1.08E+02 18'Ru-103 l'.02E+01 - - - - 4. OE O---- 3.90E+01 ; - - 1.19E+03 19 Ru-106 2.13E+02 - - - - .70E+ ----
4.12E+02 - - - - 1.38E+04 4 20 Ag-110m 9.25E+0G 8.56E+00 .08E+00 - --
1.68E+01 - - - - 3.49E+03 '
21 Sb-124 1.19E+02 2.24E+0 4.70E+01 -
9.23E+01 3.37E+03 22 Sb-125
-23. Te-125m 2.51E+012.80E-15.97E+002k87E-01----
2.46E+03 8.92 5E - - - 1.93E+01 2.76E+02 02 3.30E+02 7.4s_E+02 1.00E+04 - - - - 9.82E+03 24 Te-127m 6.50E+03 2.3S +03 7.92E+02 1.66 +03 2.64E+04 - - - - 2.18E+04 25 Te-129m 9.90E+03 3.g9E+03 1.57E+03 3.40E 3 4.13E+04 - - - - 4.99E+04 26 Te-131m 3.24E+02 4593+02 1. 32E+02 2.51E+0 1.61E+03 - - - - 1.58E+04 21 Te-132 9.42E+02 .09E+02 5 72E+02 6.73E+02 5.87E+03 - - - - 2.88E+04 38 I-131 1.34E+0 1.91E+02 1.10E+02 6.27E+04 8E+02 - - - - 5.04E+01 29 I-132 1.34E- 1 3.59E-01 1.26E-01 1.26E+01 5 1E - - - 6.74E-02 30 I-133 7.845 00 1.36E+01 4.16E+00 2.00E+03 2.3 +01 - - - - 1.23E+01 31 I-134 2.6SE-02 7.30E-02 2.61E-02 1.26E+00 1.16 - - - 6.36E-05 32 I-135 8. $E-01 2.16E+00 7.97E-01 1.42E+02 3.46E+g 0.- - - - 2.44E+00 33 Cs -134 2 99E+05 7.11E+05 5.82E+05 - - - - 2.30E+0 7.64E+04 1.24E+04 34 Cs-136 .25E+04 8.88E+04 6.39E+04 - - - - 4.94E+04 .77E+03 1.01E+04 l 35 Cs-137 3.86E+05 5.27E+05 3.45E+05 - - - - 1.79E+05 -95E+04 1.02E+04 ;
36 Cs-138 1.14E+00 2.25E+00 1.12E+00 - - - - 1.66E+00 1. 4E-01 9.62E-06 l 37 Ba-14 4.69E+02 5.89E-01 3.07E+01 - - - - 2.00E-01 3.3 -01 9.66E+02 38 La-4 0 3.95E-02 1.99E-02 5.26E-03 - - - - ---- -- -
1.46E+03 39 Co,141 3.11E-01 2.10E-01 2.38E-02 - - - - 9.76E-02 - - - 8.03E+02 40 Cp143 2.15E-03 1.59E+00 1.76E-04 - - - - 7.01E-04 - - - - 5.95E+01
.41 .e-144 2.70E+01 1.13E+01 1.45E+00 - - - - 6.70E+00 - - - - Q.14E+03 2 NP-239 1.13E-02 1.11E-03 6.10E-04 - - - - 3.45E-03 - - - - 27E+02
/
/
[ es; ,
- . , 'H.P.-0DCM-4' Page 1 of 9e i' ,
12-29-88 w
NORTH' ANNA POWER STATION' 4
0FFSITE DOSE CALCULATION MANUAL'
.SECTION 4-LIQUID EFFLUENT DOSE' LIMITS-Part Subject Pag -
1 Technical Specification Requirement 2:
'2. Calculation 2 3 Example 3 4- Quarterly Composite Analyses '5 a-__________.______m
ii.P.-0DCM-4 Page 2 of 9 12-29-88
- 1. E T_EC'INICAL SPECIFICATION REQUIREMENT Technic 3'. Specification 3.11.1.2 requires that: "The dose or dose commitment to the maximum exposed MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) 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 the critical 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 the critical organ".
- 2. CALCULATION 2.1 Dose contribution shall be calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS based on the following expressions:
D=tFIf f C.I A.1 Where:
D= the cumulative dose commitment to the total body or critical organ, from the liquid effluents for the time period t, in mrem; t= the length of the time period over which C and F are averaged for t.El liquid releases, bours; F= the near field average dilution factor for C. during any liquid effluent release. Defined as the ratio of the average undiluted liquid waste flow during release to the average flow from the site discharge structure to UNRESTRICTED AREAS;
- f. = the individual dilution factor to account for increases in concentration of longlived nuclides due to recirculation, listed in Table 4.0.
C. = the average concentration of radionuclides, i, in undiluted liquid effluent during time period, 't, from any liquid releases, in pCi/ml; A.* = the site related ingestion dose commitment factor to the total l body or critical organ or an adult for each identified principal gamma and beta emmitter listed in Table 4.0, in mrem-ml per br-pCi; A.I = 1.14E+05 (730/Dw + 21BF./D 1 a
) DF.1 where:
1.14E+05 = 1E+06 pCi/pCi x IE+03 ml/kg + 8760 hr/yr, units conversion factor;
_ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - . i
y' w, y* , .
H.P.-0DCM-4 Page 3 of 9 3: 12-29-88~
L 730 = adult water consumption, kg/yr from NUREG-0133; D" = dilution factor from the near field area within ~ one-quarter mile of the release point to the potable water intake for the adult water consumption. Dw includes the dilution contributions
'from the North Anna Dam to Doswell (0.73), the Waste Heat; Treatment Facility (C c/Cg), and Lake Anna-(C /C ). 'The potable-c L Ctconcentrations
/CR ) x .73 =
water ' mixing ratio is calculated as 1/(C CR /(C x .73) where C /C and C are'the respective fortfieconsiderednuclibeinthe.DischargeChannel,WasteHeat Treatment' Facility (Lagoon) and the Reservoir. Calculation is per Expressions 11.2 - 5, 11. 2 - . 6, and 11.2 - 8 UFSAR.
BF. =
- the bioaccumulation - factor for nuclide, i, in fish,. pCi/kg per pci/1, from Table A-1 Regulatory Guide 1.109; D = dilution factor for .the fish ' pathway,. calculated as C g/C-where - C .and C are the concentrations - for the considere8 nuclide in the# Discharge Channel and the Waste - Heat Treat-ment Facility (Lagoon). . Calculation is per Expressions-11'.2 .5, 11.2 - 6 UFSAR.
. DF'g - = the critical organ dose conversion factor for nuclide, ~i, for adults, in mrem /pci, from Table E-11 of Regulatory Guide 1.109, Rev. 1.
- 3. EXAMPLE 3.1 Compilation. of data from release- records for a 31 day period-provides the following information:
Total - Volume . of Undiluted Liquid Effluent Released = 2.00E+10 ml Total Volume of Dilution Water Used During Period = 1.59E+14 ml Average Concentration of Radionuclides in Undiluted Liquid Effluent Cs-134 = 6.23E-08 pCi/ml Cs-137 = 2.13E-07 pCi/ml l I-131 = 5.17E-07 pCi/ml Co-58 = 1.53E-07 pCi/ml Co-60 = 7.27E-07 pCi/ml H-3 = 4.62E-03 pCi/ml 3.2 31 Day Total Body Calculation:
D = t F I f. C. A.
1 1 1 Obtain total body A values from Table 4.0 f
t = 744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br />
~'
t . .
H.P.-0DCM-4 p Page 4 of 9 12-29-88 mrem-m1 amrem Nuclide- Ci -(pci/ml) 'X f g XAi Ci-hr = hr
~Cs-134 6.23E-08 'X 12.4 X 5.82E+5 = 4.50E-01 L. Cs-137 2.13E-07 X' 19.0 X 3.45E+5. =- 1.40E+00 l I-131 .5.17E-07 X 1.2 X- 1.10E+02- = 6.82E-05 Co-58 1.53E X 3.3 X 2.57E+02 =- 1.30E-04 Co-60 7.27E-07 X 16.0 X 8.32E+02 = 9.68E-03 H-3. 4.62E-03 X 18.0 X 6.25E+00- = 5.20E-01 Cf .' fg A g'= .2.38E+00' p _, 2.00E+10 (ml) 1.59E+14 (ml) 0E Therefore,' D =.744(hr) x 9 x2.38E+00"[**
D =.2.'23E-01 mrem to Total Body.
3.3 31 Day Critical Organ Calculation:
D=tFIf C fA f g 1
Obtain Critical. Organ A. values from Table 4.0.
t = 744 hours0.00861 days <br />0.207 hours <br />0.00123 weeks <br />2.83092e-4 months <br />.
mrem-al mrem C A Nuclide i (pCi/ml) X f g
X i Ci-hr = hr Cs-134 6.23E-08 X 12.4 X 7.11E+05 = 5.49E-01 Cs-137 2.13E-07 X 19.0 X 5.27E+05 = 2.13E+00 I-131 5.17E-07 X 1.2 X 1.91E+02 = 1.18E-04 l
Co-58 1.53E-07 .X 3.3 X 1.14E+02 .= 5.76E-05 Co-60 7.27E-07 X 16.0 X 3.77E+02 = 4.39E-03 l H-3 4.62E-03 X 18.0 X 6.25E+00 = 5.20E-01 I C. A.- = 3.20E+00 1 1 l F= 2.00E+10 (ml) l 1.59E+14 (ml)
~
Therefore, D = 744(hr) x 59E x3.20E+00"[**
D = 2.99E-01 mrem to critical organ.
l l
H.P.-0DCM-4 Page 5 of 9 12-29-88
- 4. QUARTERLY COMPOSITE ANALYSES For radionuclides not detennined 'in each batch or weekly composite, the dose contribution to the current monthly or calendar quarter cumulative summation may be approximated by assuming an average monthly concentration based on the previous monthly or quarterly composite analyses. However, for reporting purposes, the calculated dose contribution shall be based on the actual composite analyses.
l
d
- :i: H.P.-0DCM-4' L Page 6.of 9 29-88 l
TABLE 4.0 SITE RELATED LIQUID PATHWAY INGESTION DOSE FACTORS AND INDIVIDUAL DILUTION FACTORS ,
NAPS UNIT 1 AND 2 l Individual Total Body A f Critical Organ * : A, Radionuclides Dilution mrem /br mrem /hr Factor (fi) pCi/ml pCi/ml H-3 18.0 6.25E+00 6.25E+00 Na-24 1.0- 4.43E+01 4.43E+01 Cr-51 1.9 1.15E+00 -
Mn-54 8.3 8.67E+02 4.54E+03-Fe-55 13.6 1.30E+02 5.59E+02 Fe-59 2.5 9.79E+02 2.55E+03 Co-58 3.3 2.57E+02 1.14E+02 Co-60 16.0 8.32E+02 3.77E+02 Zn-65 '.2 7 3.30E+04 7.29E+04 Rb-86 ~1.6 3.71E+04 7.96E+04 Sr-89 2. 6 - 9.10E+02- -
Sr-90 19.0 2.40E+05 -
Y-91 2.9 3.57E-01 -
Zr-95 3.1 3.18E-01 1.81E-01 Zr-97 1.0 1.85E-04 4.04E-04 Nb-95 1.1 5.50E+01- 1.02E+02-Mo-99 1.0 8.87E+00 4.66E+01 Ru-103 2.3 4.40E+00 -
-Ru-106 9.0 2.70E+01 -
Ag-110m 7.3 5.08E+00 8.56E+00 Sb-124 2.9 4.70E+01 2.24E+00-Sb-125 13.7 2.50E+01 1.17E+00 Te-125m 2.9 3.30E+02 8.92E+02 Te-127m. 4.3 7.92E+02 2.32E+03-Te-129m 2.1 1.57E+03 3.69E+03 Te-131m 1.0 1.32E+02 1.59E+02 Te-132 1.1 5.72E+02 6.09E+02-I-131 1.2 1.10E+02 1.91E+02 I-132 1.0 1.26E-01 3.59E-01 1-133 1.0 4.16E+00 1.36E+01 1-134 1.0 2.61E-02 7.30E-02 I-135 1.0 7.97E-01 2.16E+00:
Cs-134 12.4- 5.82E+05 7.11E+05 Cs-136 1.4 6.39E+04 8.88E+04 Cs-137 -19.0 3.45E+05 5.27E+05 Cs-138 1.0 1.12E+00 2.25E+00 Ba-140 1.4 3.07E+01 5.3ys-01 La-140 1.0 5.26E-03 1. _ 02 Ce-141 2.1 2.38E-02 2.)JE- 1 Ce-143 1.0 1.76E-04 1.5/i+00 Ce-144 7.9 1.45E+00 1.13E+01 Np-239 1.0 6.10E-04 1.11E-03
- Critical Organ is defined in HP-0DCM-A2, Page 2 of 2.
A H.P.-0DCM-4 Page 7 of 9 12-29-88
- ATTACHMENT 1.0
= NORTH ANNA LAKE SPECIFIC DATA L
See'UFSAR CH 11.25 and ODCM'4 for Nomenclature 1
Volumes : Reservoir VR = 1.06E+10 cft Lagoon VL = 2.66E+09 cft Flow Rates : RRL =
Reservoir to Lagoon-Lagoon to Reservoir 4140 cfs/sec RLR = 4144.5 cfs/sec Lake over Dam ROR = 220 cfs/sec Evap. Rates : Reservoir Lagoon RER = 59 cfs/sec REL'= 21 cfs/sec
- NUCLIDE . HALFLIFE BIOACCUM. EFF.DIL. FLOW fi INDIVIDUAL (sec) FACTOR (cfs) DIL. FACTOR 1 H-3 -3.89E+08 0.9 230 18.0 2 Na-24 5.40E+04 100.0 4127 1.0 3 Cr-51 2.39E+06 200.0 2194 1.9 4 Mn-54 2.70E+07 400.0 500 8.3 ,
5 Fe-55 ~8.52E+07 ~ 100.0 305 13.6 l 6- Fe-59 3.86E+06. 100.0 1687- 2.5 7 Co-58 6.12E+06 50.0 1267 3.3 m 8 Co-60 1.66E+08 50.0 259 16.0 9 Zn-65' 2.11E+07 2000.0 575 7.2 l 10 Rb-86 1.61E+06 2000.0 2634 1.6
.. 11 -Sr-89 4.37E+06 30.0 1566 2.6.
12 Sr-90 9.02E+08 30.0 218 19.0 13 Y-91 5.06E+06 25.0 1430 2.9 14 Zr-95 5.53E+06 3.3 1351 3.1 l
15 Zr-97 6.08E+04 3.3 4124 1.0 16 Nb-95 3.12E+05 30000.0 3887 1.1 17 Mo-99 2.38E+05 10.0 3971 1.0 18 Ru-103- 3.40E+06 10.0 1816 2.3 19 Ru-106 3.18E+07 10.0 458 9.0 20 .Ag-110m 2.16E+07 2.3 567 7.3 21 Sb-124 5.20E+06 1.0 1405 2.9 22 Sb-125 8.74E+07 1.0 303 13.7 l
'23 Te-125m 5.01E+06 400.0 1438 2.9 24 Te-127m 9.42E+06 400.0 956 4.3 25 Te-129m 2.90E+06 400.0 1983 2.1 26 Te-131m 1.08E+05 400.0 4094 1.0 27 Te-132 2.82E+05. 400.0 3922 1.1
'28 I-131 6.95E+05 15.0 3440 1.2 29 I-132 8.28E+03 15.0 4140 1.0 30 I-133 7.49E+04 15.0 4116 1.0 31 I-134 3.16E+03 15.0 4140 1.0 32 I-135 2.38E+04 15.0 4137 1.0 33 Cs-134 6.50E+07' 2000.0 334 12.4 34 Cs-136 1.14E+06 2000.0 3002 1.4 35 Cs-137 9.51E+08 2000.0 218 19.0 36- Cs-138 1.93E+03 2000.0 4140 1.0 37 Ba-140 1.11E+06 4.0 3030 1.4 38- La-140 1.45E+05 25.0 4064 1.0 39 Ce-141 2.81E+06 1.0 2019 2.1 40 Ce-143 1.19E+05 1.0 4085 1.0 41 Ce-144 2.46E+07 1.0 527 7.9 42 Np-239 2.04E+05 10.0 4007 1.0 i
S- ,..
H.P.-0DCM-4 Page 8Jof 9~
.12-29-88
(
ATTACHMENT 2.0 INGESTION DOSE. FACTORS FOR ADULTS (arem/pci ingested)
.# Nuclide BONE LIVER W. BODY THYROID KIDNEY' LUNG GI-LLI 1 H-3' ----
1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 2 Na 1.70E-06.1.70E-06 1.70E-06'1.70E-06.1.70E-06 1.70E-06 1.70E-06 3 Cr-51 ---- ----
2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E 4 Mn-54 ----
4.57E-06 8.72E - - - 1.36E - - - l'.40E-05:
^ '
5 Fe-55 2.75E-06 1.90E-06 4.43E-07 ---- ----
1.06E-06 1.09E-06 6 Fe-59 4.34E-06 1.02E-05 3.91E - - - - --
2.85E-06'3.'40E-05 7 Co-58 ----
7.45E-07 1.67E-06 '- - - - - -:- ----
1.51E-05 8 Co-60 ----
2.14E-06 4.72E - - - ---- -- -
4.02E-05 9- Zn-65 4.84E-06 1.54E-05 6.96E - - - 1.03E - - - 9.70E-06{
10 Rb-86 ----
2.11E-05 9.83E - - - ---- -- -
4.16E-06.
11 Sr-89 3. 08E - -
8.84E - - - ---- ----
4.94E-05 12 Sr-90 7.58E - - - 1.86E - - - ---- ----
2.19E-04~
'13 Y-91 1.41E - - - 3.77E - - -
7.76E-05 14 Zr-95 3.04E-08 3.75E-09 6.60E - - - 1.53E - - -~ 3.09E-05 15 Zr-97 1.68E-09 3.39E-10 1.55E - - - 5.12E - - - 1.05E-04 11 6 ,Nb-95 6.22E-09 3.46E-09 1.86E - - - 3.42E - - - 2.10E-05 17 Mo-99 ----
4.31E-06 8.20E - - - 9.76E - - - 9.99E-06 18' Ru-103 1.85E - - - 7.97E - - - 7.06E-07 '- - - -2.16E 19 Ru-106 2.75E - - - 3.48E - - - 5.31E - - - 1.78E-04 20 Ag-110m 1.60E-07 1.48E-07 8.79E - - - 2.91E - - . 6.04E-05 21 Sb-124 2.80E-06 5.29E-08 1.11E-06 6.79E - - - 2.18E-06.7.95E-05 22 Sb-125 1.79E-06 2.00E-08 4.26E-07 1.82E - - - 1.38E-06 1.97E-05 23 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E - - - 1.07E-05 24 Te-127m. 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E - - - 2'.27E-05 25 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E - - - 5.79E-05 26 Te-131m 1.73E-06~8.46E-07 7.05E-07 1.34E-06 8.57E - - - 8.40E 27 Te-132 2.52E-06 1.63E-06.1.53E-06 1.80E-06 1.57E - - - 7.71E-05.
28 -I-131 4.16E-06.5.95E-06 3.41E-06 1.95E-03 1.02E - - - 1.57E-06:
29- I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E - - - 1.02E-07' 30 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E - - - 2.22E-06 31 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E - - - 2.51E-10 32 I-135 4.43E-07 l'.16E-06 4.28E-07.7.65E-05 1.86E - - - '1.31E-06:
33 Cs-134 6.22E-05 1.48E-04 1.21E - - - 4.79E-05 1.59E-05 2.59E-06 34 Cs-136 6.51E-06 2.57E-05 1.85E - - - 1.43E-05 1.96E-06.2.92E-06 35 Cs-137 7.97E-05 1.09E-04 7.14E - - - 3.70E-05 1.23E-05 2.11E-06 36 Cs-138 5.52E-08 1.09E-07 5.40E - - - 8.01E-08 7.91E-09 4.65E-13 37 Ba-140 2.03E-05 2.55E-08 1.33E - - - 8.67E-09 1.46E-08 4.18E-05 38- La-140 2.50E-09 1.26E-09 3.33E - - - ---- ---- 9.25E-05 39 Ce-141 9.36E-09 6.33E-09 7.18E - - - 2.94E - - - 2.42E-05 40 Ce-143 1.65E-09 1.22E-06 1.35E - - - 5.37E - - - 4.56E-05 !
41 Ce-144 4.88E-07 2.04E-07 2.62E - - - 1.21E - - - 1.65E-04 ,
42 Np-239 1.19E-09 1.17E-10 6.45E - - - 3.65E - - - 2.40E-05 l i
.m...-, ..
Y.
-H.P.-0DCM-4 Page 9 of 9 12-29-88 i."
' ATTACHMENT 3.0
. SITE RELATED ' INGESTION DOSE COMMITMENT FACTORS FOR ADULTS (mrem-al/uCi-hr)
- Nuclide BONE LIVER- W. BODY ' THYROID KIDNEY LUNG GI-LLI 1 H-3 ----
6.25E+00 6.25E+00 6.25E+00 6.25E+00 6.25E+00 6'.25E+00-2 Na-24 4.43E+01 4.43E+01 4.43E+01 4.43E+01 4.43E+01.4.43E+01'4.43E+01' 3 Cr-51 ---- ---- 1.15E+00 6.87E-01 2.53E-01 1.52E+00 2.89E+02 Mn-54 4.54E+03 8.67E+02 - - - -- 1.35E+03 - - - - 1.39E+04 4 ----
5 Fe-55 8.09E+02 5.59E+02 1.30E+02 - - - - ---- 3.12E+02 3.21E+02 6 Fe-59 1.09E+03-2.55E+03 9.79E+02 - - - - ---- 7.14E+02 8.51E+03 m 7 .Co-58 ----
1.14E+02 2.57E+02 - - - - ---- ---- 2.32E+03' 8 lCo-60 ----
-3.77E+02 8.32E+02.- - - - ---- ---- 7.08E+03 9 Zn-65 2.29E+04 7.29E+04~3.30E+04 ---- '4.88E+04 - - - - 4.59E+041l~
'10 Rb-86 ----
7.96E+04 3.71E+04.- - - - ---- ---- 1.57E+04
~ 11 Sr-89 3.17E+04 - - - - 9.10E+02 - - - - ---- ---- 5.08E+03 12 Sr-90 9.80E - - - 2.40E+05 - - - - - - - - ---- 2.83E+04-13 Y-91 1.34E+01 - - 3.57E - - - ---- ---- -7.35E+03 14- Zr-95 1.47E+00 1.81E-01-3.18E - - - 7.38E - - - 1.49E+03 15 Zr-97 2.00E-0314.04E-04 1.85E - - - 6.10E - - - 1.25E+02
. 16 ~Nb-95 1.84E+02 1.02E+02 5.50E+01 - - - - 1.01E+02 - - - - 6.21Ef05 17 Mo-99: -----
4.66E+01 8.87E+09 - - - - 1.06E+02 - - - - 1.08E+01 18 Ru-103 -1.02E+01 - - - - 4.40E+00 - - - - 3.90E+01 - - - - 1.19E+03 19 .Ru-106 2.13E+02 - - - - 2.70E+01 - - - - 4.12E+02 - - - - 1.38E+04 20 Ag-110m 9.25E+00 8.56E+00 5.08E+00 - - - - 1.68E+01 - - -
3.49E+03-21 Sb-124 1.19E+02 2.24E+00 4.70E+01 2.87E - - - 9.23E+01 3.37E+03 22 Sb-125 '1.05E+02 1.17E+00 2.50E+01 1.07E-01.- - - - 8.10E+01 1.16E+03 h 23 Te-125m 2.46E+03 8.92E+02 3.30E+02 7.40E+02 1.00E+04 -'- - - 9.82E+03 24 Te-127m 6.50E+03 2.32E+03 7.92E+02 1.66E+03 2.64E+04 - - - - 2.18E+04l 25 Te-129m- 9.90E+03 3.69E+03 1.57E+03 3.40E+03 4.13E+04 - - - - 4.99E+04.
26 Te-131m 3.24E+02 1.59E+02 1.32E+02 2.51E+02 1.61E+03 - - - - 1.58E+04 27 'Te-132 9.42E+02 6.09E+02 5.72E+02 6.73E+02'5.87E+03 - - - - .2.88E+04 28 I-131 1.34E+02 1.91E+02-1.10E+02 6.27E+04 3.28E+02 - - - - 5.04E+01' 29 I-132 1.34E-01 3.59E-01 1.26E-01 1.26E+01 5.71E - - - 6.74E-02 30 I-133- 7.84E+00 1.36E+01 4.16E+00 2.00E+03 2.38E+01 - - - - 1.23E+01 31- I-134 2.69E-02 7.30E-02 2.61E-02 1.26E+00 1.16E - - - 6.36E -32 I-135 8.25E-01 2.16E+00 7.97E-01 1.42E+02 3.46E+00 - - - - 2.44E+00 33 Cs-134 2.99E+05 7.11E+05 5.82E+05 - - - - 2.30E+05 7.64E+04 1.24E+04 34 Cs-136 2.25E+04 8.88E+04 6.39E+04 - - - - 4.94E+04 6.77E+03 1.01E+04
-35 Cs-137 3.86E+05 5.27E+05 3.45E+05 - - - - 1.79E+05 5.95E+04 1.02E+04 36 Cs-138 1.14E+00 2.25E+00 1.12E+00 - - - - 1.66E+00 1.64E-01 9.62E-06 37 Ba-140 4.69E+02 5.89E-01 3.07E+01 - - - - 2.00E-01 3.37E-01 9.66E+02 38 La-140 3.95E-02 1.99E-02 5.26E - - - ---- ---- 1.46E+03 39 Ce-141 3.11E-01 2.10E-01 2.38E - - - 9.76E - - - 8.03E+02 40 Ce-143 2.15E-03 1.59E+00 1.76E - - - 7.01E - - - 5.95E+01 41- Ce-144 2.70E+01 1.13E+01 1.45E+00 - - - - 6.70E+00 - - - - 9.14E+03 u4 2 Np-239 1.13E-02 1.11E-03 6.10E-04 ---- 3.45E - - - 2.27E+02
Techniczl Justification to Chengs HP-0DCM-9 I IAW T.S.-6.15.2
/ 1 A) It is requested to change the following items of HP-0DCM-9 (see attached SNW pages showing corrections and newly approved and dated pages for details).
- 1. Page 3 of 5: 1 Second sentence of step 3.3 has a typographical misspelled word, vlaues should be values.
- 2. Page 4 of 5: .
The example problem has. several mathematical errors which need to be corrected due to use of incorrect R 7py I-131 value ( see attached page 4 of 5 for ex,act corrections).
- 3. Table 9.0 ( page 5"of 5.):
The R I-131 dose factor was found to be incorrect. It needs to be-changjgVfrom 1.E63+05 to 6.72E+08. This value wes calculated using the formula in HP-0DCM-A3 which uses values taken frow HP-0DCM-Al and Regulatory Guide 1.109.(these three (3) references are provided).
(50 kg/dcy) (3301/yr)
R 7py I-131 = (RI-131) (D/Q7py) =(1.0E+12pci/Ci) (9.98E-7see ) + (5.73E-7sec~)
( 6.0E-3 day /1 ) (1) (1.39E-2 mrem /pCi)
( (0.58)(1.0) +[1.0-(0.58)(1.0)) e (9.98E-7sec-1)(7.78E+6sec},-(9.98E-7sec-1)(1. 73E+5sec) 0.7 kg/m 2 2 kg/m 8 (1.10E-9/m8 )
~
mrem /yr R7py I-131 = 6.72E+8 C1/sec O
s
B. The changes as outlined in A.1 and A.2 do not reduce the accuracy or reliability of dose calculations or setpoint determinations. The value correction for HP-0DCM-9, Table 9.0 is used in the determination of dose assessments for the Process Vent ( mixed mode ) pathwty. The current gaseous releases of I-131 are relatively small and not a large contributor to the overall dose assessment for this pathway.
C. Documentation, of these changes are being reviewed and approved by SNSOC, is provided by ADM-5.4, Attachment 3 which has the dated signature of the SNSOC Chairman.
i f
l
NORTH ANNA ADM-5.4 REQUEST TO CHANGE PROCEDURE AND ROUTING FORM Attachment 2 Page 1 of 1 Date 12-09-88 Procedure No: A^ c De - - 1 1 Unit No: //2 2 Rev. Date: /> -u - Pr 3
Title:
e " ' < n a n e ew <- <
- r ' '" *> - " ' - 50 ^ > % rc o, - < -' u rx r. ~ ,o e. t /sa n ~ a > < s :
a /% r. csrt / o c. , cor<e ,e a re p w e a ,- seu x , m . ,- s Changes Requested: '~ ."" 4 A "% "A - e s u /" J 'm r <4~
~J ':~ /:hA.?
<f 4 a e 4.s -., o ene - n/W 4 ~ 4 -t :
References:
Q c.r.M / <* 9 Reason: %. , .
. A J ~/ ~ ~~ < " ~ A : ' ~ s 2 /n e , A L 2 A .s h i , _ .7 Requested By: 4/ r 4u-8 8 Department: A4+ M /"44 9 Date: a-a 2 # 10 Safety O NSO O Non-Safety O Classification change O r New 0 Revision O Deletion O Required Distribution Date: V 71-k4-Review Record (5.3) O Review Checklist (5.4) O This Section To Be Performed By The Cognizant Supervisor *2 Does This Change The Operating Methods As Desenbed in The UFSAR? YES O No O Does TNs Change inwNe A Change To Tech. Specs? YES O No O Does This Change inwNe A Possbie Unrevewed Safety Question? YES O NO O If A!!'NO' No ' Safety Analysis' is Required. If Any 'YES*, A ' Safety Analysis'is Required.
(10CFR50.59) Appret ed Copy To Be Provided To Ucensing Coord. For inclusion h Annual Report.
ORDER TITLE INIT. DATE ORDER TITLE INIT. DATE 13 Sta. Procedures Cognizant Supv. b IM7-1i Supt. j M //-27'-45
/J M A/ "4147' Return To:
14 Approved SNSOC 15 1-YES I NO O Chairman Signature Date /N9 Immediate Selective Control Distribution O 17 htag;ered Implementadon Date: ////N 1 Action Comoteted G Date initials Typing, First Time Affix Attachment 3 if entire procedure was retyped. O'O7M M Proof Reading (!)
Typing Corrections (I)
Proof Reading (11)
Typing Corrections (II)
Proof Reading (Ill)
Typing Corrections (Ill)
Drafting Station Records. for Processing and Distnbution Special Notes / instructions K
L ,,
., PROCEDURE ADM-5.4 REVIEW Attachment 3 CHECKLIST .Page 1 of 1 12-09-88 (1
PROC. NO: /N- p a cerr - 7 CURRENT REV: Al DATE: //- / "/ -d' p ,si..re m.u e. s ec u rn s <ru~. L '- se a n... ')-
PROCEDURE TITLE: o /un*<e l ah . ' M t' m Y' ##'" ##"*~ M 1,2 s a e , 3,, r.e .7, ,, _
- " '" d>
- s k :r,r
,,.s l
FOR NEW, REVISED, OR PROGRAMMATIC UPGRADE:
(2 '
(CHECK)
Human Factors Review Criteria (Ref: Attachment 9)
- Radiological Work Practices Criteria (Ref: Attachment 10) 7 General Procedure Review Criteria (Ref: Attachment 11)
FOR CHANGES:
(3' (CHECK)
' Latest revision of existing procedure used Changes and location of their placement clear Deletions do not remove committed material /information
- Additions clearly portray equipment, readings, data, etc.
//vA Setpoints and/or acceptance criteria changed
</</ Calculational basis provided or updated per ADM-17.15 W'
Review Completed By: // d Date: /J 2J N Department: Ahe'4 /# h s
i o
e
.---_-__-------_g H.P.-0DCM-9 Page 1.of 5 12-19-85 i
l' NORTH ANNA POWER STATION OFFSITE DOSE CALCULATION MANUAL SECTION 9 Iodi . - 131, Tritium, and Ra onuclides in Particulate Form Gaseous Eff1v6nt Dose Limits Part Subject y P
1 Technica1'S ecifi tion Requirement 2-2 Calculation 2 3 Example 3
. _ _ __ N
\ .._. _ . . _ . . _ _ . . . . _ _ . . . _,
H.P.-0DCM-9 Page 2 of 5 4 12-19-85 j
- 1. TECHNICAL SPECIFICATION REQUIREMENT Technical Specification 3.11.2.3 requires that: "The dose to the maximum exposed MEMBER OF THE PUBLIC from iodine-131, from t tium, and from all radionuclides in particulate. form with half-lives reater than 8 days in gaseous effluents released, from each reactor it, from the site to UNRESTRICTED AREAS (see Figure 5.1-1) shall be imited to the fh lowing:
- a. ring any calendar quarter: Less than or eq l to 7.5 areas to t e critical organ, and
- b. Duri any claendar year: Less than or e ual to 15 areas to the criti 1 organ".
- 2. CALCULATION.
2.1 The dose to t maximum exposed ME5BER OF THE PUBLIC from iodine-131, from tritium, d from all radi nuclides in particulate form with -
half-lives great than 8 days in gaseous effluents from the site to areas at and eyond the TE BOUNDARY shall be determined as follows:
Dr = 3.17E-08 I [Rgy,k +R gy kg y]
p p .
where:
Subscripts = vv, refers to vent releases from the building ventilation at; pv, refers to the vent releases from the process ;
vent; Dr = the dose to the er ical organ of the maximum exposed MEMBER OF THE UBLIC in eres; R R = the dose factor for vent lation vent or process i iV P
vent release due to iod e-131, tritium, and from all radionuclides in p ticulate form with half-lives greater than 8 da , in ares /yr per Curie /sec. Factors are liste in Table 9.0;.
D gyy, DgPy = the release for ventilation ven or process vent of iodine-131, tritius, and fro x all radio-nuclides in particulate form with half-lives greater than 8 days in Curies (per sitch '
3.17-08 = the inverse of the number of seconds in a ear.
2.2 All gaseous releases, not through process vent, are considere ground level and shall be included in the determination of Qgyy. l
\
.q l
, .-.. -- .- . .. . ... ... /
l H.P.-0DCM-9 Page 3 of 5 12-19-85
- 3. EXAMPLE 3.1 Compilation of data from release records for the process v t for a quarter provides the following information:
Radionuclides Activity Released (C esl I-131 7.20E-0 H-3 2.45E- 1 Co-58 1.10 -06 3.2 Comp ation of data from release records f r the ventilation vent I for a uarter provides the following info tion:
, Rad -nuclide Activfty Released (Curies)
I-13 6.48E-03 H-3 2.21E+00 Co-58 9.90E-05 3.3 The dose to the critical org of the maximum exposed MEMBER OF THE PUBLIC is ca.lculated Y om:
Dr = 3.17E-08 I [k kg+R D y)
The appropriate v1 es of R g and R gPy shall be obtained from Table 9.0.
l l
R D (arem-sec Radionucli (ares /yr phEcurie/sec) (bEies) = yr i
/
G1 1.45E+09 x 6.48E-03 = 9.40E+06 I-)3
- 1.73E+03 x 2.21E+00 = 3.82E+03 o-58 -
x 90E-05 = -
IR,D i
g g
= 9.40E+06 D/e to lack of information, according to Regulatory Guide 1 109, Rev. 1, o-58 is not included in these calculations.
H.P.-0DCM-9 Page 4 of 5 r
12-19-85 ipv ipv . (arem-sec)
Radionuclides (ares /yr per Curie /sec X (Curies)
~
yr I-131 6.63E+08 X 7.20E-04 = 4.77E+05 H-3 9.36E+02 X 2.45E+01 -
2.29E+02 Co-58 -
X 1.10E-06 = -
= 4.77E+05' IR D i ipv vv Dr = .17E-08 Y# see [9.40E+06
- ~8'
+ 4.77 05 arem-sec) yr yr Dr = 3.1 -01 (ares) !
/ ,
f 1
k l
_ _mmm __ - _ _ _ _ . _ _ _ _ - - - - -- - - - - -- - - ----s---- - ' " " - - - - - - - - " - - ' ' - - " " - - - ' - ^ - - ' ' ' '
H.P.-0DCM-9 Page 5 of 5 12-19-85 TABLE 9.0 CRITICAL PATHWAY
- DOSE FACTORS FOR NORTH ANNA POWER STAT N UNIT NOS. 1 AND 2
-2 at 3250 Meters N irection
\entilstionVentD/Q=2.4E-09m ocess Vent D/Q = 1.1E-09 m
-2 at 3250 Meters N Firection ivv ipv Radionuclides eres/yr mrem /yr Curie /sec Curie /sec H-3 1.73E+03 9.36E+02 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND Rb-86 ND ND Sr-89 ND ND Sr-90 ND ND .
Y-91 ND !
Zr-95 ND ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-127m 1.97E+05 9.04E+04 Te-129m 2.95E+05 1.35E+05 I-131 1.45E+09 1.E63+05 Cs-1 4 ND ND Cs 136 ND ND s-137 ND ND Ba-140 ND ND Ce-141 ND ND Ce-144 ND ND
- Critical Pathway is defined in HP-0DCM-A3, Page 3 of 5.
ND-No data for dose factor according to R.G. 1.109, Rev.1.
m 'n_..
., 6
.. .. .. US r *
., H.P.-0DCM-9.:
1 s Page 1 of: 5:
W 12-29-88~
4 NORTH ANNA POWER STATION OFFSITE DOSE CALCULATION MANUAL SECTION 9
' Iodine 131, Tritium, and Radionuclides in Particulate Form Gaseous Effluent Dose Limits
' Part~ Subject. Page 1 Technical Specification Requirement- 2 2 Calculation 2 3 Example 3 l
1 l
C_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ - . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . . _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ ________J
i H.P.-0DCM-9 i Page 2 of 5 12-29-88 ;
- 1. TECHNICAL SPECIFICATION REQUIREMENT l i
Technical Specification 3.11.2.3 requires that: "The dose to the maximum exposed MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.3 mrems to the critical organ, and
- b. During any claendar year: Less than or equal to 15 mrems to the critical organ".
- 2. CALCULATION 2.1 The dose to the maximum exposed MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be determined as follows:
Dr = 3.17E-08 I [R ivv Divv + R. 1pv Qipv]
f where:
Subscripts = vv, refers to vent releases from the building ventilation vent; pv, refers to the vent releases from the process vent; Dr = the dose to the critical organ of the maximum exposed MEMBER OF THE PUBLIC in mrem; R iyy, R = the dose factor for ventilation vent or process gpy vent release due to iodine-131, tritium, and from all radionuclides in particulate form with half-lives greater than 8 days, in mrem /yr per Curie /sec. Factors are listed in Table 9.0; Divv,D ipv = the release for ventilation vent or process vent of iodine-131, tritium, and from all radio-nuclides in particulate form with half-lives greater than 8 days in Curies (per site);
3.17-08 = the inverse of the number of seconds in a year.
2.2 All gaseous releases, not through process vent, are considered ground level and shall be included in the determination of Q .
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ __ _____a
K i. .
.H.P.-0DCM-9 {
'Page 3'of 5 i 29-88 l.
H3 3. EXAMPLE 3.1 - Compilation of data from release records for.the process vent for a-lt quarter provides the following information:
' Radionuclides Activity Released (Curies)
I-131 '7.20E-04 l1 H 2.45E-01 Co-58 1.10E-06 i .; , 3.2 Compilation of data from release records for the. ventilation-vent for a quarter.provides the following information:
Radionuclides- Activity Released (Curies)-
'Co-58 9.90E-05 3.3 The dose to .the critical organ of the maximum exposed MEMBER OF THE -
?
PUBLIC is calculated from:
Dr = 3.17E-08 I [R g Qg+R p Qgp y]
gy The appropriate values of Rg .and R gy p shall be obtained front Table 9.0.
R. D (mrem-sec)
Radionuclides (mrem /yrpeECurie/sec) x (bEies) = yr I-131 1.45E+09 x 6.48E-03 = 9.40E+06 H-3 1.73E+03 x 2.21E+00 = 3.82E+03-Co-58 -
x 9.90E-05 = -
I R. Q. = 9.40E+06 IVV IVv Due to lack of information, according to Regulatory Guide 1.109, Rev. 1, Co-58 is not included in these calculations.
i
H.P.-0DCM-9 Page 4 of 5 12-29-88 s
N ipv ipv _
(mrem-sec)
Radionuclides (mrem /yr per Curie /sec X (Curies) yr I-131 6.72E+08 X 7.20E-04 = 4.84E+05 l H-3 9.36E+02 X 2.45Er01 = 2.29E+04 Co-58 -
X 1.10E-06 = -
= 5.07E+05 1R D i ipv ivv Dr = 3.17E-08 EI-sec[9.40E+06 mrem-sec yr + 5.07E+05 "#**~8*']
yr Dr = 3.14E-01 (mrem)
A
H.P.-0DCM-9
[.:. 'Pags 5 of 5 12-29-88 TABLE 9.0
- j.
.3 CRITICAL PATHWAY
- DOSE FACTORS FOR NORTH ANNA POWER _ STATION .-
UNIT NOS. 1 AND 2
-2 Ventilation Vent D/Q'= 2.4E-09 m at 3250 Meters N Direction
-2 Process Vent D/Q = 1.1E-09 m at 3250 Meters N Direction ivv ipv Radionuclides = mrem /yr- mrem /yr Curie /sec Curie /see H-3 1.73E+03 9.36E+02 Mn-54 ND ND Fe-59 ND ND Co-58 ND ND Co-60 ND ND Zn-65 ND ND.
Rb-86 ND ND Sr ND ND-Sr-90 ND ND Y-91' ND ND Zr-95 ND- ND Nb-95 ND ND Ru-103 ND ND Ru-106 ND ND Ag-110m ND ND Te-f27m 1.97E+05 9.04E+04 Te-129m 2.95E+05 1.35E+05 I-131 1.45E+09 6.72E+08 Cs-134 ND ND Cs-136 ND ND Cs-137. ND ND Ba-140 ND ND Ce-141- ND -
ND Ce-144 ND ND
- Critical Pathway is defined in HP-0DCM-A3, Page 3 of 5.
ND-No data for dose factor according to R.G. 1.109, Rev.1.
H.P.-ODCM-Al Page 1 of 3 03-20-86 NORTH ANNA POWER STATION
~
- OITSITE DOSE CALCULATION MANUAL
~
Section Al Meteorological Analysis
. Part~ Subject g 1 Purpose 2' 2 Meteorological Data, Parameters, and Methodology 2 3 Results 3
.- i i
t
,a - - -
to .. -w
H.P.-0DCM-Al Peg 2 2 of 3 03-20-86
- 1. PURPOSE The purpose of the meteorological analysis was to determine the annual average X/Q and D/Q values at critical locations around the station for ventilation vent (ground level) and process vent (mixed mode) releases. .
The annual average X/Q and D/Q values were used in performing a dose pathway analysis to determine the maximum exposed individual at SITE BOUNDARY and MEMBER OF THE PUBLIC. The X/Q and D/Q values resulting in the maximum exposures were incorporated into the dose factors in Tables 7.0, 7.1, 7.2, and 9.0.
- 2. METEOROLOGICAL DATAtPARAMETERS, AND METHODOLOGY
'. Onsite meteorological data for the period January 1, 1981 through
} December 31, 1981 was used in calculations. This data included windspeed, wind direction, and differential temperature for the purpose of determining joint frequency distributions for those releases characterized as ground level (e.g. ventilation vent), and those characterized asjsized mode (i.e. process vent). The portions of release characterized as ground level cere based on AT158.9ft-28.2ft and 28.2 foot wind. data, and the portions characterized as mixed mode were based on AT158.9ft-28.2ft and 158.9ft wind data.
X/Q's and D/Q's were calculated using the NRC computer code "XOQD0Q Pro ,s a for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations", September, 1977. The code is based upon a straight line airflow model implementing the assumptions outlined in Section C (excluding Cla and C1b) of Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors".
- The open terrain adjustment factors were applied to the X/Q values as recommended in Regulatory Guide 1.111. The site region is characterized by gently rolling terrain such that open terrain correction factors are considered appropriate. The ground level ventilation vent-releaseg calculations included a building wake correction based on a 1516m containment minimum cross-sectional area. The effective release height used in mixed mode release calculations was based on a process vent release height of 157.5 ft, and plume rise due to momentum for a vent diameter of 3 in, with plume exit velocity of 100 ft/sec. Ventilation vent, and vent releases other than from the process vent, are considered-ground level as specified in Regulatory Guide 1.111 for release points less than the height of adjacent solid structures. Terrain elevations were obtained from North Anna Power Station Units 1 and 2 Virginia Electric and Power Company Final Safety Analysis Repcet Table 11C.2-8.
X/Q and D/Q values were calculated for the nearest site boundary ,
resident, milk cow, and vegetable garden by sector for process vent and ventilation vent releases at distances specified from North Anna Power Station Annual Environmental Survey Data for 1981. X/Q values were also calculated for the nearest lake shoreline by sector for the process vent and ventilation vent releases.
4
.. . .... __ _ _ , _ , _ . . .l H.P.-ODCM-Al Page 3 of 3 03-20-86 According to the definition for short term in NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear . Power Stations", October, 1978, some gaseous releases may fit this category, l primarily waste gas decay tank releases and containment purges. Mcwever, i these releases are considered long term for dose calculations as part releases 'were both random in time of day and duration as evidenced by reviewing part release reports. Therefore, the use of annual average concentrations is appropriate according to NUREG-0133.
The X/Q and D/Q values calculated from 1981 meteorological data are comparable to the values presented in the North Anna Power Station UFSAR.
- 3. RESULTS .
The X/Q value that resulted in the maximum total body, skin and inhalation exposure for. ventilation vent releases was 9.3E-06 sec/mg at a site boundary location 1416 meters SE sector. 3For process vent releases, the nice boundary X/Q value was 1.2E-06 sec/m at a location 1513 meters S sector.
The shoreline X/Q value that resulted in the agximum inhalation exposure for ventilation vent releases was 1.0E-04 sec/a at a location 241 metegs NNE sector. The shoreline X/Q value for process vent was 3.7E-06 sec/m at a location 241 meters NME sector.
Pathway analysis indicated that the maximum exposure from iodine-131, and from all radionuclides in particulate form with half-lives greater than 8 days was through the grass-cow-milk pathway. The D/Q value from ventiption vent relasses resulting in the maximum exposure was 2.4E-09 per a at a location 3250 metys N sector. For process vent releases, the D/Q value was 1.1E-09 per a at a location 3250 meters N sector. For
)
tritium, the X/Q value from ventilation vent releases refulting in the .!
l maximp exposure for the silk pathway was 7.2E-07 nec/a , and 3.9E-07 sec/m for process vent releases at a location 3250 meters N sector.
.~
l .
- OO Q
1
. .,., .....u. . . . . . . .:w. - .u....- . . . . .
%d a. es
. H.P.-ODCM-A3 Page 1 of 5 03-20-86 NORTH ANNA POWER STATION OFFSITE 1)0SE CALCULATION MANUAL 1:
1 Section A3 Gaseous Pathway Analysis Part
- Subject Pm -
1 Purpose 2 2 Data, Parameters, and Methodology 2 MD M
a k
M
. . . . . . .. . . . . - . n.. .. .
H.P.-0DCM-A3 Page 2 of 5 03-20-86
! 1. PURPOSE-A gaseous effluent pathway. analysis was performed to determine the location th' at would result in the maximum doses due to noble gases for use in demonstrating compliance with Technical Specifications 3.11.2.1.a and 3.11.2.2. The Analysis also included a determination of the critical pathway, location of maximum exposed MEMBER OF THE PUBLIC, and the.
critical organ for the maximum dose due to iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for use in demonstrating compliance with Technical Specifications 3.11.2.1.b and 3.11.2.3.
- 2. , DATA,' PARAMETERS, AND METHODOLOGY Annual average X/Q values were calculated, as described in ODCM-AI, for the nearest SITE BOUNDARY in each directional sector and at other' critical locations beyond SI'J BOUNDARY. The largest X/Q value was 3
determined to be 9.3E-06 sec/m at SITE BOUNDARY for ventilation gent releases at a location 1416 meters SE direction, and 1.2E-06 sec/m at SITE BOUNDARY for process vent releases at a location 1513 meters S detection. The maximus doses total body and skin, and air doses for gamma and beta radiation due to noble gases would be at these SITE BOUNDARY locations. The doses from both release points are summed la ODCM calculations to calculate total maximum dose.
Technical Specification 3.11.2.1.b dose limits apply specifically to the inhalation pathway. Therefore, the locations and X/Q values determined for maximum noble gas doses can be used to determine the maximum dose from iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days for the inhalation pathway.
The NRC computer code GASPAR, " Evaluation of Atmospheric Releases",
Revised 8/19/77, was run using 1979, 1980, and 1981 North Anna Power Station Gaseous Effluent Release Report Data. Doses from iodine-131, tritium, and particulate j for the inhelation pathway were calculated using the 9.3E-06 sec/m SITE BOUNDARY X/Q. Except for the source term data and the X/Q value, computer code default parameters were used. The results for each year indicated that the critical age group was the child and the critical organ was the thyroid for the inhalation pathway.
The gamma and beta dose factors Kivv, Livv, Mivv, and Nivv in HP-0DCM Table 7.0 were obtained by performing a units conversion of the approriate dose facto 3s from Table B-1, Regul tory Guide 1.109, Rev. 1, f
to area /yr per Ci/m or arad/yr per Ci/m , and multiplying by the i ventilation vent SITE BOUDNARY X/Q value of 9.3E-06 sec/m . The same approach was used in calculating the gamma and beta dose factors Kipv, Lipv, Hipv, Nipv in HP-0DCM-Tably 7.1 using the process vent SITE BOUNDARY X/Q value of 1.2E-06 sec/m .
The inhalation pathway dose factors P and P and P in HP-0DCM-Table 7.2werec'alculatedusingthe$11owingeqYation: i iV P
,g P = 0 (BR) nag (X/Q) (mrem /yr per Curie /sec)
L ,
1 o ,? l
_. . . . - --. =.
H.P.-0DCM-A3 Page 3 of 5 1 03-20-86
- I where, i
E' '= A constant of unit coversion, 1E+12 pCi/Ci; BR = the breathing rate of the child age group, 3700 m3 fy,,
from Table E-5, Regulatory Guide 1.109, Rev. 1; DFA g a the thyroid organ inhalation dose factor for child age group for the ith radionuclides, in eres/pci, from Table E-9, Regulatory Guide 1.109, Rev. 1; 3
X/Q = the ventilation vent SITE' BOUNDARY X/Q, 9.3E-06 sec/m , or
, the process vent SITE BOUNDARY X/Q, 1.2E-06 sec/a 3 as
. appropriate.
Technical Specification 3.11.2.3. requires that the dose to the maximus exposed MEMBER OF THE PUBLIC from iodine-131, critius, and from all ,
radionuclides in particulate form with half-lives greater 8 days be less than or equal to the specified limits. Dose calculations were performed for an ~ exposed MEMBER OF THE PUBLIC within SITE B0UNDARY UNRESTRICTED AREAS, and to an exposed MEMBER OF THE PUBLIC beyond SITE BOUNDARY at locations identified in the North Anna Power Station Annual Environmental Survey Data for 1981.
~
It was determined that the MEMBER OF TB PUBLIC within SITE BOUNDARY would be using I4ke Anna for recreational purposes a maximum of 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br /> per year. It is assumed that this MEMBER OF TE PUBLIC would be.
located the entire 2252 hours0.0261 days <br />0.626 hours <br />0.00372 weeks <br />8.56886e-4 months <br /> at the lake shoreline with the largest annual X/Q of 1.0E-04 at a location 241 meters NNE sector. . The NRC '
computer code GASPAR was run to calculate the inhalatica dose to this individual.
- h GASPAR results were corrected for the fractional year the MEMBER OF YdE PUBLIC.would be using the lake.
Using the' NRC computer code GASPAR and annual average X/Q and D/Q values -
obtained as described in HP-0DCM-A1, the MEMBER OF THE PUBLIC- receiving the largest dose beyond SITE BOUNDARY was determined to be located 3250 meters N sector. N critical pathway was the grass-cow-milk, the anximum age group was the infant, and the critical organ the thyroid.
For each year 1979, 1980, and 1931 the dose to the infant fron the grass-cow-milk pathway was greater than the dose to the MEMBER OF THE )
PUBLIC within SITE BOUNDARY. b refore, the maximum exposed MEMBER OF.
- THE PUBLIC was determined to be the infant, exposed through the grass-cow-silk pathway, critical organ thyroid, at a location 3250 meters N sector.
hR i#
and R dose factors, except for tritium, in HP-0DCM-Table 9.0 were calculatedly multiplying the appropriate D/Q value with the following equation: ,
I R av F, (r) (DFL ) }.fgf_s + (1-fpfs)e" ]e i = K, Ag + A, i Y, Y,
i
- - . . . . c ,
)
H.P.-0DCM-A3 Page 4 of 5 03-20-86 where, K' = a constant of unit conversion, 1E+12 pCi/Ci; QF ""' ^the cow's consumption rate, 50, in Kg/ day (wat weight); 4 U,, a the infant milk consumption rate, 330, liters /yr; Y
p
= the agricultural Productivity by unit area of pasture feed grass, 0.7, Kg/m2; a
Y, the agricultural productivity'by unit area of stored feed, 2.0, in Kg/m ;
F-
= the stable element transfer coefficients, from Table E-1, Regulatory Guide l.109, Rev. 1; r = fraction of deposited activity retained os cow's feed. grass, 1.0 for radioiodine, and 0.2 for* particulate;'
DFLg a the thyroid ingestion dose factor for the ith radionucide-for the infant, in area /pci, from Table E-14, Regulatory Guide 1.109, Rev. 1; Ag = the decay constant for the ith radionuclides, in sec"I ,
A, a the decay constant for removal of acyvity on leaf and plant surfaces by weathering, 5.73E-07 sec (corresponding to a 14 day half-life);
tg = the transport time from pasture to cow, to milk, to receptor, 1.73E+05, in'sec; t
h
= the transport time from pasture, to harvest, to cow, to milk, to receptor, 7.78E+06, in see; f
p a fraction of the year that the cow is on pasture, 0.58 (dimensionless), 7 months per year from NUREG-0597; f, = fraction of the cow feed that is pasture grass whole the cow is on pasture, 1.0 (dimensionless).
Parameters used in the above equation were obtained from NUREG-0133 and i Regulatory Guide 1.109, Rev. 1.
Since the concentration of tritium in silk is based on the airborne concentration rather than the deposition, the following equation is used:
b-3 = F,QF ap (b-3) [0.75(0.5/HH x X/Q where, '
K = a constant of unit conversion, 1E+03 ga/kg; j absolute humidity of the atmoshpere, 8.0, gs/m3 ;
H =
H. K-0DCM-A3 Pega $ of 5 03-20-86 0.75 = ,_ the fraction of total feed that is water; 0.5 = the ratio of the specific activity of the feed grass to the atmospheric water;
=
X/Q ~the annual average cogeentration at a location 3250 meters N sector,7.2E-g7sec/m for ventilation vent releases, and 3.9E-07 sec/m for the process vent releases.
Other parameters have been previously defined.
gens 4
9 8
6 m
t I
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1 . . .- . . . . . . .
I Revision 1 ,
October 1977
[,,,g U.S. NUCLEAR RESULATOOY CCMMISSION I,e- Q)ff % ,,,..
REGULATORYGUIDE OFFICE OF, STANDARDS DEVELOPMENT .
REGULATORY GUIDE 1.109~
CALCULATION OF ANNUAL DOSES TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENTS FOR THE PURPOSE OF EVALUATING COM"PL la CFR.PART 50, APPENDIX I l - .
t .
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- o u.s ,. = s.
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The substantial number of changes in this revision has made it impractical to 3
1 indicate the changes with lines in the margin.
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s APPEN0!K E NUMERICAL DATA FOR THE CALCULATION 0F ANNUAL DOSES TO MAN FROM ROUTINE
-ftELEASE5 0F REACTOR EFFLUENTS This appendix contains data for use in the equations presented in the Regulatory Position and in Appendices A, B, C, and 0 of this guide. The neerical values presented in this appendix are those routinely used by the NRC staff. In instances where more appropriate information of a site-specific nature has been developed and documented, that information should be used.
In a ne ber of instances the staff has found it necessary to provide guidance as to the 4 value of a particular parameter in the absence of substantial empirical data. In such instances the staff has exercised judgment and has considered
- values used by others and the sensitivity of the results,to the value assumed. ,
Information is provided below under four broad categories: environmental data, human data, dose factors, and other parameters.
- 1. Envirossental Data Table E-1 provides values for the following stable element transfer coefficients:
- s. Og , for the estimation of produce, leafy vegetable, or pasture grass radioactivity from that in sell (pC1/kg in vegetation por pCf/kg in soil):
- b. F, for the estimation of cow milk activity from that in feed (pC1/s in milk per PC1/
day ingested by the animal): and
- c. Ff for the estimation of meat activity from that in feed (pct /kg in meat per PC1/ day l
ingestedbytheanimal).
j The data are largely derived from Reference 1. The value of the cow milk transfer coefficient for radiotodine is based on the staff's review of the literature (Refs. 2-g). ,
Values of transfer coefficients for goet milk are presented in Table E-2 for a limited -
number of nuclides. For nuclides not itsted in Table E-2 the milk transfer coefficient from j Table E-1 should be used.
Various animal parameter values are presented in' Table E-3 for use in estimating animal product activity levels as functions of the corresponding levels in feed and water supplies, j '2. Musen Data Tables E-4 and E E present usage rate's of various environmental media by average individuals and maximum individuals, respectively, according to age group. " Seafood" is used to indicate intake of aquatic invertebrates such as lobster, crab, class, and oysters. Ingestion of aquatic plant material is not normally assumed. ,
- 3. Oose Factors Dose factors for external irradiation from a uniformly contaminated ground plane are 2
presented in Table E-6 (Refs. 10 and 11). in units of aram/hr per pC1/m . These factors are applicable for surface contamination via deposition of liquid effluents on shoreline sediments or airborne effluents on ground surfaces. Dose factors are provided for the total body and skin only. Doses to other organs are assumed equal to the total body dose.
Dose factors provided'in Tab'le E-6 are derived from a consideration of the dose rate to air .
1 meter above the ground plane and the penetration of the radiation into the body. The total body dose is computed at a penetration depth of 5 ca; the skin dose is computed at a depth of 2
7 ag/cm . These tissue depths are indicated by Reference 12, where it,is suggested that, for 1.10g-36 O _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _
h, 3
, i .
s I -
\
.- e , , ,
l - .
. TA8tt E-1
- t. ,.
. STA8LE ELDIENT TRAM 5FDI 0ATA* -
I I,g F,M F f
g Vee / Soil Milk (d/t) Meat (d/ke)
He* 4.8E 00 1.0E-02 1.2E-02 l, '! 5.5E 00 1.2E-02 3.1E.02
' C" Na 5.2E-02 4.0E-02*" 3.0E-02 P 1.1E 00 2.5E-02 4.6E-02 Cr 2.5E-04 2.2E-03 2.4E-03 Mn 2.9E-02 2.5E-04 8.0E-04
'1
- Fe 6.6E-04 1.2E-03 4.0602 Co 9.4E-03 1.0E-03 1.3E-02
! N1
- 1.9E-02 6.7E-03 5.3E-02 8
Cu 1.2E-01 1.4E-02 8.0E-03 3
In . 4.0601 3.9602 3.0E-02 l* Rb 1.3601 3.0bO2 3.1E-02 Sr 1.7E-02 8.0E-04** 6.0E-04 y 2.6b03 1.0605 4.6E-03
. Zr 1.7E-04 5.0E-06 3.4E-02' Mb 9.4E-03 2.5E-03 2.8601 Mo 1.2E-01 7.5E-03 8.0E-03
. Tc 2.5E-01 2.5602 4.0601 Av 5.0E-02 1.0E-05 4.0E-01 I
Rh 1.M 01 1.0E-02 1.5603 p*
l Ag 1.5E-01 5.0E-02 1.7b02 Te 1.3E 00 1.0E 03 7.7E-02 f I 2.0E-02 6.0E-03e 2.9603.
Cs . 1.0E 1.2E-02"* 4.0603
'- 5.0E-03 4.0E-04* " 3.2E-03
' 8e La 2.5E-03 5.0E-06 2.0E-04 .
Ca 2.5E-03 1.0604*** 1.2E-03 Pr- 2.5E-03 5.0E-06 4.7E-03 Nd 2.4E-03 5.0E-06 3.3E-03 W 1.8E-01 5.0E-04 1.3E-03 Np 2.5E-03 5.0E-05 2.0004tt
! "Osta presented in this table is from Reference 1 'unless otherwise indicated.
- " Meet and milk coefficients are based on specific activity considerations.
- e 8
From Reference 15.
'See text.
tty,,,g,f,,,,,,13, 4
I b
i s-1.109-37
- d = * - * $ . i
- ...- .-- . -. .. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , - - - - = - - -
j? '. .. ,
TASLE E-14 . .
P AG E 1 ('F 3
)
INGESTION 005E FACTORS FOR INFANT (MREM PER PCI INGESTE01 ,
',, NUCLICE SONE.
LIVER T.800Y THYROID RIONEY LUNG GI-LL1
~
'l H 3 NO DATA
] C 14 3.08E-075.06E-06 2.37E-05 5.06E-06 3.08E-07 3.08E-07 3.08E-07 3 08E-07 3.08E -I 5.06E-06 5.06E-06 5.06E-06 5.06E-06 NA 24 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 I
P 32 1.70E-03 1.00E-04 6. 59E-05 NO DATA NO DATA NO DATA CR 51 NO DATA' NO DATA 2.30E-05
- 1. 41 E-0 8 9.20E-09 2.01E-09 1.79E-08 4 11E-07 MN 54 NO DATA 1.99E-05 4.51E-06 NO DATA
--- 4.41E-06 NO DATA 7:31E-06 MN 56 NO DATA S.18E-07 1.41E-0 7 NO DATA FE 55 1.39E-05 7.03E-07 NO DATA 7.43E-05 8.98E-06' 2.40E-06 NO DATA NO DATA 4.39E-06 1.14E-06 FE 59 3.08E-05 5.30E-05 2.12E-05 NO DATA NO DATA 1.59E-05 2 5 7E-05 CO 58 NO DATA 3.60E-06 8 98E-06 NO DATA NO DATA CO 60 40 DATA 1.00E-05 2. 55E-0 5 NO DATA NO DATA 4.9tE-06 NO DATA NO. DATA 2.5TE-05 N143 6.34E-04 3.92E-05 2 20E-05 NO DA TA NO DA TA NO DATA 1.95E-04 NI 65 4.70E-06 5.32E-07 . 2 42E-0 7 NO DATA NO DATA NO DATA CU 64 40 DATA 6.09E-07 '- 2 82E-0 7 NO DATA 4.0$E-05
', 1.03E-06 NO DATA 1.2SE-05
- IN 65 + 1 84E-05 6.31E-05 2 91E-05 NO DATA 3.065-05 40 DATA 5.33E-05 IN 69 1.33E-04 1.68E-07 1 25E-08 NO DATA BR 83 6. 90 E NO DATA 1.37E-05 NO DATA NO DATA 3 63E47 NO DATA NO DATA SR 84 NQ DAT& . NO DATA NO DAT A. LT E-24 3 82E-07' N0' O&TA NO DATA NO DATA LT E-24
.- =
' #R 85 NO DATA NO DATA 1 94E-08 NO DATA NO DATA R$ 86 NO DATA NO DATA L7 E-24 M8 88 1.70E-04 8 40E=05 NO DATA NO DATA NO DATA 4.35E-06 NO DATA +.98E-07 2 73E=07 NO DATA
.,, NO DATA. NO DATA. 4.85E-07 f1 RS 49 NO DATA 2.865-07. 1.9 7E-0 7 NO DATA
'b SR- 89 NO DATA NO DATA 9.74E-08 2.51E-03 NO DATA 7.20E-0 5 NO OATA NO DATA NO DATA
'3 SR 90 1.85Er02. NO DATA 5.16E-05 4.71E-a0 3 NO 04TA NO DATA NO DATA 2.31E-04 SR 91 5.00E-05 NO DATA 1.81E-06 NO DATA NO DATA NO DATA 5.92E-05 SR 92 1.92E-05 NO DATA 7.13E-07 NO OATA i Y 90 NO DATA NO DATA 2.07E-04 0.69E-08 40 DATA 2.33E-0 9 NO DATA NO DATA 1 NO DATA 1.20E-04 l V 91M 8 10E-10 NO DATA Y 91 2 76E-11 NO DATA NO DATA NO DATA 2.70E-06 l 1.13E-06 NO DATA 3.01E-08 NO DATA NO DATA NO DATA Y 92 7.65E-09 NO DATA 8.10E-05 2.15E-10 NO DATA NO O&TA NO DATA 1.46E-04 4
Yh j 1.109-65 m _
4.4.WB.. . . . .
N@, s. M.M.. W... E..2... 5.2. ., T. .e. m..w..,.
. . .a..e. .- .s c.v.wa.m,
_ . .s._ s... . ., . . e- . _ . . .... . . . _ _~..~ _ _
'h j -
l
, f' a
i
=
I i . TA8LE,E-14e CONT'O l
i PAGE 2 0F 3 l
INGESTION 005E FACTORS POR INFANT
, (NREN PER PCI INGESTEDI NUCLICE SONE LIVER T.800Y THYROID KIONEY LUNG GI-LLI Y 93 2.438-08 NO DATA 6.62E-10 NO DATA NO DATA NO DATA 1.92E-04 l ZR 95 2.06E-07 5.02E-08 3.56E-08 NO DATA 5.41E-08 NO DATA 2.50E-05' ZR 97 1.48E-08 2.54h09 1.16E-0 9 NO DATA 2.56E-09 NO DATA. 1.62 E-M j . - . -
- 18 95 4 20 F 08- 1.73E-04 1.00E-08 NO DATA 1.24E-08 NO DATA 1.46E-05
, MO 99 NO DATA 3.40E-05 6.63E-06 NO DATA 5.08E-05 NO DATA 1.12E-05
. TC 99N 1.92E-09 3.96be9 5 10E-08 NO DATA 4 26E-04 2.07E-09 1.15E-06 l
TC101 2 27E-09 2.86E-09 2.83E-04 NO DATA 3.40E-08 1.56E-09 '4.86E-07 j !. RU103 1.488-06 NO DATA 4. 95E-0 7 40 DATA 3.08E-04 NO DATA 1.80E-05 l -
RU109 1.368-07 NO DATA 4 58E-08 NO DATA 1.00E-06 NO DATA 5.41E-05 RU106 2.41E-05 NO DATA 3.01E-0 6 NO DATA 2.85E-05 NO DATA 1.83E-04
, AG110N 94964-07 7.27E-07 4.81E-07 40 DATA 1.04E-06 NO DATA 3.T7E-05 TE125N 2.338-05 7.79E-04 3.15E-04 7.84E-04 NO DATA NO DATA 1.11E-05
. TE127N 5.458-05 1.94E-05 7.04E-04 1.49E-05 1.445-04 NO DATA 2.36E-05 TE127 1.00E-04 3.1SE-0 T . 2.15E-0 7 8.14E-07 2.448-06 NO DATA- 2 10E-05
. TE129N 1.005 3.43E-05 1.545-0 5 . 3.84E-05 2.54E-04 NO DATA 5.97E-05 TE129 2.84E-07 9.79E-08 6.63E-04 2.384-07 7.0TE-07 NO DATA 2.27E-05
' TE131N 1 52E-09 4.12E-06 5.058-06 1 24E-05 4.21bOS NO DATA 1.03E-04 TE131 1 765,07 4 50E-08 4.948-08 1.5 7E-07 ' 4.50E-07 NO DATA T.11E-06 TE132 2.088-09 1.038-05 9.61E-06 1 52E-05 4.44E-05 NO DATA 3.81E-05
. ! 130 4.00E-06 1.325-05 5.30E-06 1 48E-03 1.45E-05 NO DATA 2.83E-06 t 131 3.59E-05 4 235-05 1 868-05 1 395-02 4.94E-09 NO DATA 1.51E-06 1 131 1.668-04 .3.37f-06 1.20E14- 1.588-04 3.768-06 NO DATA 2.73E-06 I 133 1.255-05 1.82E-OS 5.33 E-06 3.31E-03 2.145-05 NO DATA 3.08E-06 t 134 8.498-07 1 78E-04 6.33PO T 4.15E-05 1 998-06 NO DATA 1.84E-06 .--
- 1 135 3.64E-06 7.245-06 2.64E-06 6.49b04 0.07E-06 NO DATL 2.62E-04 l C5134 3 77E-04 -7.03E-04 7 10E-05 NO DA TA 1.81E-04 7.42E-05 1.91E-06 '
C5136 4.59E-05 1.355-04 5. 04E-0 5 NO DATA 5.38605 1 10E-05 2.05E-06 C5137 5 22E-04 6.11F04 4.'33E-0 5 NO 0474 1.64E-04 6.64E-05 1.91E-04 C5134 4 818-07 7.82E-07 3.79E-0 7 NO 04TA 3.90E-07 6.09E-08 1.2SE-06
. 84139 8.81E-07 5.845-10
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- 2. 55 E-0 8- 40 DATA 3.51E-10 3.54E 10 5.58E-05 O O
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TASLE E-14, CONT'O' PAGE 3 0F 3 INGESTION DOSE FACTORS ~ POR INFANT (NREM PER PCI INGESTE08 NUCLICE, BONE- LIVER T'.800Y THYP.0!0 K!DNEY J
LUNC 'GI-LLt JBA140 1.71E-04 1.71E-07 8.81E-04 NO DATA 8A141 4.25E-07 2.91E-10 1.34E-0 8 ' No D AT A 4.06E-08 1.05E-07 4 20E-05 1 t
BA142 1.84E-07 1.53E-10 9.06E-49 ,NO DATA 1.75E-10 1.77E-10 .5 19E-06 '
8.81E-11 9.26E-11 7.59E-07 LA140 LA142 2 11E-08 8.32E-09 2 14E-09 NO DATA NO CATA NO DATA 9.77E-05
,- ( CE141 1.10E-09 4 04E-10 9.6TE-11 ' NO DATA NO DATA NO DATA 6.86E-05
, 7.87E-04 4.80E-04 5. 65E-0 9 NO DATA 1 48E-08 NO DA7A i
. 2.48E-05 CE143 CE144 1.44E-08 9.82E-06 1 12E-09 NO DATA 2 86E-09 NO DATA 5.73E-05
,a 2.98E-06 1 22E-06 1. 67E-07 . NO D A T A 4.93E-07. NO DATA
.PR143 8 13E-04 3 04E-08 4. 03 E-09 NO DATA 1.71E-04 1 13E-08 NO _. .
DATA' '4.29E-05 PR144 2.74E-10 1.06E-10 1. 38E-11 NO DATA
,~'
' NO147. 5.53E-08 5.68E-04 3. 48E-09 NO CATA 3.844-1L NO DATA 4.93E-06 W 187 9.03E-07 6.28E.-07 2 17E-0T NO 04TA 2 19E-05 NO DATA - 3.60E-05
.t NO DA TA ' NO DATA 3.69E-05
- _ ~ .--
NP239 1 11E-08 '9.93E-10 5. 61E-10 NO DATA t 1.98E-09 NO DATA 2.87E-05 6
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- I RADIATION PROTECTION ICRP PUBLICATION 38 I
Radionuclides Transformations Energy and Intensity of Emissions
~
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Report of a Task Group of Committee 2 of the International Commission on Radiological Protection !
on data used in ICRP Publication 30 l
I l
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l PUBLISHED FOR The International Commission on Radiological Protection ,
by PERGAMON PRESS OXFORD NEW YORK TORONTO SYDNEY PARIS FRANKFURT
4 '
UK Pergamon Press Ltd., Headington Hill Hall, Oxford OX3 OBW, England USA Pergamon Press Inc., Maxwell House, Fairview Park, Elmsford, New York,10$23, USA .
CANADA Pergamon Press Canada Ltd., Suite 104, 150 Contumers Road, Willowdale, Ontario M21 1P9, Canada AUSTRALIA Pergamon Press (Aust.) Pty. Ltd., PO Box 544, Potts Point, NSW 2011, Australia FRANCE Pergamon Press SARL,24 rue des Ecoles, 75240 Paris, Cedes 05, France FEDERAL REPUBLIC Pergamon Presa GmbH, Hammerweg 6, OF GERMANY D-6242 Kronbery Taunus,1 ederal Republic of Germany Copyright C 1983 The laternational Comminion on Radiological Protection The fmernational C-W- on Radiological Protection encouropes the puNication of translations of tkis reparr.
Permissionfor such transsarians and their puNication will-nmnally be pioenper ofcharge. he part of tins publication neay be reproduced, stored in a retrieval system or trannnitted in anyforms or by any neeens, electronic, electrostatic, mapneric tape - * '-4 photocop pay, recording or otherwise or repuNisand in anyform, w(thout pernaission in writingfem 3 8w copyright owner.
d First editice 1983 ISBN 0 08 030760 4 (Hardcover)
- ISBN 0 08 0307612 (Fleiucover)
Printed and bound in Great Britain by William Clowes Limited, Beccles and Imdon k
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. IODINE 453
._ . .53-IQDI.NE-131
" . HAIIldFE -6.04 DAYS - 21-JAN-76 DECAY MODE (S): 4-
- ~ ~
y(i) E(i)
~ RADIATION
~- (Bo-s)-8 .
. (MnV) y(i)= E(i)
- ~ 1 2.13E-02 6.935E-02* 1.48E-03
- 2 6.20E-03 8.693E-02* 5.39E-04
- ~ 3 - - 7.36E 9.660E-02*- 2.11E-03 44 8.94E-01 1.915E-01* 1.71E-01
'# 6
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42E 2.832E-01* - 1.19E-03 y1 _. .2.62E-02 6018E-02 2.10E-03 co-K, y 1 3.63E-02 4.562E-02 1.66E-03 co-14 , y 1 -
4.30E-08 -
7.473E-02 3.21E-04 y' 4 2.65E-03 1.772E 4.70E-04 y7 6.06E-02' ~ ' Y.843E-01 ~ 1.72E-02 co-K, y 7 2.48E-03 -. - 2.497E-01 6.20E-04 y 12-
- 2.51E 3.258E-01 8.18E-04 .
y 14 8.12E-01 3.645E-01 2.96E-01 co-K, y 14 1.55E-02 3.299E-01 5.10E-03 co-In, y 14 ._. 1.71E-03 3.590E-01 6.13E-04
~
y 16 3.61E-03 5.030E-01 1.82E-03 y .17 - 7.27E-02 6.370E-01 4.635 -02 y 18 --- - -22E-03-- ~6.427E-01 1.41E-03 y 19 1.80E-02 7.229E-01 1.30E-02 Kai X-ray 2.59E-02 2.978E-02 7.72E-04 Ka, X-ray 1.40E-02 2.946E-02 4.12E-04 LISTED X. y AND ys RADIATIONS 3ADE-01 OMITTED X, y AND y: RADIATIONS" 1.00E-03 LISTED #, ce AND Auger RADIATIONS 1.90E-01 OMITTED # cs AND Auger RADIATIONS" - ~ ~~
1.86E-03 LISTED RiA !AT!D S 5.70E-01' OMITTED RADIATIONS" 2.96E-03 AVERAGE ENERGY _.(MeV) .
" EACH OMTITED TRANSITION CONTRIBITTES
<0.100% 'IO Zy(i)xE(i)..IN .lTS . CATEGORY.
XENON-131M DAUGHTER, YIELD 1.11E-02, IS RADIOACTIVE.- ----
XENON-131 DAUGHTER. YIELD 9.889E-01 IS STABIE - - - - - - - .- - . .
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Page 8 ATTACHMENT 4' (07/88 - 12/88)
REVISIONS TO PROCESS CONTROL PROGRAM fPCP)
As required by Technical Specification 6.14, revisions to the PCP for the time period covered by this report are included. The revisions are synopsized below. A list of the supporting documentation and affected pages of the PCP are provided on the next page.
The Process Control Program (PCP) manual was deleted and replaced by- the new Radiation Plan procedure HP-7.2.20 " Process Control Program."
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r .: ,
LIST OF SUPPORTING DOCUMENTATION FOR CHANGES TO THE PROCESS CONTROL PROGRAM SNSOC Approved Request to Change Procedure and Routing Forms with the Procedure Review Checklist for replacing the PCP Hanual with the Radiation Protection Plan Procedure HP-7.2.20, " Process Control Program."
- New North Anna Health Physics Procedure, HP-7.2.20
-i l
REQUEST TO CHANGE PROCEDURE ADM-3.4 NORTH ANNA POWER STATION Attachmanc 3
. VIRGINIA POWER Page 1 of 1 06-23-88 1 JPERVISOR RESPONSIBLE FOR FOLLOWING PROCEDURE: ;
O ABNORMAL O CURVE BOOK OOPERATING O WELDING O ADMINISTRATIVE O EMERGENCY O PERIODIC TEST O ANNUNCIATOR Q
O IN-SERVICE INSPECTION @i!EALTH PHYSICS O O CALIBRATION MAINTENANCE- O SPECIAL TEST O CHEMISTRY Q
O NON-DESTRUCTIVE TEST O START-UP TEST Q PROCEDURE NO: g ., 2 UNIT NO: //y 3 REVISION DATE: 4 T M E: ff,(,,g [,7,g, L g, gy, 5 t CHANGES REQUESTE: (GIVE STEP NUMBER, EXACT SUGGESTED WORDING, AND LIST REFERENCES.. STAPLE 6 COPY OF PROC EURE WITH SUGCEST E CHANGES MARKED TO THIS FORM.)
Imhanta rA nos) of Alew /keteDM ES3 NOT bCONTRnr1e,
~
--mtw 4 4 sr s
- REASON FOR CIAiiGzi: '
/ fugg 7
_ Ron n,~
ImhrmrarAnon of 72Je 2Anoproca 6*nm p orJ ftAu - a mcapu ,, _.
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8 DATE:
r "GE REQUESTE BY: %> vus ar 0 E+.
9 ol's ks
~TCTION TAKEN: 10 DOES THIS CHANGE TR OPERAnmG innamUus AS DESCRIBD IN Tas UFSARf QTES 3 180 DOES TRIS CHANGE INYOLVE A CRANGE TO TEE TECE. SPECS? QYES STto DOES THIS CHANGE INVOLVE A POSSIBLE UNRkFIEWED SAFETT QUM TION? QYES G180 IF ALL "NO", NO "SAFETT ANALYSIS" IS REQUIRE. IF ANY "YES", A " SAFETY ANALYSIS" IS REQUIRE.
(10CF150.59) AFFROVD C0FT TO BI PROVIDE TO LICESING COORD. FOR INCLUSION IN ANNUAL RDORT.
~
RECOMENDD ACTION: 11 AFFROFID QDISAFFROVID DOES TEIS PROCDURE CREATE A i
QA DOCUM NTT OTES 9N0-BY: (COGNIZANT Suraava.umJ 12 13 REVIEWED BY QUALITY As vaABCE:
ES MADE: Q YES d 14
'BY:
g 15 DATE 16 REVIEWID BY STATION NUCLEAR SAFETT AND OPERATING COMNInsas f 17
@KFFROVID O DISAFFROVED LJ AFFROVED AS WDIFIE BY, COMMITTEE i l
~I RMAN SIGNATURE: 18 DATE: 9 NEW PROCEURE REVISION DATE:
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,/ ,j g 4 20
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ACTION COMFLETE BY: oF.Ol" M . 21 DATE: 22
\
ADM-5.4-j- Attachasnt 2 c NORTH ANNA POWER STATION' Page 1 of 1 PROCEDURE ROUTING SLIP . 06-23-88
~
RUCTIONS: When you have completed the action for which this procedure was sent to you, date and initial, then forward the procedure to the next person or department.
DO NOT REMOVE this routing slip.
UNIT /?'2 PROCEDURE NO. M/ 7 7. 70
@ NON-SAFETY DOES THIS CHANGE INVOLVE A CHANGE IN PROCEDURE CLASSIFICATION 7 YES @
(EWPROCEDU$ REVISION c.7j g DELETION REQUIRED APPROVAL DATE /5/^lM REQUIRED DISTRIBUTION DATE COMMITMENT DATE i2/89$E
- ROUTING IS REQUIRED TO ALL DEPARTMENTS THAT COULD BE AFFECTED BY THIS CHANGE.***
(I through the routing order if no routing required.)
ORDER TITLE INITIALS DATE ORDER / TITLE INITIALS DATE COGNIIANT SUFV. $ (2, g/s.//rV SUPT.- TECE. SERV.
ASST STATION MANAGER r. SUPT.- PROJECTS .
O&M ASST STATION MANAGER NS&L g/4 SUFT. - HEALTE PHYS $5[//-/M MANAGER, QUALITY ASSURANCE M ggg SUFY. ADN. SERVICES SUPT. - OPERATIONS y g . - PLANNING SUPT. - TRAINING DOCUMENT SUPT. - MAINTENANCE J AN 1 '/1989 RETURN TO COGN1?.4NT DEFAITMENTNOR NECESSARILY THE THIS PROCEDURE IS BEING L " M D FR(Mt FORWARDED T0/FOR: '
FR(0FING ACTION COMPLETED CRECR ~ DATE -
DATE INITIAL Cognisant Supervisor ~for Review / Approval l' or SN50C Eaview/ Approval Typing, First Time Affia Attachment 3 if entire procedure ses retyped Cognisant Dept Froof Reading (I) ~~
Typing Corrections (I)
Cognisant Dept: Froof Reading (II)
Typing Corrections (II)
Cognizant Dept: Froof Reading (III)
Typing Corrections (III)
Cognizant Dept: Froof Reading (IV)
Drafting
, Cognizant Dept: Froof Reading-Drafting Station Records, for Processing and Distribution SPECIAL NOTES / INSTRUCTIONS:
ORIGINATOR / DEPARTMENT $ A E~ m / NO
e PROCEDURE ADM-5.4 REVIEW Attachmsnt 4 CHECKLIST Page 1 of 1 06-23-88 PROC. NO: MN 7/ 20 CURRENT REV: DATE:
PROCEDURE TITLE: koctSS (l>arjec c Bcus-FOR NEW, REVISED, OR PROGRAMMATIC UPGRADE:
(CHECK) (2)
[ Human Factors Review Criteria (Ref: Attachment 10) s/ Radiological Work Practices Criteria (Ref: Attachment 11)
/
General Procedure Review Criteria (Ref: Attachment 12)
FOR CHANGES:
(3),
(CHECK)
Latest revision of existing procedure used Changes and location of their placement clear Deletions do not remove couaitted edi:erial/information Additions clearly portray equipment, readings, data, etc.
Setpcints and/or acceptance criteria changed Calculational basis provided or updated per ADb 17.15 l (4)
Review Completed Bys wr k. L en 5. Date: bl>3lTf Department: hhntrH ysses
. HP 7.2.20 Page 1 of 9
~ 1 FROCEDURE HP,7.2.20 k NORTH ANNA POWER STATION NUMBER i PROCEDURE DATE TYPE PROCEDURE: Health Physics UNIT # 1&2 TITLE: PROCESS CONTROL PROGRAM List of Effective Revisions:
Section Date l
1.0 PURPOSE . . . . . . . . . . . . . . . . .
2.0 REFERENCES
3.0 PRECAUTIONS AND LIMITATIONS . . . . . . .
i 4.0 INSTRUCTIONS . . . . . . . . . . . . . .
5.0 ATTACHMENTS . . . . . . . . . . . . . . .
6.0 FORMS . . . . . . . . . . . . . . . . . .
NOT A CONTROLLED DOCUMENT
,' AN 171989 NOR NECESSARILY THE L
\ 'llS!ON RECOMMEND APPROVAL:
DATE:
APPROVED CHAIRMAN SNSOC:
1 DATE:
HP-7.2.20 Page 2 of 9 PROCESS CONTROL PROGRAM 1.0 PURPOSE The Process Control Program provides instructions for processing and packaging of wet radioactive wastes to assure compliance with applicable Federal and State regulations for disposal of solid radioactive waste.
2.0 REFERENCES
2.1 Virginia Power Radiation Protection Flan. Chapter VII, " Radioactive Material and Effluents Control," Section 2, " Solid Radioactive Waste Control."
2.2 Code of Federal Regulations, Title ib, snergy, Parts 0 to 190.
2.3 Code of Federal Regulations, Title 49. Transportation, Parts 100 to 199.
2.4 North Anna Power Station Unit I and Unit 2 Technical Specifications; 1.20, 3/4.11.3, 6.8.1.g, 6.9.1.9, 6.14.1, and basis 3/4.11.3.
2.5 US NRC Low Level Waste Licensing Branch, " Technical Position on Radioactive Waste Classification" and " Technical Position on Waste Form," May 1983, Rev O.
2.6 US NRC, Standard Review Plan 11.4, " Solid Waste Management Systems" Rev 2, July 1981 (NUREG-0800).
3.0 PRECAUTIONS AND LIMITATIONS 3.1 The Supervisor Health Physics (Radwaste and Decontamination) is responsible for ensuring that applicable types of radioactive waste are processed and packaged in accordance with this procedure.
3.2 Changes to the PCP shall be in accordance with Technical Specification 6.9.1.9, including changes being submitted to the NRC as part of the Semiannual Radioactive Effluent Release Report.
3.3 The PCP consists of'instructi n steps describing the major elements of the PCP and establishing requirements for additional operational requirements and descriptions which may be included as attachments to this procedure.
- 3.4 For ion exchange resins and filter elements, the PCP provides specific requirements to l casure waste will be processed suitable for disposal at the Barnwell disposal site.
Applicable steps generally indicate the applicable condition of the Barnwell site license for reference.
3.5 Attachment 3," Definitions and Additional Descriptions" provides definitions of terms and other de:cription or consideration specific to the PCP. Procedure users should be familiar with the contents of this attachment.
4 HP-7.2.20 Page 3 of 9.
4.0 INSTRUCTIONS 4.1 System Descriptions. Waste Sources and Requirements
)
4.1.1 The types of wet radioactive waste produced at the station which must have means to process for disposal are:
- a. Ion exchange bead resin,
- b. Filter elements.
- c. Waste oil. -
- d. Liquid waste.
4.1.2 Station systems which normally process radioactive liquids with the subsequent generation of spent radioactive ion exchange bead resin and/or filter elements which must be processed for disposal are:
- NOTE
- The following systems are briefly described in Attachment 3,
[ Definitions and Additional Descriptions."
- a. Primary Coolant System,
- b. Boron Recovery System,
,. c. Spent Fuel Pit Purification System,
- d. Vent and Drain System, and
- c. Liquid Waste Processing System.
4.1.3 If primary to secondary leakage exists, and the Condensate Polishing System is processing secondary condensate, the ion exchange resin and filter elements usert in the system may become radioactive, and if so must be processed for disposal.
4.1.4 If lubricating / cooling oil becomes contaminated with radioactive material and if the oil will be disposed of as radioactive waste in a licensed land disposal facility, the oil should be considered and processed as wet radioactive waste.
4.1.5 If liquid wet waste is produced which must be disposed of (for example evaporator bottoms or decontamination solutions) such waste is to treated as wet radioactive waste.
4.2 Cleanirientian of Radioactive Waste 4.2.1 Processed wet waste shall be classified in accordance with 10CFR61.55.
4.2.2 Classification of wet waste shall be in accordance with procedure HP-7.2.21,
" Sampling, Analyzing and Classifying Solid Radioactive Waste."
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HP 7.2.20 Page 4 of 9 4.3 Processinn Wet Radioactive Waste 4.3.1 Ion exchange resins shall be processed by dewatering and/or solidification.
4.3.2 Filter elements shall be processed by dewatering or encapsulation in a solidification binder. j 4.3.3 Waste oil shall be processed by solidification, absorption with stabilization, or transferred to a licensed waste processor for disposal.
NOTE: The following applies if a liquid is not to be released as an effluent or treated further, but must itself be processed as radioactive waste.
4.3.4 Liquid wet waste must be processed by solidification.
4.3.5 If waste is classified as Class B or Class C waste; it shall be stabilized prior to disposal (10CFR61).
- a. Filter media shall be encapsulated in a solidification media prior to disposal or disposed of in a high integrity container (NRC BTP, C.5).
4.3.6 If required by the disposal site and/or the ' disposal site license conditions, certain categories of Class A waste shall be stabilized prior to disposal.
4.4 Procedures and Instructions to Imolement the Process Control Pronram 4.4.1 Acceptable methods which shall be used for waste stabilization are provided for specific disposal sites by procedures in the HP-7.2.4x series for the following waste types:
- a. Ion exchange bead resin.
- b. Filter elements.
- c. Waste oil.
- d. Liquid waste.
4.4.2 Acceptable methods which shall be used to dewater ion exchange resin and filter elements are provided in procedure HP-7.2.70, " Packaging Radioactive Waste."
4.4.3 Acceptable methods which shall be used to transfer wet waste to licensed waste processors are provided in procedure HP 7.2.50," Shipments of Radioactive Waste to Processing Sites."
l 4.4.4 If solidification of wet waste is applicable, it shall be performed in accordance with Attachment 1, " Solidification of Wet Waste."
4.4.5 If filter elements are to be stabilized by encapsulation, it shall be consistent with l the procedure provided in Attachment 2," Dewatering and Encapsulation of 4 Filter Elements." .
H P-7.2.20 Page 5 of 9 4.5 Requirements When Contractor Services are Used NOTE: If an outside contractor is used to provide a temporary solidification system on site for waste solidification, the following steps are applicable prior to solidification of radioactive waste. The solidification system is not to be used for solidifying radwaste until the operating procedures are approved.
4.5.1 Obtain the following, as a minimum, for review and evaluation:
- a. A detailed system description, which may be included in a topical report or equivalent documentation.
- b. Solidification system operating procedures which include process control parameters.
- c. A list of physical interfaces and station materials / services required.
- d. A list of expected utility / contractor responsibilities.
4.5.2 Compare the system description and operating procedures to the requirements provided in Attachment 1," Solidification of Wet Waste" to ensure the system can be operated within requirements.
4.5.3 The solidification system operating procedures shall be submitted to de Station Nuclear Safety Operating Committee for review and approval.
- a. After approval by SNSOC, procedures are to be considered applicable station procedures, and procedure compliance is required.
4.5.4 Ensure the contractor:
- a. Provides a system as proposed, described, and as approved for use at the station.
- b. Complies with approved procedures.
4.6 Process Records and Documentation 4.6.1 If an outside contractor's temporary solidification system is used for waste solidification, ensure the following are forwarded to Records Management:
- a. The system description, which may be included in a topical report or equivalent documentation.
- b. Approved solidification system operating procedures.
4.6.2 Data sheets shall be used to record solidification data, including test specimen data.
- a. Completed data sheets shall be forwarded to Records Management following final review.
L-__-_-__-_--_____----_-_---__----_--------_---------- -
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' HP-7.2.20 '
Page 6 of 9 4
' 5.0' ATTACHMENTS ' a 5.1 Attachment 1, " Solidification of Wet Waste."
5.2 Attachment 2 " Dewatering and Encapsulation of Filter Elements." !
5.3 Attachment 3, " Definitions and Additional Descriptions."
. 6.0 FORMS j 6.1 None i
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, HP 7.2.20 Attachment I-L[ + Page 1 of 3 l
21 SOLIDIFICATION OF WET WASTE NOTE: North 1nna Power Station currently has no installed solidification system on site to
. process wet wastes. If solidification is required, a solidification system is to be provided by an outside contractor / vendor. The contractor shall be advised of their responsibilities for ensuring compliance with the PCP.
1.0 Process Control Pronram Solidification Parametern 1.1 Solidification parameters may include but are not limited to waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, and mixing and curing times.
R 1.2 Once established, the process control parameters will provide boundary conditions which assure solidification will be complete and that the requirements for waste form stability and for no detectable free standing liquids are met.
2.0 Samnline. Analvnin_ and Proccan Surveillance 2.1 Wet radioactive waste to be solidified shall be sampled and analyzed as required and compared to process control parameters as required by procedure.
2.2 Wet radioactive wtste which will be solidified and/or absorbed shall have a representative test specimen from at least every tenth batch of waste to be processed.
2.2.1 Analysis shall be performed on waste to define process control parameters,
- s. Results of analysis will be recorded on waste solidification data sheets, as
. required by solidification procedures.
2.2.2 The test specimen shall be solidified to verify the process to be performed is satisfactory.
2.2.3 If any test specimen fails to solidify:
- a. The solidification batch under test shall be suspended until such time as additional.sampics can be obtained, alternative solidification parameters can be determined.
- b. Subsequent test (s) mu:t verify solidification,
- c. Solidification of the batch may then be resumed using the alternative solidification parameters determined.
- d. Test specimen shall be obtained for each subsequent batch of the same type of waste to be solidified, and test solidification performed.
I. Obtaining test specimen shall continue until three consecutive initial test specimen demonstrate solidification.
I 2.2.4 If required, the Process Control Program and/or applicable procedures shall be revised to ensure solidification of subsequent batches of waste.
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H P-7.2.20 Attachment l' ,
Page 2 of 3, ..
2.3 Samples of waste shall be obtained in accordance with procedure HP-7.2.21. " Sampling, Analysis and Classification of Solid Radioactive Waste" as required to determine 10CFR61 waste class.
3.0 Record and Data Sheets _
3.1 . Data sheets shall be used to record test sample solidification data.
3.2 The data sheets may include but are not limited to, type of waste to be solidified, major constituents, pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, mixing and curing times.
3.3 Data sheets should include batch numbr, batch volume, and date processed for each batch solidification and/or absorption.
4.0 Solidification Accentance Criteria NOTE: The following are general considerations. Specific site disposal criteria must be addressed based on the site to be used.
4.1 Solidified containers shall be verified as required by procedure to ensure containers are filled to at least 85% of capacity.
4.1.1 If a container is solidified with less than 85% of capacity,it shall not be shipped for disposal without prior approval from the disposal site.
NOTE: Free standing liquid criteria is based on the quantity ar received at the disposal site, not at the time of package closure. Normally the determination that a solidification batch will meet the criteria is made by ensuring the batch is solidified in accordance with'the PCP and may include additional visual or instrumentation inspections.
4.2 Solidified containers shall be verified as required by procedure to ensure free standing liquid meets disposal site criteria as follows:
4.2.1 Free liquid must be non corrosive.
4.2.2 If a high integrity container is not used, the maximum free liquid is 0.5% of the waste volume.
4.2.3 If a high integrity container is used, the maximum free liquid is 1.0% of the waste volume.
43 Stability of solidified waste may be considered to be satisfactory based on the following:
4.3.1 Solidification media and processes used to stabilize Class A liquids, waste forms as required by specific site criteria, or Class B or C waste shall meet and have been evaluated in accordance tvith NRC BTP C.2 (stability guidance) or other evaluation criteria specifically approved by the NRC or site license.
- s. If evaluation is required, refer to applicable documents as indicated.
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' HP-7.2.20 -
Attachment 1-'
Page 3 of 3 t
4.3.2 Solidified Class A' liquids shall meet the requirements of NRC BTP C.1:
- a. If Cisss A is segregated for Class B and C wastes, it should be' a free
^
Standing monoliths with no more than 0.5% free liquid.
- b. If Class A is not segregated for Class B and C wastes, it should met the requirements for Class B and C wastes. '
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HP-7.2.20 -
Attachment 2 Page 1 of 2 i DEWATERING AND ENCAPSULATION OF FILTER ELEMENTS NOTE: Filter elements are normally mechanical filters with wound fiber cartridges used for removing particulate from liquid systems. This attachment is only applicable to filter elements which are of the cartridge type.
1.0 Spent filter elements are normally removed from systems and placed in storage bunkers to await' processing and shipment.
2.0 Processing will be based on the waste classification of the filter.
NOTE: The following are general considerations. Specific site disposal criteria must be addressed based on the site to be used.
2.1 If filter media is classified as Class A waste and does not contain nuclides with half-lives greater than 5 years which have a total specific activity of I uCi/cc or greater, it may be disposed of as Class A waste.
2.2 If filter media is classified as Class B or Class C waste (per 10CFR61.55), it shall be encapsulated in a solidification media prior to disposal or disposed of in a high integrity container (NRC BTP, C.5).
3.0 Fif ter Elements to be Disnosed of as Clann A Waste 3.1 Filters should be allowed to drain dry in such a manner that any liquid trapped in voids is allowed to drain.
3.2 Filters shall not be compacted unless they are first allowed to dry essentially free of moisture.
3.3 If moist filters are to be packaged without compaction:
3.J.1 There shall be no indication of moisture on the filter in the form of drops or surface wetness.
3.3.2 Place the filters in a container or plastic bag to which absorbent material has been placed to absorb unintentional and incidental amounts of liquids.
- a. The amount of absorbent material should be equal to at least one-fourth tne volume of the filter.
3.4 Ensure the documentation indication package contents describes the presence of the filters.
4.0 Filter Flements to be Diseased of as Class B or C Waste 4.1 If the filter is to be solidified by being encapsulated in a solidification media:
4.1.1 Place the filter (s) in a suitable container such that the filter (s) will be completely surrounded by the solidification media when added.
- a. A basket type arrangement of thin wirc 4 recommended to hold the filter (s) in a fixed geometry.
H P-7.2.20 Attachment 2 Page 2 of 2 NOTE: The solidification media, including absence of free liquid, must be tested and documented in a manner required for solidification described in Attachment 1, " Solidification of Wet Waste."
4.1.2 Intro.1uce the solidification media into the container and fill the container to completely cover the filter (s) and to at least 85% of the capacity of the container.
4.1.3 The solidified filter container may be disposed of in any appropriate container for shipping and disposal at the disposal site. A high integrity container is recommended to ensure compliance with all requirements.
NOTE: Use of high integrity containers is addressed by procedure HP-7.2.70,
" Packaging Radioactive Waste."
4.2 If an encapsulated filter is to be disposed of in a high integrity container:
4.2.1 Properly place the container with the encapsulated filter in a high integrity container.
4.3 If an un-er.capsulated filter is to be disposed of in a high integrity container:
4.3.1 Place the filter (s) in the container such that the filter (s) will be h' e ld in a fixed geometry and such that liquids will not be trapped within the filter (s).
- a. A basket type arrangement of thin wire is recommended to hold the filter (s) provided the container C of C will not be violated.
4.3.2 If resin will be added, proceed with resin addition as appropriate.
4.3.3 Dewhter the container as applicable in accordance with procedure HP-7.2.70,
" Packaging Radioactive Waste."
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I H P-7.2.20 Attachment 3 Page 1 of 3 DEFINITIONS AND ADDITIONAL DESCRIPTIONS 1
Wet versus Dry Wastes (from NRC SRP 11.4, Branch Technical Position, ETSB 11-3)
Radioactive waste is generated in the form of " wet" and " dry" wastes. Wet wastes, including spent bed resins, filter sludge, spent powdered resins, evaporator concentrates, and spent cartridge filter elements, normally result as byproducts from liquid processing systems. Dry wastes, including activated charcoal, HEPA filters, rags, paper, and clothing, normally result as byproducts from ventilation air and gaseous waste processing systems and maintenance and refueling operations.
Liquid wet wastes such as evaporator concentrates are solidified prior to shipping, to render the waste immobile, to from a homogeneous solid matrix, absent of free water.
Adsorbents, such as vermiculite, are not acceptable.
Spent bead and powdered resins, and filter sludges, if acceptable to the receiving burial site, may be either solidified or dewatered (to less than the free liquid criteria) prior to shipping. In addition, the activity of dewatered wastes may dictate the type of container to be used.
Spent cartridge filter elements may be packaged in a shielded container with suitable absorbers such as vermiculite, although it would be desirable to solidify the elements in a suitable binder.
Process Control Pronram (PCP)
The PCP shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10CFR20,10CFR71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste (Tech Spec definition).
Stabilization or Stability A structurally stable waste form will generally maintain its physical dimensions and its form under the expected disposal conditions. Structural stability can be provided by the waste form itself, processing the waste to a stable form, or placing the waste in a disposal container or structure that provides stability after disposal (10CFR61.56(b).
Solidification Solidification shall be the conversion of wet waste into a form that meets shipping and burial ground requirements (Tech Spec definition).
Free Liould Free liquid is liquid which is still visible after solidification or dewatering is complete, or is drainable from the low point of a punctured container (NRC SRP !!.4, ETSB 11-3).
H Po7.2.20 Attachment 3 j
. Page 2 of 3 1
H.il.tidL A quantity of waste'which has been mixed or may be mixed to produce a homogeneous I mixture for th'e purposes of sampling, testing, and processing. Different samples of the homogeneous mixture would be expected to exhibit similar chemical and physical properties. A batch should not to be considered to be smaller than the quantity of waste s which fills one disposable liner or drum.
Test Soecimen i
A sample obtained from a batch of waste to be processed (solidified or absorbed), or a
' simulated sample of similar chemical and physical characteristics, on which a test can be-performed to verify the intended process will perform satisfactory.
l Comoosite A mixture of samples proportional by volume to the individual transfers making up a batch, thus resulting in the test specimen being representative of the batch, Soent Ion Exchanne Re1181 Resins are considered spent when decontamination factors indicate a significant decrease or when activity levels reach a pre-determined level.
Non-Corrosive Liauid in lieu of specific tests, a liquid may be considered to be non corrosive if it has a pH between 4 and 11 (based on NRC BTP C.2.h).
Hinh Intenrity Container A container designed to provide long term structural stability to contained waste during the required disposal period. May be used as an alternative to waste solidification. See NRC BTP C4 for more details if desired. High integrity containers must be approved by the appropriate agency.
Primary Coolant and Chemical and Volume Control Svstems Reactor coolant is purified by processing the letdown flow thr ugh a letdown filter, a mixed bed demineralized, a cation demineralized (optional), deborating demineralized (optional), and a reactor coolant filter to the volume control tank. From the volume control tank, part of the coolant is recycled to the reactor via charging pumps. Part of the coolant is routed thru a seal water filter to the charging pumps, and to a seal water injection filter (for coolant pump seals). Part of the coolant may be routed to the boron recovery system (normally to reduce reactor coolant boron concentration).
Boron Recovery System Reactor coolant routed to this system may be processed thru cesium removal ion exchangers and/or boron recovery filters for removal of additional radioactivity prior to interim storage in the boron recovery tanks. If boron evaporators are used to reconcentrate the boric acid solution, the concentrated solution is passed thru boron evaporator bottoms filters. Evaporator distillate may be processed thru boron cleanup lon exchangers and boron clean-up filter prior to discharge or recycled as primary grade water.
' HP 7.2.20
. Attachment 3 Page 3 of 3 Sun,t Fuel Pit Purification System Water in the spent fuel pit is recycled thru the fuel pit ion exchanger and fuel pit filter to remove. radioactivity. Spent fuel pit skimmer filters are provided to remove debris trapped by the spent fuel pit skimmers.
Vent and Drain System Waste water collected in sumps may be passed thru the high level waste drain filter or the low level waste drain filter based on optional valve lineups. The water is then routed to holdup tanks for processing.
Liould Waste Processine System Liquid waste collected in the high and low level waste tanks may be processed thru disposable ion exchangers and filters for removal of radioactivity prior to discharge.
Laundry waste is normally not processed, but is sampled and analyzed prior to discharge to ensure compliance with applicable limits. The liquid waste evaporator is normally not used in lieu of the disposable ion exchangers.
Soent Resin Transfer System Spent resins from primary coolant purification systems are transferred (flushed) to the Spent Resin Holdup Tank which collects resins for decay pending disposal. These resins are then transferred to disposable containers for shipment and disposal.
Dewaterine Sv3 tem for Soent Resin Holduo Tank The dewatering system consists of the Spent Resin Holdup Tank, the Radwaste Metering Pump, the Spent Resin Recirculation Pump, the dewatering container, the shipping cask, the dewatering pump, interconnecting hose and piping, valves, instrumentation and l controls, and sample taps.. Station procedures specify minimum periods for dewatering I pump operation and other considerations.
Dewaterina System for Himh Level Lianid Waste Treatment System Resins are loaded into pressure vessels and when chemically exhausted, the resins arc sluiced to an applicable container for dewatering. The system contains from two to seven disposable filter / demineralized vessels. The dewatering system for the HLLW disposable filter / demineralized vessels consists of the disposable vessel and the dewatering pump.
Dewaterine System for Condenante Polishina System
~
Powdered resins are transferred to a separate holdup tank for later transfer to a dewatering container. The spent powdered resin recirculation pump is a progressive cavity pump, which takes suction frcm the bottom of the spent powdered resin tank and discharges to the top of the tank and provides positive pressure to the suction of the pump which pumps the resin into a dewatering container. As the dewatering container is filled, a dewatering pump removes excess water.
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Page 9 ATTACHMENT 5
( 07/88 - 12/88 )
MAJOR CHANCES TO RADIOACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS As required by Technical Specification 6.16, major changes to radioactive liquid, gaseous and solid waste treatment systems for the time period covered by this report are reported below. Supporting information as to the reason (s) for the change (s) and a summary of the 10 CFR Part 50.59 evaluation are included.
i i No major changes to the radioactive liquid, gaseous, and
-solid waste treatment systems were made for the time period l
covered by this report.
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Page 9 ATTACHMENT 6
( 07/88 - 12/88 ) l IN0PERABILITY OF RADI0 ACTIVE LIQUID AND GASE0US ;
EFFLUENT INSTRUMENTATION As required by Technical Specification 3.3.3.10.b and 3.3.3.ll.b, a list and explanation for the inoperability of radioactive liquid and/or gaseous effluent monitors is provided in this report.
On April 11, 1987, 1-RM-SW-108, the radiation monitor for service water discharged to Lake Anna was declared inoperable, due to sample pump (1-SW-P-10) not supplying adequate flow. It was found that the pump suction piping from the Service Water Return Headers A and B was clogged, and not able to be isolated. EWR 88-052 was initiated to modify the suction piping to 1-SW-P-10 to prevent future clogging and to allow isolation of the pump for future repairs, if necessary, without requiring a Service Water outaae.
Work Order - 064303 to accomplish the EWR has now been completed and the pump and 1-RM-SW-108 have been returned to service. This was completed on December 27, 1988.
On November 27,19881-LW-P-28 was declared inoperable. The clarifier discharge sample pump has had operability problems due to clogging of the pump head check valves and the back pressure regulator. EWR 88-157 was initiated to solve these problems. Scheduled completion of EWR 88-157 is March 1, 1989. The Work Orders to accomplish EWR 88-157 should be completed by May 1, 1989. Every effort is being made to keep the pump in service until the EWR is implemented.
Page 11 ATTACHMENT 7
( 07/88 - 12/88 )
UNPLANNED RELEASES As required by Technical Specification 6.9.1.9, a list of unplanned releases, defined according to the criteria presented in 10 CFR part 50.73, from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents made during the reporting period is made below.
No unplanned releases, as defined according to the criteria presented in 10CFR Part 50.73, occurred during the time period covered by this report.
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Page 12 1 Attachment 8
'( 07/88 - 12/88 )
Changes Required By The Land Use Census Evaluation As required by Technical Specification 3.12.2 and 6.9.1.9, Evaluation of the Land Use Census is to be made for identifing the new location (s) for dose calculations and/or environmental monitoring pursuant to Technical Specification 3.12.2 requirements.
No new location (s) for dose calculations and/or environmental monitoring pursuant to Technical Specification 3.12.2 requirements were identified by the evaluation of the lastest Land Use Census.
Page 13 ATTACHMENT 9 Lower Limits of Detection For Effluent Sample Analysis
( 07/88 - 12/88 )
Gaseous Effluents Required L.L.D. Typical L.L.D.
Radioisotope (uCi/ml) (uCi/ml)
Krypton - 87 1.0E-4 1.52E 4.27E-7 Krypton - 88 1.0E-4 1.98E 4.95E-7 Xenon - 133 1.0E-4 1.24E 3.10E-7 Xenon - 133m 1.0E-4 5.63E 1.58E-6 Xenon - 135 1.0E-4 6.39E-8 - 1.81E-7 Xenon -
135m 1.0E-4 2.81E 7.90E-7 Xenon - 138 1.0E-4 7.86E 2.28E-6 Iodine - 131 1.0E-12 4.49E 6.18E-14 Manganese - 54 1.0E-11 4.13E 4.89E-14 Cobalt - 58 1.0E-11 4.12E 5.04E-14 Iron - 59 1.0E-11 9.11E 1.10E-13 Cobalt - 60 1.0E-11 6.72E 1.15E-13 Zine - 65 1.0E-11 8.28E 1.36E-13 Strontium - 89 1.0E-11 4.00E 7.00E-15 Strontium - 90 1.0E-11 7.00E 1.00E-15 Molybdenum - 99 1.0E-11 2.58E 3.70E-13 Cesium - 134 1.0E-11 3.55E 5.34E-14 Cesium - 137 1.0E-11 4.67E 6.56E-14 Cerium - 141 1.0E-11 4.32E 5.20E-14 Cerium - 144 1.0E-11 1.90E 2.43E-13 Gross Alpha 1.0E-11 1.02E 2.10E-14 Tritium 1.0E-6 1.22E 1.44E-7
Page 14 ATTACHMENT 9 Lower Limits of Detection For Effluent Sample Analysis.
( 07/88 - 12/88 )
( cont. )
Liquid Effluents Required L.L.D. Typiccl L.L.D.
Radioisotope (pCi/ml) (uC1/ml)
Krypton - 87 1.0E-5 2.21E-8 - 6.93E-8 Krypton' - 88 1.0E-S 5.90E-8 - 9.65E-8 Xenon - 133 1.0E-5 4.90E 6.81E-8 Xanon - 133m 1.0E-5 2.09E-7 - 2.78E-7 Xenon -
135 1.0E-5 2.60E 3.10E-8 Xeron - 135v 1.0E-5 1.13E 1.25E-7 Xenon - 138 1.0E-5 2.96E 3.85E-7 Iodine - 131 1.0E-6 2.57E-8 - 3.45E-8
___, Manganese - 54 5.0E-7 2.32E 2.87E-8 Iron - 55 1.0E-6 1.00E-6 Cobalt - 58 5.0E-7 2.32E 3.04E-8 Iron - 59 5.0E-7 5.14E 6.18E-8 Cobalt - 60 5.0E-7 3,.98E-8 - 6.47E-8 Zinc - 65 5.0E-7 4.92E 7.89E-8 Strontium - 89 5.0E-8 5.00E-8 Strontium -- 90 5.0E-8 7.00E-9 l Molybdenum - 99 5.0E-7 1.53E-7 - 2.25E-7 l
l Cesium - 134 5.0E-7 2.23E-8 - 3.09E-8 Cesium - 137 5.0E-7 2.86E 4.02E-8 Cerium - 141 5.0E-7 3.45E 4.19E-8 Cerium - 144 5.0E-7 1.52E-7 -
1.96E-7 Gross Alpha 1.0E-7 5.89E 1.21E-8 Tritium 1.0E-5 3.36E-6 - 3.96E-6
Page 15 ATTACHMENT 10
( 07/88 - 12/88 )
UNAVAILABILITY OF MILK OR LEAFY VEGETATION SAMPLES As required by Technical Specification 3.12.1.c., the identification of the causes of the unavailability of milk or leafy vegetation samples, required by Technical Specification Table 4.12-1, and the identification of the new location (s) for obtaining replacement samples are listed.
No unavailability of milk or leafy vegetation samples, as required by ,
Technical Specification Table 4.12-1, occurred during the time period covered by this report.
l' VIRGINIA ELECTRIC AND PownN COMPANY RIcnxoxn,VIRoINIA 2G261 March 2, 1989 United States Nuclear Regulatory Commission Serial No.89-108 Attention: Document Control Desk N0/JBL:jmj Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER CONPANY NORTH ANNA POWER STATION UNITS 1 AND 2 SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT Enclosed is the North Anna Power Station Semi-Annual Radioactive Effluent Release Report for July 1, 1988 through December 31, 1988. The report, ,,
submitted pursuant to North Anna Station Technical Specification 6.9.1.9, includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released during'the previous six months, as outlined in Regulatory Guide 1.21, Revision 1, June 1974.
Very truly oyrs,g W. R. Cartwri ht
[. Enclosure cc: U. S. Nuclear Regulatory Commission 101 Marietta Street, N.W. l Suite 2900 Atlanta, GA 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station 1 y
i h
j RADI0 ACTIVE EFFLUENT RELEASE REPORT l~ NORTH ANNA POWER STATION
( JULY 01, 1988 TO DECEMBER 31,1988)
PREPARED BY: #4d.4--F Assistant Supervisor Health Physics (Count Room & Environmental)
REVIEWED BY:
'C.
Supervisor Health Physics (Technical Services)
APPROVED BY:
g[ '
//
Superintendent Health Physics
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FORWARD This report is submitted as required by Appendix A to Operating License Nos.
NPF-4 and NPF-7, Technical Specifications for North Anna Power Station, Units 1 and 2, Virginia Electric and Power Company, Docket Nos. 50-338, 50-339, Section 6.9.1.9.
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