ML20099J801

Semiannual Radioactive Effluent Release Rept, Jul-Dec 1984

ML20099J801
Person / Time
Site:	North Anna
Issue date:	12/31/1984
From:	Barnes W, Harrell E, Stafford A, Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
85-141, NUDOCS 8503200062
Download: ML20099J801 (55)

Text

_ _ _ _ _ _ _ _ _ _

s SEMI-ANNUAL RADIOACTIVE Ei rLUENT RELEASE REPORT NORTH ANNA POWER STATION JULY 1, 1984, TO DECEMBER 31, 1984 1

PREPARED BY: d "

W. C. Barnes Assistant Supervisor Health Physics REVIEWED BY: / h A.H.Sta/TA5rd Superviso Health Ph sics APPROVED BY:

E. Wayne Haf f ell /

Station Mahager hp DO 8 R

f626 ll1

c_

i-f 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 s

~

Station, Units 1 and 2, Virginia Electric and Power Company, Docket Nos. 50-338, 50-339, Section 6.9.1.12..

4 l

RADIOACTIVE EFFLUENT RELEASE REPORT FOR THE NORTH ANNA POWER STATION JULY 1, 1984 TO DECfMBER 31, 1984 INDEX SECTION NO. SUBJECT PAGE 1 PURPOSE AND SCOPE . . . . . . . . . . . . . . . . . 1 2 DISCUSSION .. .................. 2 3 SUPPLEMENTAL INFORMATION . .. . . . .. . . . . . 3 Attachment i Effluent Release Data . . . . . . . 4 Attachment 2 Annual and Quarterly Doses . . . . 5 Attachment 3 Revisions to Offsite Dose Calculation Manual (ODCM) . . . . . 6 Attachment 4 Revisions to Process Control Program (PCP) . . . . . . . . . . . 7 Attachment 5 Major Changes to Radioactive Solid Waste Treatment Systems . . . 8 Attachment 6 Radioactive Liquid and Gaseous Effluent Monitoring ,

Instrumentation Inoperable. . . . . 9 Attachment 7 Unplanned Releases 10 l

f Pag 2 1

! 3 t

l 1.0 PURPOSE AND SCOPE l

The Radioactive Effluent Release Report includes a summary of the L quantities of radioactive liquid and gaseous effluents and solid waste as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting l

l Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Caseous Ef fluents from Light-Water-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following i

the format of Appendix B thereof. The report submitted within 60 days after January 1 of each year includes an assessment of radiation doses to the l maximum exposed member of the public due to radioactive liquid and gaseous l

effluents released from the site during the previous calendar year. The report also includes a list of unplanned releares during the reporting period.

As required by Technical Specification 6.15, ciianges to the ODCM for the time period covered by this report are included. 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 l

by Technical Specification 6.14.

Major changes to radioactive solid waste treatment systems are reported as required by Technical Specification 6.16. Information to support the l reason for the change and a summary of the 10 CFR Part 50.59 evaluation are included. In lieu of reporting major changes in this report, major changes to the radioactive solid waste treatment systems may be submitted as part of the annual FSAR update.

As required by Technical Specification 3. 3. 3.10.5 and 3. 3. 3.11.b a lis t I

and explanation for the inoperability of radioactive liquid and/or gaseous

. ' P gt 2 effluent monitors is provided in this report.

2.0 DISCUSSION The basis for the calculation of the percent of technical specification for the critical organ in Table 1A 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 fhr 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 end 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 1A 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 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin.

The basis for the calculation of the percent of technical specification in Table 2A 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 concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 microcuries/ml.

Percent of technical specification calculations are based on the total 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 doses due to noble gases released from the site were calculated at site boundary. The maximum exposed member of the public from iodine-131,

i Psga 3 c

from tritium, and-from all radionuclides in particulate form with half-lives greater than 8 days is defined as an infant, exposed through the grass-cow-milk. pathway, with the critical organ being the thyroid. The maximum exposed member of the public from radioactive materials in liquid ef fluents in unrestricted areas is defined as an adult, exposed by the fish pathway, with the critical organ being the liver. - The total body dose from liquid effluents is also determined for this individual.

Unplanned releases presented in Attachment 7 are defined according to the criteria presented in 10 CFR Part 50.73. Gaseous unplanned releases are those radioactive releases that exceed 2 times the applicable concentrations of the limits specified in Appendix B, Table II of 10 CFR Part 20 in unrestricted areas, when averaged over a time period of one hour. Liquid 7

unplanned releases are those effluent releases that exceed 2 times the limiting combined Maximum Permissible Concentration (MPC) specified in Appendix B, Table II of 10 CFR Part 20 in unrestricted areas for all radionuclides except tritium and dissolved noble gases, when averaged over a time period of one hour.

3.0 SUPPLEMENTAL INFORMATION Not included in this report are the Fourth Quarter, 1984, results of the following analysis:

Particulate Filters Ventilation Vent - A: SR-89/90, Fe-55 Ventilation Vent - B: SR-89/90, Fe Process Vent: SR-89/90, Fe-55 Liquid Composites Turbine Building Sump: SR-89/90, Fe-55 Clarifier Effluent: SR-89/90, Fe-55 When furnished these results by the' vendor, an addendum to this report

\

will be 7 properly submitted.

, j

. - Paga 4 ATTACINE'TT l_

EF57.UEST* RELEASF DATA _

(1/94 - 12/84)

This attachment includes a sumnary of the nuantities of rad.') active liquid and gaeous effluents and solid waste, as outlined in Regulatory Guide 1.21.

_. l

_.=* ~ qwx. ~ ' L 2 l ,.;  :- ':*~~ ~ '--~ i TABLE lA ~

EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1984 )

GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES ,

NORTH ANNA POWER STATION ,

Est. Total Units 3rd Quarter 4th Ouarter % Error A. Fission and Activation Gases: -

1. Total Release Ci 9.767 E + 3 2.065 E + 1 1.50 E'& 1

2. Average Release Rate for Period: pCi/sec 1.23 E + 3 2.60 E + 0 B. Iodines: -

1. Total Iodine-131 Release Ci 1.87 E - 3 1.87 E - 4 1.50 E + 1

2. Average Release Rate for Period: pCi/see 2.35 E - 4 2.35 E - 5 C. Particulates ( T > 8 days):

1. Total Particulate ( T > 8 days) Release Ci 8.169 E - 4 3.491 E - 4 1.50 E + 1

2. Average Release Rate for Period pCi/sec 1.04 E - 4 4.39 E - 5

3. Cross Alpha Radioactivity Release Ci 4.72 E - 6 2.37 E - 5 D. Tritium:

1. Total Release Ci 6.097 E - 1 3.285 E + 0 1.50 E + 1

2. Average Release Rate for Period DCi/sec ' 7.67 E - 2 4.13 E - 1 Percentage of Technical Specification Limit's:

~

E.

1.77 E - 1 3.93 E - 3

1. Total Body Dose Rate  %

2. Skin Dose Rate r  % 2.96 E - 2 . 9.02 E - 4

3. Critical Organ Dose Rate  % 2.32 E - 3 4.97 E - 4 L

TABLE IB Peg 2 1 cf 2 EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1984)

CASEOUS EFFLUENTS MIXED MODE REIEASES NORTH ANNA POWER STATION ,

' CONTINUOUS MODE BATQ! MODE I NUCLIDES REllASED UNIT 3rd QUARTER ' ' 4 th' QUARTER ' '1 rd QUARTER 4th QUARTER I Fission and Activation Cases:

I Ci *

• 1.43 E + 1

• Krvoton - 85 I Krvoton - 85m Ci 8.67 E - 3

[ Krypton - 87 Ci * .* *

• l Krypton - 88 Ci * *

• 6.47 E - 3 I Xenon - 131m Ci *

• 2.65 E + 2

• I Xenon - 133 Ci 6.35 E + 1 7.58 E + 0 9.08 E + 1 7.16 E + 0 I Xenon - 133m Ci *

• 4.34 E + 1 6.22 E - 2

. Xenon - 135 Ci 7.81 E - 1 1.31 E + 0 1.89 E - 3 4.52 E - 1 I Xenon - 135m Ci * * *

• Xenon - 138 Ci * * * *

. Other (Specify) Ci

. Argon - 41 Ci 3.08"E - 3 Ci Total for Period Ci 6.41 F + 1 8.89 E + 0 9.40 E + 3 7.69 E + 0

, Ci

! Iodines: Ci '

Iodine - 131 Ci 5.69 E - 6 9.30 E - 6 8.29 E - 5 5.12 E - 6 Ci *

• 3.44 E - 4 1.80 E - 8 Iodine - 132 Iodine - 133 C1 3.09 E - 6 3.23 E - 5 1.60 E - 8 5.73 E - 6 j Iodine - 134 Ci * * *

• l Iodine - 135 Ci

• 7.01 E - 8

• 1.30 E - 6 l Ci Total for Period Ci 8.78 E - 6 4.17 E - 5 4.27 E - 4 1.22 E - 5 l Ci Particulates: Ci i Strontium - 89 Ci

• Strontium - 90 Ci

• I Cesima - 134 Ci * * .*

• Cesium - 137 Ci 3.13 E - 8 1.15 F -R *

• Barium - 140 Ci * * * *

[ Lanthanum - 140 Ci * * *

• I Other (Specify) C1 Cobalt - 58 C1 2.07 E - 8'

• 2.22 E - 6 1.58 E - 8 Cobalt - 60 Ci

• 8.20 E - 8 5.85 E - 7 3.86 E - 9 Strontium - 85 *

• 1.15 E - 9

• Ci .

i Tellurium - 131m ( Tb < 8 days) Ci *

• 2.68 E - 9

• I Ci

• 1. css than lower limits of detection -

TABE IB EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1984) P gs 2 ef 2 CASEOUS EFFLUENTS MIXED MODE RELEASES .

NORTH ANNA POWER STATION CNTINUOUS HODE- RATG MODE

' i NUCLIDES RELIASED UNIT 3rd QUARTER' ' 4th' ' QUARTER ' '3rd QUARTER 4th QUARTER I

Particulates (con't)

I Sodium - 24 ( T < 8 days) Ci * * * ?_6n R -0 I Rubidium - 88 ( T < 8 days) Ci *~ *

• I Iron - 55 6.23 E - 5 t

Ci

• i Total for Period Ci 5.40 E - 8 1.14 E - 7 2.81 E - 6 l 6.23 E - 5 1 Cross Alpha Ci

• 5.89 E - 10 *

• Ci i Tritium Ci 3.69 E - 3 1.76 E - 1 7.99 E - 3 1.91 E - 4 Ci Ci C1 Ci

, Ci

. Ci s., Ci '

C1 Ci

! Ci i Ci

! Ci Ci I

ci i Ci '

i Ct '

C1 Ci i Ci '

C1 -

Ci Ci

} Ci l Ci Ci i

Ci .

I Ci l

Ci

• 1. css tisan lower limits of detection

! TABLE 1C Pcga 1 cf 2 ' '

l EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL PIPORT (1984) -

GASEOUS EFFLUENTS - GROUND - LEVEL RELEASES NORTH ANNA POWER STATION l .

CCNTINUOUS HODE' BATQi MODE

NUCLIDES REilASED UNIT 3rd QUARTER ** *l.rVQUARTER* '1'r'd QUARTER '4 th QU ARTE R

• Fission and Activation Cases:

I Krvoton - 85 Ci * * * *

• 5.43 E - 2 4.48 F + O 7.81 E - 5 I Krvoton - 85m Ci

• 1.08 E - 1 5.27 F + 0 1.64 F -4 I Krvoton - 87 Ci Ci

• 9.01 E - 2 7.07 E + 0 1.97 E - 4 i Krvoton - 88 Ci * *

• i Xenon - 131m

• I Xenon - 133 1.58 E + 2 3.30 E - 1 5.36 E + 1 4.65 F - 1 Xenon - 133m Ci *

~

Xenon - 135 Ci 1.98 E + l 8.82 E - 1 1.19 F4 1 6.24 E - 4 Xenon - 135m Ci 1.20 E + 0 1.71 F + 1 i ox F 1 Xenon - 138 Ci

• 5.01 E - 1 *

• Other (Specify) Ci 4.32'E - 1 Argon - 41 Ci

• 6.04 F + O 6.25 E - 4

~

Ci Total for Period C1 1.78 E + 2 3.60 E + 0 1.25 F + 9 L.64 F - 1

~

C'i 5 Iodines: Ci i Iodine - 131 Cl 1.78 E - 3 1.69 E - 4 1.81 E-6 1.27 F - 6 Iodine - 132 Ci *

• 1.54 E - 8 1.25 E - 5 s Iodine - 133 Ci 9.97 E - 4 1.52 E - 3 1.93 E - 6 2.25 E - 5 C'i *

• 1.06 E - 6 1.42 E - 5 I Iodine - 134 ~

C1 *

• 1.50 E - 6 2.34 E - 5 I Iodine - 135

Ci

! Total for Period C1 2.78 E - 1 1.60 F -1 '

6.32'E - 6 7.59 E - 5 i Ci 5

! Particulates: Ci I Strontium - 89 Ci '

C1'

• I Strontium 90 I Cesitsu - 134 C1 4.30 E - 5 2.15 E - 5 1,46 E - 4 1.87 E - 5 j Cesium - 137 C1 1.66 E - 4 1.11 E - 4 ~

2.92 E - 4 3.85 E - 5 Ci * * *

• i Barium - 140 Ci * * * *

! Lanthanum - 140 1 I Other (Specify) CL Manganese - 54 Ci *

• 3.37 E - 7 - 1.05 E - 8 l Cobalt - 58 Ci 7.81 E - 5 4.37 E - 5 1.65 F -R 4.60 E - 5

coDalt -ou Ci 7.89 E - 5 5.81 E - 5

• 1.87 E - 6 7.81 F - 7 i Strontium - 85 ~

Ci 7.82 E - 6 * *

• l C1

• 1. css than lower limits of detection

k TABE 1C , ,

EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1984) Page 2 cf 2 f CASEOUS EFFLUENTS - CROUND - LEVEL RELEASES t NORTil ANNA POWER STATION ,

'CCNTINUOUS HODE' BATQi MODE I NUCLIDES REEASED UNIT 3rd QUARTER ' ~4t'h ' QUARTER ' 3rd QUARTER 4th QUARTER i Particulates (con't) ,

I Iron - 55 Ci

• Rubidium - 86 Ci *'~ *

• 4.72 E - 6 l

• 5.72.E - 6 *

• Niobium - 95 Ci Sodium - 24 (T < 8 days) *

• 3.35 E - 9 4.93'E - 7

• 1.59 E - 3 6.34 E - 2

• Rubidium - 88 (T4 < 8 days) Ci 1 Rubidium - 89 (Tis < 8 days) Ci

• 7.32 E - 4 *

• i Antimony - 122 (Tis < 8 days) Ci *

• 1.71 E - 5 1.27 E - 6

. Tellurium - 131m (T < 8 days) Ci 6.61 E - 5 * * *

! Tellurium 132 (Th < 8 days) Ci

• 1.15 E - 5 1.38 E - 3

• Cesium - 138 (T < 8 days) Ci 8.11 E - 6 2. 31 E - 3 1.29 E - 1 2.83 E - 5 Ci Total for Period C1 4.48 E - 4 4.88 E - 3 1.94 E - 1 1.39 E - 4 C1 Cross Alpha C1 4.72 E - 6 2.37 E - 5 *

• Ci

! Tritium Ci 3.86 E - 1 3.09 E + 0 2.12 E' - 1 1.85 E - 2 C1 C1 Ci j Ci l Ci Ci

! Ci '

i Ci Ci '

C1 Ci i Ci -

Ci +

C1 Ci

! Ci

! Ci

[ Ci i C1 -

I Ci i C1

• Less than lower limits of detection -

L-- _ - _ _ _1

TABLE 2A -

- EFFLUENT AND WASTE DISPOSAL SE})I-ANNUAL REPORT ('1984)

LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES

' NORTH ANNA POWER STATION TST. TOTAL I

UNIT 3rdQUARTER 4th QUARTER ERROR %

1. Fission & Activation products

1. Total release (not including' tritium, gates, alpha) Ci 6.63 E - 1 5.25 E - 1 1.50 E'+ 1

2. Average diluted concent, ration during period pCi/ml 1.61 E - 9 8.13 E - 10

3. Percent of applicable limit (T.S)  % 9.84 E - 3 5.41 E - 3

1. Tritium

1. Total release activity. Ci 2.01 E + 1 1.72 E + 2 1.50 E + 1

~ 2. Average diluted concentration during -

period pCi/ml 4.89 E - 8 2.66 E - 7

3. Percent of applicable limit (T.S.)  % 1.63 E - 3 8.87 E - 3 .

2. Dissolved and entrained eases .

1 .

1. Total release: activity. Ci 1.59 E + 0 1.63 E - 1 1.50 E + 1

2. Average diluted concentration during period pCi/ml 3.87 E - 9 2.52 E - 10

3. Percent'of apolicable limit (T.S.) '

% 1.94 E - 3 1.26 E - 4 .

). Gross Alpha radioactivity'

1. Total release activity. -

. Ci < I LD 1.16 E - 4 1. 50 E + 1

. V21ume of waste released (prior to '

dilution)

• Liters 4.36 E +.7 8.08 E + 7

'. V:lume of dilution water used during period 4.11 E + 11' 6.46 E + 11 Liters

, , , - - _. --- _. -. ._. m. .a.- _ ~. .- -

TABLE 2B '

EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1984) Page 1 of 2 LIQUID EFFLUENTS NORTH ANNA POWER STATION .

' CONTINUOUS HODE' BATOi H0DE

- NUCLIDES RELIASED UNIT 3rd QUARTER ** ~4't'h' ' ' QUAR 1ER ' hs QUARTER '4th QUARTER i Fission and Activation Products:

I Strontium - 89 Ci

• I Strontium - 90 Ci '*

i Cesium - 134 Ci 1.52 E - 1 1.54 F - 1 4.55 E - 6 3.26 E - 4 i Cesium - 137 Ci 2.78 E - 1 3.17 E - 1 9.88 E - 6 6.97 E - 4 1 Iodine - 131 C1 1.07 E - 3 *

• 2.60 E - 6 i Cobalt - 58 C1 3.84 E - 2 1.59 E'- 2 9.70 E - 7 1.15 E - 4

Cobalt - 60 Ci '3.24 E - 2 1.29 E - 2 5.28 E - 7 'l.58 E - 4 Iron - 59 C1 2.21 E - 3 * *

• Zine - 65 Ci * * *

• Chromium - 51 Ci 6.28 E - 3 *

• 5.65 E - 6 Manganese - 54 ci 1.46 E - 3 *

• 5.75 E - 6 Niobium - 95 Ci 6.71 E - 3 5.16 E -8 2.91 E - 6 Zirconium - 95 Ci 4.56 E -- 3 3.32 E - 4 *

• Molybdenmn - 99 Ci * * * *

. Technetium - 99 Ci * * * *

[ Barium - 140 Ci * * *'

• Lanthanum - 140 Ci * * *

• Cerium - 141 C1 4.06 E - 5 * *

• i Other ( Specify) Ci i Antiruony - 122 Ci 5.25 E - 2 1.74 E - 2 2.55 E - 7 3.03 E - 5 i Sodium - 24 C1 1.63 E - 4 1.13 E - 5

• 1.02 E - 6 Antimony - 124 Ci 5.89 E - 3 7.45 E - 4 * *

! Silver - 110m Ci 5.05 E - 3 5.15 E - 4

• 4.74 E - 6 i C1 2.12 E - 3 * *

• Ruthenium - 103

. Tellurium - 129m Ci 3.93 E - 3' * *

• I fron - 55 Ci 7.03 E - 2

• I Iodine - 132 Ci * *

• 1.69 E - 5 i Iodine - 133 Ci * *

• 3.17 E - 5 j Iodine - 134 Ci * *

• 4.16 E - 6

-l Iodine - 135 Ci f' * *

• 2.28 E - 5 I Tellurium - 131m Ci * * * '3.38 E - 6 1 Cesium - 138 Ci * *

• 1.63 E - 6 i

Cobalt - 57 Ci = *

• l.34 E - 6 i Ci j Ci x 3 Total for Period C1 6763 E - 1 5.23 E - 1 1.62 E - 5 l.43 E - 3 I Ci '

' "r8" _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ ___

6 6 3 R

E - - -

T R E E E

• A U 4 4 3 Q * * *

• 5 5 7

• 2 E 1 1 3 D h f O t o M 4 1

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d r 6 3 * * * * * -

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• 2

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• 4 8 R E 1 2 4 1 2 4 9

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A E U E E E '

E E E T D Q R O '

7 3 3 2 6 iO 4 P H h 4 5

• 2

• 6 7 1 E t .

S 2 1 1 1 R U '4 1 1 L O A U U N ,ll ll

- I

• N T '

N N N *

- A O O R 0 2 3 2 O 1

- I C E T

I MSA T R '+ - - - + +

ETT A E E E E E E SNS- U E Q 3 9 1 BLUR 4 0 0 2 ALE 5 3 5 2

• 5 0

• SFW d 2 9 2 1

2 EOFO r 1

LPEP 3 BS AI DA TDIN UN EQA l TI SLH A T W R O

T I i i i

D N N C C C i

M 1C C i 1 i 1 i i i i i i 1 1 i i i i i i i i i C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C 1 i 1 i i i i i i i N U A

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E y r R d f n i m m m e S a 3 5 c 13 33 5 P E 3 3 e 11 3 D d 1 1 p 1 r I e S o

L f m C lv - -

( - - -

u U o n n r n n n l i a N s o o e o o o a t h p

s n n h n n n t i o r e e t e e e l

i D X X O X X X , T T A i I I II i I 1 . I'.* !i ' ! I! Ii i }I! i !iI  ! t

TABLE 3 t EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (7/84-12/84)

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) 6-MONTil EST. TOTAL

1. Type of Waste UNIT PERIOD ERROR, %

a. 3 Spent resins, filter sludges, evaporator m 1.07 E + 2 1.0 E+1 bottoms, etc. Ci 7.85 E + 2 1.0 E+1

b. Dry compressible waste, contaminated m 5 14 E + 2 10 E+1 equipment, etc. Ci 6 63 E + 1 10 E+1

c. Irradiated components, control m . E . E <

rods, etc. Ci . E . E

d. Other (describe) m 3 37 E + 1 10 E+1 SOLIDIFIED OIL Ci 2 22 E -1 10 E+1

2. Estimate of major nuclide compositon (by type of waste)

a. Cr 51 ~  % 5. 8 E-1 Mn 54  % 1.17 E O Co 58  % '9.9 -E O CO 60  % 1. 82 ' E + 1 Sb 122  % 9.8 E-1

Nb 95  % 3.5 E-1

! Cs 134  % 7,64 E O Cs 137  % 1. 4 7 E + 1 Ba 140  % 5.0 E-2 La 140  % 5.0 E-2 Ce 144  %  ?.1 E-1 Zr 95  % 1.7 E-1 Ru 103  % 3,0 E-2 Tc 99  % l.0 E-3 Sr 90  % 2.6 E-2 I 129  % 3.~ 0 E-6 C 14  % 7. 0 E-2 Fe 55  % 1. 7 7 E + 1 11 3  % 2. 34 E + 1 Ni 63  % 3.78 E G Pu 241  %- 1.9 E-1 Cm 242  % 1. 0 E-3 Pu 238 Pu 239 g 7, 0 E-4

% 9. 0 E-4 Pu 242  % 4.0 E-5 Am 241  % 5.0 E-4 On 244  % 1.0 E- 4

% . E

b. Cr 51  % 6.2 E-1 Mn 54  % 2.65 E O Co 58  % 4.72 E O Co 60  % 2. 25 E+ 1 Sb 122  % 1.4 E- 1 Nb 95  % 1.14 E O Cs 134  % 5.54 E O Cs 1374  % 1.92 E+ 1 Ce 144  % 1.8 E O Zr 95  % 5.1 E - l' Ru 103  % 1.9 E- 1

TABIZ 3-CGtTINUED PACE 2

2. Estimate of major nuclide composition (by type of waste)

Tc 99  % 1. 0 r-3 Sr 90 y 1. 6 v-2 I 129 2 8. 0 v-6 C 14 -

% 2, 2 p- 1 Fe 55 y 2.73p + 1 H 3 x 8.53 p 0 N1 63 <

Y 181 E O Pu 241  % 1.09 E O Cm 242  % 8, 0 E-2 Pu 238  % 6, 0 E-4 Pu 239  % 6. 0 E - 4 Pu 242  % 2. 0 E - 4 Am 241  % 7. 0 E-5 Cm 244  % 3. 0 E-4 e I , E

c.  % . E

% . E

d. Cr 51 *

% 3. 31 E + 1 Co 60  % 2. 66 E + 1 Cs 134  % 3. 87 E O Cs 137  % 2. 33 E + 1 Tc 99  % 5. 8 E-3 Sr 90  % 8. 3 E-2 I 129  % 1. 3 E-5 C 14  % 1. 7 E-1 Fe 55  % 1. 20 E + 1 Ni 63  % -5.36 E - 1 Pu 241  % 1. 7 E-1 Cm 242 g 1, 7 E3 Pu 235  % 9, 7 E-4 Pu 239  % 6. 9 E-4 Pu 242  % 1. 7 ' E - 4 Am 241  % 4. 8 E-4 Cm 244  % 4. 7 E-4

% . E

% . E

% . E

% . E

% . E

% . E

% . E

% . E'

% . E

% . E

% . E

% . E

% . E

% . E

% . E

% . E 3.' Solid Waste' Disposition NUMBER OF SilIPMENTS MODE OF TRMJSPORTATION DESTINATION 3 Private Vehicle Richland, WA 51 Private Vehicle Barnwell, SC B. IRRADIATED FUEL SilIPMENTS(Disposition)-

, NUMBER OF SIIIPMENTS MODE OF TRANSPORTATION DESTIN ATION NUN t.

..e Peg 3 5

\

, ATTACW%'T 2 ,

ANNUAL AND OUARTERLY DO9ES '

(1/84 - 12/84) m This attachment includes an assessment of radiation doses to the maximum exposed member of the public due to radioactive liouid

. and gaseous effluents released from the site.

m.

D l

l c- J

'4 TABLE 4 EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT 1984 ANNUAL DOSE ASSESS!iENT NORTH ANNA PO'4ER STATION LIQUID EFFLUENTS:

1st 2nd 3rd 4th Annual Units Quarter Quarter Quarter Quarter Total Total Body Dose mrem 0.93 0.86 1.11 1.15 4.04 Critical Organ Dose mrem 1.27 1.17 1.54 1.59 5.57 CASE 0US EFFLUENTS:

Units 1st 2nd 3rd 4th Annual Quarter Quarter Quarter Quarter Total Noble Gas Camma Dose mrad 0.50 0.51 0.27 0.005 1.28 Noble Gas Beta Dose mrad 0.83 0.62 0.57 0.003 2.02 Critical Organ Dose for I-131, H-3 particulates with T > 8 days mrem 0.69 2.95 0.10 0.008 3.75 4

4e 0

.m -

. . Pag 2 6 ATTACHMENT 3 (7/84-12/84)

REVISIONS TO OFFSITE DOSE CALCULATION MANUAL (ODCM)

As required by Technical Specification 6.15, revisions to the ODCM for the time period covered by this report are synopsized below. Supporting documentation and affected pages of the ODCM are attached.

10-11-84: Revisions were made to the ODQf to more clearly identify sampling stations and their relative positions to effluent release points.

H.P.-0DCM-13 Page 3 of 12 10-11-84 I

-Exposure P2thway Sample and/or Sample (Station) Number Location Airborne: (cont'd):

Particulate &

Radioiodine (cont'd) 21 air sampler is mounted on the power pole at the end of Rt.685 at the barricade, approx. 1.5 miles from the junction of Rt. 652 & 685 22 air sampler is mounted on the power pole approx. 0.8 miles from North Anna's Visitor's Center, West on Rt. 700, at the exclusion area boundary on the West side of Rt. 700 23 air sampler is mounted on the power pole approx. 25 feet from the shoreline, inside the exclusion area boundary, located on lot #34

( on Carr Circle in the Aspen Hill subdivision 24 (control) air sampler is mounted on top of the Orange Va Region-al Vepco Office Building approx. 1 mile South of the junction of Rt. 20 & 15 on the right Soil 1,3,4,5,6,7 Previously Identified 21,22,23,24 Direct:

,: hbTI.Ds '

~

N-1/33 are located at Bearing Cool-ing Tower on stairwell structure approximately 150 feet from Vepco security building on Vepco North Anna Power Station site N-2/34 are located on power pole

, approximately - 200 feet from entrance of Sturgeon's Creek Marina off Rt. 208 i

H.P.-0DCM-13 Page 4 of 12 10-11-84 Exposure Pathway Sample and/or Sample (Station) Number Location Direct (cont'd):

TLDs (cont'd):

NNE-3/35 are located on Vepco North Anna Power Station site at rain gauge in construction parking lot "C" near weather tower NNE-4/36 are located on power pole at Good Hope Church at the intersection of Rt. 208 and Rt. 601.

NE-5/37 are located on weather tower across from security build-ing and behind Vepco's park-ing lot "B" NE-6/38 are located off Rt. 601 on Rt. 713 at Lake Anna Marina entrance on power pole e ENE-7/39 are located on island across from training facility on weather tower fence ENE-8/40 are located on power pole at entrance of Rt. 689 off Rt.

601 on left E-9/41 are located on the island across from the training facility on the second power pole on the right headed toward the recreation area E-10/42 are located on 500 Kv Vepco power line off Rt. 601 at

" Morning Glory Hill": near home of J. R. Humphries ESE-11/43 are located on island after crossing dike from Vepco "A" parking lot on first power pole approximately 0.1 mile

. from island entrance ESE-12/44 are located on a power pole on Rt. 622 off Rt. 601 across the road f rom R. L.

Garlic's home, approximately 25 feet from the road i

l' H.P.-0DCM-13 s

Page 5 of 12 10-11-84 '

Exposure Pathway Sample and/or Sample (Station) Number Location i

Direct (cont'd):

C' -

TLDs (cont'd):

SE-13/45 are located on island across from Vepco training facility at Vepco Biology Lab en-trance gate i

SE-14/46 are located at entrance to Dam off Rt. 601 at inter-section with Rt. 701 SSE-15/47 are located on power pole approximately 25 feet fron the shoreline on Lot #34 on Carr Circle in the Aspen Hill subdivision SSE-16/48 are located on. power pole at intersection of Rt. 614 and Rt. 652 at Elk Creek

,S-17/49 are located on entrance to

(

warehouse compound on gate 0.35 miles from Vepco main security building e

S-18/50 are located on power pole at Rt. 614 intersection with Rt. 652 approximately 1/2 mile from Elk Creek Church

, SSW-19/51 are located on light pole along Vepco site access road approximately 0.7 mile from Vepco main security building a

SSW-20/52 are located on a power pole approximately 200 yards from the fork of Rt. 618 and Rt.

614 on Rt.) 618 North side approximately 100 feet from the road i

SW-21/53 are located on light pole on Vepco North Anna Power Station site access road approximately 0.85 miles from Vepco main Security Building

, H.P.-0DCM-13 Page 6 of 12 s

10-11-84 Exposure Pathway Sample and/or Sample (Station) Number Location ect (cont'd):

M. LDs (cont'd):

SW-22/54 are located on power frole

' approximately 2.5 miles from intersection of Rt. 652 and Rt. 700 exiting North Anna Power Station at abandoned house and dumpster site WSW-23/55 are located on first power line (500 Kv) upon entering Vepco, North Anna Power Station site to right off Rt. 700 ,

WSW-24/56 are located'on a power pole approximately 0.8 miles from '

North' Anna Power Station Visitor's Center, West on f

Rt. 700 W-25/57 are located at North Anna

( Power Station radio tower approximately 0.15 East on Rt. 700 from Visitor's Center entrance W-26/58 are located on power pole on Rt. 685 off Rt. 652 approxi-mately 1.0 miles from inter-section at abandoned trailer WNW-27/59 are located at power pole at the end of Rt. 685 at barricade, approximately 1.5 miles from the junction of Rt. 652 and Rt. 685 WNW-28/60 are located on a power pole at the end of H. Purcells' private road off Rt. 685 NW-29/61 are located past Unit I and 2 intake at end of road on f

transformer fence - on North Anna Power Station site NW-30/62 are located on a power pole at Lake Anna Campground on Rt. 208 o - - - .

H.P.-0DCM-13 Page 7 of 12 s

10-11-84 Exposure Pathway Sample and/or Sample (Station) Number Location

. Direct (cont'd):

TLDs (cont'd):

NNW-31/63 are located on fence near Unit 1 and 2 intake struc-ture on North Anna Power Station site NNV-32/64 are located approximately 3.3 miles from intersection of Rt. 652 and 208 going East on light point at Sam Hairfield and Bro. Store on Rt. 208 C-1/2 are located on a power pole approximately 175 feet be-hind the Bumpass Post Office at the fork of Rt.601 and 618 C-3/4 are located - on the top of the Orange Va Regional Vepco

( Office Building approxi-mately 1 mile South of the junction of Rt. 20 and Rt.

15 on the right C-5/6 are located in Mineral on a power pole approximagely 200 feet off Rt. 618 on Albermale Street behind the house on the corner C-7/8 are located on power pole at Glen Marye Shopping Center

,, in Louisa Water Borne: 1,2,3,4,5,5A,6,7, Previously Identified 21,22,23,24

a. Surface 8 sampled from 2nd lagoon on discharge 9 sampled from Lake near Rt.

208 bridge 11 sampled from North , Anna river just below the lake dam

b. Ground 1A sampled from well at North Anna Power Station's Biology Lab on island across from discharge canal

i II . P . -0DCM- 13 P:g2 9 cf 12 10-11-84 ,

~

2.0 SAtlPLING STATION IDENTIFICATION AND RELATIVE POSITION TO EFFLUENT RELEASE POINT STATION DISTANCE DEGREES AND NUtillER LOCATION HILES C0t! PASS DIRECTION ret 1 ARKS 1 On Site 0.2 42 NE 3 tlineral 7.1 243 WSW 4 Wares Crossroads 5.1 287 WNW 5 Good flope Church 4.2 20 NNE 6 Levy 4.7 115 ESE 7 Bumpass 7.3 167 SSE 8 Discharge Lagoons 1.1 148 SSE 9 Lake Anna, upstream 2.2 320 NW Control 1I North Anna River, downstream 5.8 128 SSE 12 llolladay Dairy 8.3 310 NW 13 Fredericks llall 5.6 205 SSW 14 Route 713 1.2 43 NNE ,

15 Route 614 1.7 133 ESE 16 Route 629/522 12.6 314 NW Control 21 Route 685 1.0 301 WNW 22 Route 700 1.0 242 WSW 23 Aspen Ilills 0.9 158 SSE

II . P. -0DCM- 13 Page 10 of 12

l. 10-11-84 .

2.0 SAMPLING STATION IDENTIFICATION ANO RELATIVE POSITION TO EFFLUENT RELEASE POINT (cont.)

u STATION DISTANCE DECREES AND NUtillER LOCATION HILES COMPASS DIRECTION REMARKS 24 Orange 22.0 -325 NW Control e SA Sturgeon's Creek 3.2 11 NNE 2 Fredericks IIall 5.3 225 SSW N-1/33 On Site (Bearing Cooling Tower) 0.06 10 N N-2/34 Sturgeon's Creek 3.20 11 N NNE-3/35 On Site (Parking Lot "C") 0.25 32 NNE NNE-4/36 Rt. 208 (Good flope Church) 4.96 25 NNE NE-5/37 On Site (Parking Lot "B") 0.20 42 NE NE-6/38 Rt. 713 (Lake Anna Marina) 1.49 34 NE ENE-7/39 On Site (Weather Tower Fence) 0.36 74 ENE ENE-8/40 Rt. 689 2.43 65 ENE E-9/41 On Site (Near Training Facility) 0.30 91 E E-10/42 " Morning Glory liill" 2.85 93 E

! ESE-ll/43 On Site (Island Dike) 0.12 103 ESE ESE-12/44 Rt. 622 4.75 115 ESE SE-13/45 On' Site (Vepco Biology Lab) 0.75 138 SE SE-14/46 Rt. 701 (Dam Entrance) 5.88 137 SE

/

}llgPh. ii.P.-0DCM-13

"'. Page 11 of 12 10-11-84 -

2.0 SAMPLING STATION IDENTIFICATION AND RELATIVE POSITION TO EFFLUENT RELEASE POINT (cont.)

STATION DISTANCE DEGREES AND NtitlBER LOCATION MILES COMPASS DIRECTION REtlARKS SSE-15/47 Aspen 11111 0.93 158 SSE SSE-16/48 Elk Creek 2.33 165 SSE S-17/49 On Site (Warehouse Compound Gate) 0.22 173 S S-18/50 Rt. 690 (Elk Creek Church) 1.55 178 S SSW-19/51 On Site (Access Rd.) 0.36 197 SSW SSW-20/52 Rt. 618 5.30 205 SSW SW-21/53 On Site (Access Rd.) 0.30 218 SW SW-22/54 Rt. 700 4.36 232 SW WSW-23/55 On Site (500KV Tower) 0.40 237 WSW WSW-24/56 On Site (Rt. 700) 1.00 242 WSW W-25/57 On Site (Radio Tower) 0.31 279 W W-26/58 Rt. 685 1.55 274 W WNW-27/59 End of Rt. 685 1.00 301 WNW WNW-28/60 II. Purcell's Private Road 1.52 303 WNW NW-29/61 On Site (End of #1 & #2 Intake) 0.15 321 NW NW-30/62 Lake Anna Campground 2.54 319 NW

k h .P.-0DCH-13

1. - Page 12 of 12 10-11-84 ,

2.0 SAMPLING STATION IDENTIFICATION AND RELATIVE POSITION TO EFFLUENT RELEASE POINT (cont.)

STATION DISTANCE DEGREES AND NUtillER LOCATION HILES COMPASS DIRECTION REMARKS NNW-31/63 On Site (#1 & #2 Intake) 0.07 349 NNW NNW-32/64 Rt. 208 3.43 344 NNW C-1/2 Bumpass Post Office 7.30 167 SSE CONTROL C-3/4 O rac.ge 22.00 325 NW CONTROL C-5/6 Nineral 7.10 243 WSW CONTROL C-7/8 Louisa 11.54 257 WSW CONTROL

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HP-0CDM-13 Page 4 of 7 10-11-84

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PKga 7 j ATTACHMENT 4 (7/84-12/84)

REVISIONS TO PROCESS CONTROL PROGRAM (PCP)

As required by Technical Specification 6.14, revisions to the PCP

- for the time period covered by this report are synopsized below, Supporting documentation and affected pages of the PCP are attached.

12-27-84: The PCP was revised to incorporate radwaste solidifiention system operating procedures, by i

reference, and to reflect the guidance of NRC Generic Letter 84-12, Compliance with 10 CFR Part 61 and Implementation of the Radiological 2 Effluent Technical Specifications (RETS) and Attendant Process Control Program (PCP).

I i

SERIAL i J 7 C DISTRIBUTION LIST - NRC CORPJSPOCENCE J. M. DAVIS

?

W. R. BENTRALL b*T R. F. DRISCOLL p g'8 A. L. BOCC. JR.

J. O. EAS W OOD

• • cP o nrRS**Y M."k

, f L k fu

- u A t, 5, eca R.nr krr1 *.-

a,IL veb -m

+4 'o,

, UNITED STATES

! n

.I NUCLEAR REGULATORY COMMISSION

i. c wAssancrow. o. c. rosss

• • "t

/ April 30,1984 ROTS MAY 101984 J.O.E.

TO ALL OPERATING REACTORS AND APPLICANTS FOR OPERATING LICENSES Gentlemen:

SUBJECT:

COMPLIANCE WITH 10 CFR PART 61 AND IMPLEMENTATION OF THE RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS) AND ATTENDANT PROCESS CONTROL PROGRAM (PCP) (GENERIC LETTER 84-12 l

l l

l

1. UNJTED STATES

! o -

NUCLEAR REGULATORY COMMISSION i .I wassinc row. o. c. rosss o t i

\ **"? ,o# April 30,1984 Il0TED W6' 1019M J.01 TO ALL OPERATING REACTORS AND APPLICANTS FOR OPERATING LICENSES Gentiemen:

SUBJECT:

COMPLIANCE WITH 10 CFR PART 61 AND IMPLEMENTATION OF THE RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS) AND ATTENDANT PROCESS CONTROL PROGRAM (PCP) (GENERIC LETTER 84 This letter is to inform you that the waste manifest provisions of 10 CFR

' 20.311 became effective on December 27, 1983. The manifest system is closely related to certain requirements of 10 CFR Part 61 that place new requirements on classification and acceptable forms for low-level radioactive wastes being 4

shipped from comercial nuclear power plants to commercial disposal facilities.

The NRC staff has been made aware of the fact that neither the states nor the disposal facility operators currently have sufficient resources to assure that all incoming low-level radioactive waste is in compliance with these new regulations. Consecuently, the NRC has been asked to provide reasonable assurance Part 61.

that its licensees are complying with all applicable provisions of During the development of Part 61, the NRC staff determined that compliance with the radioactive waste form requirements of Part 61 and the certification requirements of 10 CFR 20.311 could be achieved by the develop-ment and use of a Process Control Program (PCP) as an attendant part of the licensee's Radiological Effluent Technical Specifications (RETS). This approach was determined to be acceptable by the responsible state regulatory 4

agencies that license the disposal sites. It is now apparent, however, that  ;

many licensees do not yet have approved PCPs and that no licensee has a PCP which specifically addresses the new requirements of Part 61.

As an interim measure, the responsible state regulatory agencies and the disposal site operators have agreed to continue to accept nuclear power plant

- low-level radioactive wastes based upon the -NRC staff's assurance that reasonable progress is being made toward demonstration of full compliance with new requiremants of Part 61 and Part 20.

able to offer such assurances for those plants for which there are NRCThe NRC s approved and implemented RETS and the attendant PCP';. The NRC staff will assume a good-faith effort on the part of these licensees to modify in a timely fashion the PCPs to accommodate all new and applicable Part 61 and Part 20 requirements. We are prepared to assist, when requested, those licensees which presently have approved PCPs to assure that they are upgraded

! to meet the new requirements of Part 61; however, the NRC staff cannot offer the same type of assurances for those operating plants which do not possess currently approved RETS and PCPs. Prompt action may be necessary if radio-i active waste shipments from these plants are to continue without interruption.

l 1

8405010084 .

. The NRC staff will make every effort to avoid any interruption of low-level radioactive waste shipments by its licensees. We a.e prepared to expedite the implementation of NRC approved RETS and PCPs for 'all licensees who request assistance.

If you have any questions concerning this subject, please contact either

,W. Gamill or F. Congel via your Project Manager.

h.

f G. Eisenh)utLk rr D ector Division of'.icensing Office of Nuclear Reactor Regulation

i

, Process Control Program Page 3 of 19 12-27-84 i

1.0 SCOPE

1.1 Purpose i he purpose of the North Anna Nuclear Power Station Process Control Iwr'. - Program is to ensure that

a. Solidified liquid wastes, dewatered resins, and aqueous filter

, media are packaged in the proper container based on their activity level;

b. Dewatered resins and aqueous filter media are packaged with no

detectable free standing liquid prior to transportation; b c. Liquid wastes are solidified and contain no detectable free standing liquid; and i 5.!

d. Resultant solidified wastes are in compliance with the provisions

, E# '

of 10 CFR Part 61.

1.2 Applicability This Process Control Program (PCP) shall be implemented by all ,

personnel who operate. dewatering equipment, operate solidification equipment, package spent filter cartridges, collect and process samples

[ used to verify conditions required by this Program, and prepare documentation for shipping of radioactive waste.

2.0 SYSTEM DESCRIPTION 2.1 Dewatering System l Two sources of spent resins for dewatering and disposal exist at the ..

North Anna Station. The first is the Spent Resin Holdup Tank which collects for decay end disposal spent resins from the primary coolant purification system. These resins are dewatered in disposable high integrity containers. The second is the M Level Liquid Waste

, Treatment System. This system utilizes resins which are received on-site in disposable containers. When the resins in these contait.ars are chemically exhausted or the activity level of the resins approaches the limit for the container, the resins are devatered. The following _,

sections provide a description for each dewatering system.

Process Control Program s Page 5 of 19 12-27-84 l

2.1.1.e Shipping Cask 1 i

The shipping cask can be any licensed shipping cask, designed for use with the high integrity container  ;

utilized, which provides adequate radiation shielding and l package integrity for transportation of dewatered spent '

resins.

. h 2.1.1.f Dewatering Pump The dewatering pump is an air driven, 1-1/2" " Sandpiper" pump or equivalent.

2.1.1.g TV Monitor The TV monitor is a remote display television utilized, when required, to monitor the container filling operation. This monitor provides information to the operator on container resin and water level and resin slurry consistency.

2.1.1.h Interconnecting Hose and Piping The resin transfer line used to sluice resin from the "

primary coolant purification demineralizers to the Spent Resin Holdup Tank, as well as connections from the tank l co the Spent Resin Transfer Pump. Spent Resin Recirculation Pump, and plant ventilation system are permanently installed stainless steel piping. The connection from the Spent Resin Transfer Pump to the dewatering container, the connection from the dewatering container to the dewatering pump, and the connection from the dewatering pump back to the Spent Resin Holdup Tank

.. are all flexible rubber hose. -

• .. 2.1.1.1 valves Valves are installed in the system, as required, to select flow paths and isolate portions of the dewatering system as may be necessary.

l l

l 1

l l

i Process Control Program Page 6 of 19 j- ~

12-27-84 t

2.1.1.j Instrumentation and Controls Controls are provided to remotely start and stop pumps, remotely open and close valves, remotely monitor the container fill-head level and influent flow, and remotely indicate the container fill and discharge flow.

, 2.1.1.k Sample Taps As an example, a sample tap is provided on the discharge side of the Spent Resin Recirculation Pump.

2.1.2 Dewatering System for Resins Received In Disposable Containers The High Level Liquid Waste (HLLW) System contains from two to seven disposable filter /demineralizer vessels. The dewatering system for the HLLW disposable filter /demineralizer vessels i consists of the disposable vessel and the dewatering pump.

2.1.2.a Disposable Vessel The disposable filter /demineralizer vessel can be any 1

container designed for use as a disposable ~

filter /demineralizer vessel, with provisions for dewatering.

2.1.2.b Devatering Pump The dewatering pump is an air driven, 1-1/2" " Sandpiper"

W

_ pump or equivalent.

2.1.3 Dewatering System For Filter Cartridges The High Level Liquid Waste Filter utilizes disposable cartridge filters. The cartridges are air dried prior to disposal, i therefore no dewatering system or_ equipment is needed nor exists gg .,,,, for devatering spent filter cartridges.

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.- 2.2. Solidification System There is no permanently installed, operating solidification system on _

site. In the event contractor services are utilized for solidification, the contractor will be required to provide to VEPCO a e

Process Control Program Page 7 of 19 12-27-84

• b complete description of the solidification system to be used. The system description provided will be subject to VEPCO review and acceptance and once accepted, will be incorporated by reference into this Process Control Program. The solidification system will not be operable until the system description is accepted by VEPCO.

3.0 CHARACTERISTICS OF WASTE FEEDS 3.1 Dewatered Resins Resins to be devatered are either sluiced to the disposable dewatering high integrity container or are contained in a disposable container.

Low activity resin may be shipped in a carbon steel liner.

3.2 Filter Elements Spent filter elements are removed from the filter vessel housing and are processed as individual units.

3.3 Liquids For Solidification Presently liquid wastes are not fed to a solidification system. In the event contractor services are utilized for solidification, the waste ,

stream characterization will be provided by either:

( a. VEPCO personnel collecting and analyzing samples in accordance with station approved procedures; or

b. the contractor collecting and analyzing samples in accordance with station approved procedures.

The procedures will consider the solidification process utilized and will analyze for parameters or constituents which may affect the -

solidification process. Sample collection and analysis procedures will be subject to VEPCO review and acceptance and once accepted, will be incorporated by reference into this Process Control Program. The solidification system will not be operable until the sample collection and analysis procedures are accepted by VEPCO.

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Process Control Program Page 8 of 19 12-27-84 4.0 SYSTEM OPEPATION 4.1 Dewatering

4.1.1 Dewatering Resins from the Spent Resin Holdup Tank Spent resins from the primary loop cleanup system are sluiced to l the Spent Resin Holdup Tank for decay and storage. The spent resin s torage time in the tank is maximized to the extent possible to allow for maximum radioactivity decay. During the period of resin storage, the tank contents are periodically mixed using the Spent Resin Recirculation Pump to ' prevent excessive settling and resin packing. Prior to resin transfer to the dewatering container, the tank contents will be mixed. A sample for isotopic analysis may be taken from the sample tap on the recirculation line. Sample requirements are defined in Section 5.0. If a sample is taken from the recirculation line, a portion of the recirculating slurry is drawn into a sample container, the

! sample container contents decanted, and an isotopic analysis performed in accordance with approved procedures in the station Health Physics Procedures Manual, Section 3.

Following mixing and sampling (if performed), the radwaste ,

metering pump is used to transfer resin slurry to the disposable dewatering container. The Spent Resin Recirculation Pump will be

( operating during resin transfer. The operator fills the

, container with slurry until the container has been filled.

This is verified by the remote TV monitor. When the container is filled, the transfer of slurry to the container is stopped. The excess water is then removed by the dewatering pump. This water is transferred back to the Spent Resin Holdup Tank. After the excess water is removed, slurry is again transferred to the container until the container is filled with the resin and water mixture. The container is again drained using the dewatering pump. This process continues until the container is filled with resin (i.e., sluicing water removed). The resin-filled container is dewatered in accordance with one of the following approved station dewatering procedures:

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VEPCO Procedure 1-OP-20.2 for dewatering disposable I containers.

Chem-Nuclear Systems Inc. procedure FO-0P-003,

" Dewatering Procedure for CNSI Conical-bottom High Integrity Containers Containing Bead-Type Ion Exchange Resin, 1% Free-Standing Water."

Each procedure specifies a series of minimum periods for devatering pump operation and shutdown and provides a minimum time period for sample collection. These procedures are not included as they are considered proprietary. The procedures have been tested for compliance in dewatering to less than the specified percentage as required by Chem-Nuclear Systems. Inc.,

., State of South Carolina Radwaste Material License No. 97.

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.? The approved procedures must be strictly followed and the documentation required by the above procedures must be completed.

4.1.2 Dewatering Resins In Disposable Filter /Demineralizer Vessels-Spent resins from the High Level Liquid Waste System are contained in disposable filter /demineralizer vessels. The .

vessels may be removed from service when either calculations indicate the resins contain close to I uCi/gm radioactivity or

( the resins have insufficient ion exchange capacity remaining for adequate liquid processing. Once removed from service, the vessels are drained and dewatered in accordance with one of the following procedures. The procedure selected will be determined by the container to be dewatered.

VEPCO Procedure 1-0P-20.2 for dewatering disposable containers. -

Chem-Nuclear Systems Inc. procedure 06601-27-01,

" Dewatering Procedure for the Annular L14-195 and L21-300 Demineralizer Liners, 1% FSW."

Process Control Prograa

, Page 10 of 19 12-27-84 Chem-Nuclear Systems Inc. procedure FO-0P-005, '

" Dewatering Procedure for the 10 Cu. Ft. Filtration Unit Containing Ion Exchange Resins, 1% FSW." i l

• l Chem-Nuclear Systems Inc. procedure FO-0P-001,." Dewatering  !

Procedure for the 24-inch Diameter Pressure Demineralizer Vessel Containing Activated Carbon. 0.5% Free-Standing Water."

Chem-Nuclear Systems Inc. procedure FO-OP-004

" Dewatering Procedure for the 24-inch Diameter Pressure Domineralizer Vessel Containing Ion Exchange Resins. 0.5%

FSW."

Chem-Nuclear Systems Inc. procedure FO-OP-007,

" Dewatering Procedure for the L14-195 .and 14-170 Conical-Botton Demineralizer Vessels, 0.5% Free-Standing 3

Water."

Each procedure specifies a series of minimum periods for dewatering pump operation and shutdown and provides a miniaua time period for sample collection. These procedures are not included as they are considered proprietary. The procedures have been tested for compliance in dewatering to less than the ,

specified percentage as required by Chen-Nuclear Systems. Inc.,

( State of South Carolina Radwaste Material License No. 97.

M The above procedure (s) must be strictly followed and the documentation required by the above procedures must be completed.

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' Page 11 of 19 12-27-84 4

4.2 Spent Filter Cartridge Disposal When used, High Level Liquid Waste cartridge filters are to be removed from service, the vessels drained, and the elements. removed. [ Note:

The wound-type elements used do not contain void spaces which can trap j liquid in pockets.] In accordance with Health Physics Department  ;

Procedures, a sample of the element will be removed for radioassay

, following removal of the element from the filter vessel. The eleuent will be allowed to air dry. The element will be considered to contain less than 0.5% free-standing water when it can be placed on a clean, dry plastic sheet for a minimum of four hours and, when removed, the plastic contains no free liquid. If free liquid remains on the

, plastic, the filter element will continue to be placed on a dry piece l of plastic for four hour increments until, when the element is removed from the plastic, no free liquid remaine on the plastic. Drying of the elements may be assisted by placing the elements in a flow of warm, dry air or any other means which meets station HP procedures for control of airborne particulates. However, the four-hour dryness verification shall occur with no drying air flow present. Contractors may use j

mechanical filters for this service that would be controlled by contractor procedures.

4.3 Solidification There is no permanently installed, operating solidification system on

( site. In the event contractor services are utilized for solidification, the contractor will be required to provide to VEPCO system operating procedures. The system operating procedures must i

define the process control parameters which assure the proper i

proportioning and mixing of wasta and solidification agents. The operating procedures shall also specify minimum data logging requirements to document system operation within the specified ranges.

i The system operating procedures will be subject to VEPCO review and -

acceptance and once accepted, will be incorporated by reference into this Process Control Program. The solidification system will not be

,, operable until the system operating procedures are accepted by VEPCO.

d%-2 .3 Presently accepted procedures for use are:

Chem-Nuclear Systems, Inc. procedure SD-OP-027 " Operating i

Procedure for CNSI Portable Cement /0il Solidification-Unit No. 1 (PSU/0il-C-1)"; _

@* Chem-Nuclear Systems, Inc. procedure SD-OP-026, " Process Control 1

Program for Cement /0il Solidification"; and rs .

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• Chem-Nuclear Systems, Inc. procedure SD-OP-003, " Process Control Program for CNSI Cement Solidification Units." -

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Procesa Control Program Page 12 of 19 i i -

12-27-84 l j

5.0 COLLECTION AND ANALYSIS OF SAMPLES 5.1 General Requirements 5,1.1 Definitions 5.1.1.a Batch - The amount of waste which fills one disposable i liner or drum.

7 M 5.1.1.b Transfer - The delivery of liquid or sluiced radioactive a vaste to a solidification or disposable container. There may be several transfers made from several sources that compose a batch. If the transfers are made from the same source and the contents of the source are known not to change during the time of the transfers, the volume of each transfer need not be known. If the transfers are made from several different sources, the volume of each transfer must be known. If several transfers are made from one source which has inputs to that source during the duration of the transfers, the volume of each transfer must be known.

5.1.1.c Sample - A portion of a transfer or batch that represents ,'

the contents of that transfer or batch.

( 5.1.1.d Composite - 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.

5.1.1.e Direct In-Container Sample -A sample removed from the disposable container after the container has been filled; the contents of which are a composite of samples taken from various positions within the container.

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Process Control Program s

• Page 13 of 19 12-27-84 5.1.2 Frecuency of Samples 5.1.2.a Dewatered Resins - A sample for radioassay purposes shsll be taken from each batch. The sample may be a representative sample taken at the source (when the source remains unchanged during the transfer of c batch),

a composite sample composed of samples collected at the source, or a direct in-container sample.

5.1.2.b Spent Filter Cartridges A sample of each spent filter cartridge shall be obtained for radioassay purposes.

, , , .1.2.c Solidification A sample of at least every tenth batch of each type of liquid or sluiced radioactive waste (e.g., boric acid solutions, spent resins, evaporator bottoms) shall be used to demonstrate solidification.

If any test specimen fails to solidify, the batch under test shall be suspended until such time as additional test specimens tan be obtained, alternative solidification parameters can be determined in accordance

(' vith the procedures incorporated by reference into this Process Control Program, and a subsequent test verifies solidification. Solidification of a batch may then be resumed using the alternate solidification parameters determined.

If the initial test specimen from a batch of waste fails to verify solidification, then representative test specimens shall be collected from each successive batch of the same type of waste until three (3) consecutive

' initial test specimens demonstrate solidification. The l 3

operating procedure incorporated by reference into this Process Control Program shall be modified, as required, to assure solidification of subsequent batches of waste.

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l Process Control Program

. Page 16 of 19 12-27-84 5.3.3 Acceptance Criteria l The results of the radioassay are considered acceptable when it has been verified and documented that the spent filter cartridge is packaged 8.n a container which is acceptable for transportation and burial, considering the radioactivity concentrations which exist in the waste. Standard liners and drums are acceptable for waste containing less than 1 pCi/gm activity and High Integrity Containers are required for waste containing more than 1 pCi/gm activity, k h 5.4 Liquids for Solidification There is no permanently installed, operating solidification system on-site. In the event contractor services are utilized for ,

solidification, the contractor or VEPCO shall provide procedures for l sample collection and sample analysis. The sample collection and l sample analysis procedures shall be for both radioactivity determinations and solidification verification. These procedures will be subject to VEPCO review and acceptance and once accepted, will be incorporated into this Process Control Program by reference (See Section 4.3). The solidification system will not be operable until the sample collection and analysis procedures are accepted by VEPCO. ,

All chemicals used to condition or solidify waste or simulated waste in

( solidification tests shall be representative of the actual chemicals to i be used in full scale solidification. If chemicals of a different type l or from a different manufacturer are used, the new material shall be '

tested to verify it produces a solid product prior to full scale solidification.

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Process Control Program

. Page 19 of 19 12-27-84 7.1.3 Physical Interfaces 7.1.3.a Process Fluids

1. High Level Liquid Waste Tank Discharge: 5-25 gpm at 30-150 psig; maximum temperature-125F; 1-1/2", 150 psi hose with Kam-Lock fitting.

2. Filter Atmospheric Vent Hose: 1/2" tygon hose connected to a 1/2" needle valve (150 psi rating) connected to a Kam-Lock fitting.

3. Demineralized Water Discharge: 5-25 gpa at 150 psig; 1-1/2", 150 psi hose with Kam-Lock fitting.

4. Resin Transfer: 1-1/2", 150 psi hose; 25 spm max at 150 psi max; quick disconnect fitting at disposable liner fill head.

5. Dewatering Line: 1-1/2" hose connection at disposable liner fillhead; quick disconnect fitting.

6. Crane Services for loading and unloading vessels. ,

7. Protective clothing and dosimetry devices for

( Chem-Nuclear Systems Inc. operators.

8. Shielding.

7.2 Filter Cartridges The utility is responsible for removing the cartridges from the filter housing. All actions following cartridge removal are utility actions, therefore, there are no additional interfaces.

i ,7.3 Solidification There is no permanently installed, cperating solidification system on site. In the event contractor services are utilized for solidification, the contractor will be required to provide a list of physical interfaces, services required, and breakdown of ,

utility / contractor responsibilities. These documents will be subject to VEPCO review and acceptance and once accepted, will be incorporated into this Process Control Program by reference (See Section 4.3). The solidification system will not be operable until the definition of physical interfaces, services required, and utility / contractor ;

responsibilities are accepted by VEPCO.

. + p Paga 8 ATTACHMENT 5 (7/84-12/84)

MAJOR CHANGES TO RADIOACTIVE SOLID WASTE TREATMENT SYSTEMS No major changes to the radioactive solid waste treatment systems were made for this reporting period, i

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- " e Pzg2 9 ATTAC10fENT 6 (7/84-12/84)

RADIOACTIVE LIQUID AND CASEOUS EFFLUENT MONITORING INSTRUMENTATION INOPERABLE As required by Technical Specification 3.3.3.10.b and 3.3.3.11.b, an account of inoperable radioactive. effluent monitors for the time period covered by this report is provided below.

On July 2, 1984, RMSW-108 (radiation monitor for Service Water discharged to Lake Anna) was declared inoperable. Repair parts were ordered from the manufacturer, the monitor repaired, and returned to service on August 7, 1984. The monitor was inoperable greater than 30 days because of the need to order repair parts from the manufacturer, l

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• " ' Paga 10 ATTACHMENT 7 (7/84-12/84)

UNPLANNED RELEASES No unplanned releases, as defined according to the criteria presented in 10 CFR Part 50.73, occurred during the time period covered by this report'.

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1 tr:notL Snur Ntxiter Cpmstasikpartment VrceIhulent itzt($eikrrXtX:6 Nudtr Cpmaio:ts o,,e p w .rlt a Ridxront 11,gnxs 23261 L

March 1, 1985 /l/g ,.

VIRGINIA POWER Dr. J. Nelson Grace Serial No.85-141 Regional Administrator N0/JHL:acm ys.

U. S. Nuclear Regulatory Commission Docket Nos. 50-338' g Region II 50;339 101 Marietta Street, Suite 2900 License Nos. NPF-4 Atlanta, Georgia 30323 NPF-7

Dear Dr. Grace:

Enclosed is the Radioactive Effluent Release Report for North Anna Power Station for the period July 1,1984 to December 31, 1984.

Very truly yours,,

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W. L. Stewart Enclosure (2 copies) cc: Mr. James R. Miller, Chief Operating Reactors Branch No. 3 Division of Licensing Mr. M. W. Branch NRC Res. dent Inspector North Anna Power Station l

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