ML17286A281

Washington Nuclear Plant 2 Semiannual Effluent Release Rept,Jan-June 1990. W/900828 Ltr

ML17286A281
Person / Time
Site:	Columbia
Issue date:	06/30/1990
From:	John Baker WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GO2-90-144, NUDOCS 9009070188
Download: ML17286A281 (47)

Text

ACCELERATED DIS UTION DEMONST TION SYSTEM

'J V REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9009070188 DOC.DATE: NOTARIZED: NO DOCKET g FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION BAKER,J.W. .Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION R

Effluent Release Rept Jan-June

SUBJECT:

"WNP-2 S 1990."

ann

/900828 l Radioactive r.

DISTRIBUTION CODE: COPIES RECEIVED:LTR ENCL SIZE: D TITLE: 50.36a(a)(2) Semiannual Effluent Release Reports NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 3 3 PD5 PD 1 1 ENG,P.L. 1 1 D INTERNAL: ACRS 1 1 AEOD/DS P/TPAB 1 1 "D IRM TEC 1 1 NRR/DREP/PRPB11 2 2 G E 0 1 1 RES BROOKS I B 1 1 RGN5 DRSS/RPB 2 2 RGN5 FILE 02 1 1 EX'TERNAL NRC PDR I

BNL T CHLER I J 0 3 1 1 EGGG SIMPSON,F 2 2 1 1 D

D D

NOTE TO ALL",RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 19 ENCL 19

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aa WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352 August 28, 1990 G02-90-144 Docket No. 50-397 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

NUCLEAR PLANT NO. 2, OPERATING LICENSE NPF-21 SEMI-ANNUAL EFFLUENT REPORT JANUARY 1, 1990 - JUNE 30, 1990 In accordance with the Tile 10 of the Code of Federal Regulations, Part 50.36a (a)(2), the subject report is herewith being submitted.

Should you have any question, please contact Mr. R. G. Graybeal, Manager, WNP-2 Health Physics Chemistry.

Very truly yours,

. W. Baker WNP-2 Plant Manager bk cc: JB Martin - NRC PL Eng - NRC DL Williams - BPA/399 NRC Site Inspector - 901A CR Wallis - EFSEC D Shrman - Amer. Nuclear Insurers TR Strong - DSHS 90090701SS 900630 l3 r PDR ADOCK 05000397 PDC

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Reference:

10CFR50.36a(a)(2)

WNP-2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT REPORTING PERIOD JANUARY THROUGH JUNE 1990 WASHINGTON PUBLIC POWER SUPPLY SYSTEM LICENSE NO. NPF-21

II 1

TABLE OF CONTENTS

1.0 INTRODUCTION

. 1 2.0 LIQUID EFFLUENTS 1 3.0 GASEOUS EFFLUENTS 5 4.0 SOLID RADIOACTIVE WASTE 18 5.0 METEOROLOGICAL DATA 24 6.0 DOSE ASSESSMENT IMPACT ON MAN 25 7.0 REVISIONS TO THE ODCM 26 8.0 REVISIONS TO THE PROCESS CONTROL PROGRAM (PCP) 27 9.0 NEW OR DELETED LOCATIONS FOR DOSE ASSESSMENTS AND/OR ENVIRONMENTAL MONITORING LOCATIONS. 28 10.0 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS. 29

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LIST OF TABLES 2-1 WNP-2 LIQUID EFFLUENTS SUMMATION OF ALL RELEASES JANUARY JUNE 1990 . ~ ~ ~ 2 2-2 WNP-2 LIQUID EFFLUENTS SOURCE TERHS-JANUARY JUNE 1990. ~ ~ ~ 3 3-1 WNP-2 GASEOUS EFFLUENTS SOURCE TERMS-HIXED MODE RELEASES HAIN PLANT VENT JANUARY JUNE 1990. ~ ~ ~ 8 WNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES TURBINE BUILDING JANUARY JUNE 1990 3-3 WNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL, RELEASES RADWASTE BUILDING JANUARY JUNE 1990 14 3-4 WNP-2 GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES JANUARY JUNE 1990 16 3-5 WNP-2 GASEOUS EFFLUENTS BATCH RELEASES JANUARY JUNE 1990 17 4-1 SCALING FACTORS FOR REQUIRED NUCLIDES 21 4-2 SCALING FACTORS FOR CONDITIONAL NUCLIDES 21 4-3 WNP-2 SOLID WASTE SHIPMENTS JANUARY JUNE 1990 22

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1.0 This report is submitted in compliance with Technical Specification 6.9.1.11. It includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from HNP-2 during the previous six months of operation, with data summarized on a quarterly basis.

2.0 The radwaste liquid effluents were released in "batch mode" during the reporting period. No liquid releases occurred during the first calendar quarter and 23 batch releases occurred during the second calendar quarter. The total time period for the batch releases was 76.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, with the maximum, minimum and average time periods for a release being 13, 1.63 and 3.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> respectively. The volume of dilution water considered is assumed to be the total volume of recirculating cooling tower blowdown flow for the period. The average flow rate of the Columbia River during January through June 1990 was 1.53E+05 cubic feet per second.

Computer runs, using LADTAP II, were performed to verify compliance with Technical Specification limits. There were no liquid releases during the first quarter. The second quarter calculated dose for the maximum individual (adult age group) was 3.5E-03 mrem whole body and 5.0E-03 mrem for the maximum organ. No abnormal liquid releases occurred during this reporting period.

The liquid batch releases were recirculated prior to sampling. A repre-sentative sample was obtained and analyzed for each batch release. A composite of the batch samples for each quarter was analyzed for strontium and iron analyses. The methods used for measuring the total radioactivity were gamma spectroscopy, liquid scintillation and propor-tional counting. Table 2-1 provides a summation of all liquid releases during this reporting period.

The percent of MPC l,imit in Table 2-1 is based on the total of the HPC fractions using the nuclides in Table 2-2 and the concentrations listed in 10CFR20, Appendix B, Table 2, Column 2.

Estimated total errors are listed in Table 2-1, and are propagated from individual error estimates of sample activity, sample volume, tank volume, and tank homogeneity. The estimated total errors were calculated by obtaining the square root of the sum of the squares of the individual error contributions and multiplying by 1.96 for a 95% confidence level.

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Table 2-1 HNP-2 LIQUID EFFLUENTS SUMMATION OF ALL RELEASES Report Period: January June 1990 1st 2nd Est.

Unit Quarter Quarter Total A. Fission and activation products

1. Total release (not including E- 2 .2+ 1 2 ~ Average diluted concentration

-0 B. Tritium

2. Average diluted concentration d r 'i

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C. Dissolved and entrained gases 2 i~~EQ

2. Average diluted concentration r

D. Gross alpha radioactivity E. Volume of waste (prior to F. Volume of dilution water NA .5 01

" At 95% confidence level

Table 2-2 NNP-2 LIQUID EFFLUENTS SOURCE TERMS Report Period: January June 1990 BATCH MODE 1st ** 2nd Nuclides Released Unit Quarter Quarter E-E-

~ 2 5

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TABLE 2-2 (Continued) 1st A* 2nd Nuclides Released Unit Quarter Quarter E-

"* There were no liquid releases during the first quarter of 1990.

NIIN: Less than (<) values are not included in the Total For Period values.

3.0 The gaseous radwaste effluents from HNP-2 were released from three (3) release points:

1. Main Plant Vent mixed mode release

2. Turbine Building ground level release

3. Radwaste Building ground level release The gaseous source terms from each release point are listed in Tables 3-1, 3-2, and 3-3. Table 3-4 provides a summation of the total activity released, the average release rate, the percent of Technical Specifica-tion limit, gross alpha radioactivity and the estimated total error associated with the measurements of radioactivity in the gaseous ef f1uents.

Radioactivity measurements for gaseous effluent releases are performed for fission and activation gases by collecting the samples on charcoal traps and analyzing them using gamma spectroscopy. Tritium is sampled by freeze trapping and analyzed by liquid scintillation counting. Particu-lates and iodines are sampled using particulate filters and charcoal cartridges and are analyzed using gamma spectroscopy.

The percent of Technical Specification limit for fission and activation gases (air dose) was determined for locations 1 through 8 and were based on quarterly limits of ten (10) milli rads for beta and five (5) mi llirads for gamma. Locations 3 through 8 were used to determine the most restrictive value to be used in Table 3-4, Section A.3.

The percent of Technical Specification limit calculations for iodines, particulates with half-lives greater than eight (8) days and tritium are based on the quarterly limit of 7.5 mrem to any organ. Locations 3 through 8 listed below were used to determine the most restrictive value to be used in Table 3-4 for each quarter. The nearest milk sampling location was changed during the second quarter from 6.4 miles SE, to 7.2 miles ESE. Please refer to Section 9 of this report for new or deleted locations for dose assessment and/or environmental monitoring locations.

Total error estimates are propagated from individual error estimates of sample volume, sample activity and effluent flow rate measurements. The overriding uncertainty in all cases is in the measurement of the effluent and sample volumes. The estimated error was determined to be 36/ at the 95% confidence level.

Calculations were performed for releases using the GASPAR II computer program and parameters as outlined in the ODCM. Quarterly doses were determined at the following locations:

Site Boundary; 1.2 miles (ground and inhalation pathways) 1st Qtr. 9. 7E-02 0. 97 1 . 8E-01 3.60 2nd Qtr. 2,0E-02 0.18 1.6E-02 0.32 mrrI 1st Qtr 2. 4E-01 3.20 2nd Qtr 8.0E-02 1.07 Beyond Site boundary: 4.0 miles SE and 3.6 miles ESE respectively (ground and inhalation pathways) at locations having the highest X/Q values for mixed mode release.

1st Qtr. 1. 5E-02 0.15 2.6E-02 0.52 2nd Qtr. 1.1E-02 0.11 5.2E-03 0.10 1st Qtr 3. 1E-02 0. 41 2nd Qtr 1. 7E-03 0.23 l~~i '): 4.8 miles SE (ground, vegetables and inhalation pathways) r 0 1st Qtr.

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4.5E-03 0.05 4.6E-03

~l 0.09 2nd Qtr. 4.7E-03 0.05 2.3E-03 0.05 1st Qtr 9. 7E-02 e ~m, 1.29 2nd Qtr 7.0E-02 0.93

~L jgil 4: 6.4 miles SE (ground, vegetables, meat, cow milk, and inhalation pathways) 1st Qtr. 2.9E-03 0.03 3.1E-03 0.06 2nd Qtr. 3.0E-03 0.03 1,4E-03 0,03 RZesMMU .

1st Qtr 5.6E-02 0.56 2nd Qtr 1.2E-01 1.60 L~~~: 4.2 miles ESE (ground, vegetables and inhalation pathways) 1st Qtr.

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5.0E-03 0.05 3.9E-03 0.08 2nd Qtr. 9.1E-03 0.09 3.9E-03 0.08 V.Tc 1st Qtr 7.6E-02 1.01 2nd Qtr 1.1E-01 1.47

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~MS 1.2E-03 0.02 2nd Qtr. 1 . 2E-03 0. 01 6.0E-04 0.01 HiaheQ&raanJhm 1st Qtr 8. 1E-03 0.11 2nd Qtr 3.9E-03 0.05 I.~jsu~: 4.1 miles ENE (ground, vegetables and inhalation pathways) 1st Qtr.

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~h.0.14 2nd Qtr. 7.9E-03 0.08 3.3E-03 0.07 1st Qtr 8.7E-02 1.16 2nd Qtr 6,5E-02 0.87 7.2 miles ESE (ground, cow milk and inha'lation pathways)

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1. 4E-03 0.03 MKQB1 2nd Qtr 1.2E-01 1.60 In addition to the reactor site, WNP-2 has a permanent laundry facility located approximately 0.75 miles from the site. Its vent) lation system contains HEPA filters on the discharge and is continuously monitored for particulates. Also near this location is a backup chemistry laboratory within the Emergency Operations Facility (EOF). The radiochemical hood within the chemistry lab contains HEPA filters and is monitored for radioactive releases when in operation. Gamma spectrometry indicated no radioactive material present other than that attributable to natural background.

There were no abnormal releases of gaseous effluent during the first and second quarters of 1990.

Table 3-1 HNP-2 GASEOUS EFFLUENTS SOURCE TERMS MIXED MODE RELEASES HAIN PLANT VENT Report Period January June 1990 CONTINUOUS MODE 1st 2nd r

1. Fission gases

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Table 3-1 (Continued

2. Iodines 1st 2nd 2 E-

.1 2 3, Part) cul ates E-E-0 hr -1 E- 4 E- 4 E- .5-

Table 3-1 (Continued)

3. Particulates (continued) 1st 2nd

4. E-

4. Tritium Note: Less than (() values are not included in the Total For Period Values.

1 ~ l Table 3-2 HNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES TURBINE BUILDING Report Period January June 1990 CONTINUOUS MODE 1st 2nd

1. Fission gases 1.2 BKtQQ E+

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2. Iodines 1 . E- ,2E-0 1.2E- 2 I di -1 4 1. E-3 <5. f-V E- 2 11

Table 3-2 (Continued

3. Part)culates 1st 2nd B i

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Table 3-2 (Continued)

3. Particulates (continued) 1st 2nd n m- E- 2

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4. Tritium Note; Less than (<) values are not included in the Total For Period Values.

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Table 3-3 WNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES RADWASTE BUILDING Report Period January June 1990 CONTINUOUS NODE 1st 2nd

1. Fission gases E 1 2, Iodines I -12 .4 4 I e- 1.4E- 2 E- 3 I in -1 4 E-04 <. E-5 I -1 2.

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Table 3-3 (Continued 1st 2nd

3. Particulates r 1 E-0 2,7

4. Tritium

l. E- 1 il n i E 1 .4E Note: Less than (<) values are not included in the Total For Period Values.

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Table 3-4 WNP-2 GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES Report Period January June 1990 1st 2nd Est. Total r 4 O

A. Fission and activation gases

2. Average release rate

3. Percent of Technical B. Iodines

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2. Average release

3. Percent of rate Tec hni ca 1 C. Particulates

1. Particulates with half-

2. Average release rate

3. Percent of Technical ivi D. Tritium E-

2. Average release rate

3. Percent of Technical 0 i 0 1. E+0

" At 95% confidence level Table 3-5 WNP-2 GASEOUS EFFLUENTS BATCH RELEASES Report Period January June 1990 Total Haximum Hi nimum Hean Purge 46.0 16.4 4.6 11.5 17

4.0 A total volume of 10,593 ft (300 m ) of solid waste was transported in 32 shipments during the January through June, 1990 reporting period. The total activity of the waste shipped was 635.7 Ci; 576 Ci contained in dewatered spent resins, 59.7 Ci were contained in Dry Active Waste (DAH).

A.

Dewatered resins accounted for 4,131 ft (117 m ) of the radioactive wastes shipped during the reporting period. The burial containers were ES-190 and EA-142 liners provided by NUPAC Services, Inc. The total activity of the resins shipped during the reporting period was 576 Ci. The principle nuclides and their percent contribution to the total activity are listed in Table 4-3. The solid wastes were shipped to the U.S. Ecology, Hanford burial site using flat bed trailers and NUPAC 10-142, LN-14-170, or USEcology 14-D2.0 casks.

The counting error associated with the total activity was 1.64%.

Other parameters considered in estimating the total error of the activity shipped included the error in measuring the absolute volume, the weight of the waste in the liners, the representative-ness of the sample taken, the homogeneity of the nuclide distribu-tion within a batch or liner and the geometry error in the gamma spectroscopy analysis. The gamma spectroscopy calibration error is approximately 5%. The best estimate of the total error in the activity of spent resin shipped was assumed to be less than or equal to 20%.

B.

t, A total of 6462 ft (183 m ) of DAH was shi pped in 62 Container Products Corporation, B-25 steel boxes, 4 NUPAC Services ES-190 carbon steel liners, and 1 NUPAC Services EA-50 Ferralium HIC. The total activity of the DAH shipped was 59.7 Ci. The values for the activities shipped were determined by using dose rate-to-curie conversion factors. The conversion factors were based on nuclide distribution taken from analysis of'ontamination found in each of the major DAW production areas. The nuclide distribution is updated monthly. Short-lived nuclides were eliminated based on decay of the DAH prior to shipment. A meaningful counting error cannot be gener-ated for the DAW; however, the total error may be assumed to be less than or equal to 20%, since DAW would be subjected to similar error contributions as the spent resins.

4.1 Scaling factors are based on outside laboratory (SAIC) analysis of hard-to-measure nuclides. Scaling factors are updated on an annual basis. For those waste streams where the scaling or the scaled

nuclide concentration is not sufficient to provide a viable scaling factor, the final EPRI Report "Updated Scaling Factors in Low Level Radwaste", NP-5077, March 1987 has been used as a basis for the determination of a scaling factor.

Sampling of individual waste streams was performed with analyses performed by an outside lab. The H-3 concentration was measured per gram of waste material. This value was compared to the Reactor Coolant System H-3 concentration. The scaling factor is derived from the ratio of the H-3 concentration in the waste stream to RCS H-3 concentration.

Sampling of the individual waste stream was performed with analysis by off-site lab to determine isotopic concentration. Ratios were developed between the scaled nuclide to the scaling nuclide concen-tration determined by analysis. In those cases where the scaling nuclide is not available in large enough quantities to develop reliable (viable) scaling factors, the recommendations made in Section 3 of the referenced EPRI report will be followed.

TRU nuclides are scaled to Ce-144, as recommended by the AIF report "Methodolgies for Classification of Low Level Radioactive Waste from Nuclear Power Plants". These nuclides are not considered to be present if the scaled values are less than: 1 nCi/g for TRU, 35 nCi/g for Pu-241 or 200 nCi/g for Cf-242. TRU nuclides will be reported if the scaling nuclide (Ce-144) is reliably detected and Cs-137 is also present.

Sampling of individual waste streams has been performed with analyses by an outside laboratory. Cs-137 and Sr-90 concentrations were measured in each waste stream except waste oil. The ratio of Cs-137 to Sr-90 has been determined and is used as the scaling factor for Sr-90 from Cs-137, For waste oil, the values from the referrenced EPRI Report will be used for scaling factors. Co-60 and Ni-63 con-centrations were measured in each of the sampled waste streams. The ratio of Co-60 to Ni-63 has been determined and is used as the scal-ing factor for Ni-63 from Co-60.

Table 4-1 lists scaling factors by waste stream for those nuclides that are required to be reported. Table 4-2 lists scaling factors for the conditional nuclides that are reported only when the scaling nuclide is found to be present.

4.2 The Process Control Program (PCP) used to control solidification at NNP-2 will be provided by the vendor waste processor, Pacific Nuclear Inc. in accordance with Contract C-20452, and will be subjected to POC review prior to any solidification of radwaste. Normally approved High Integrity Containers (HIC's) are used for the trans-port of wastes requiring stabilization. Other portions of the rad-waste program are controlled by the NNP-2 procedures PPH 1.12.1, "Radwaste Management Program", PPH 1.12.2, "Radwaste Process Control Program", and 1.12.3, "Contract (Vendor) Haste Processing". There were no significant changes during the reporting period.

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SCALING FACTORS b - i 1 RWCU CFD EDR/FDR EDR/FDR POWDER POWDER POWDER BEAD

~D EGIU BKSLH RGiIH KITH 2JLK QLL H-3/Rx Coolant 4.5E-1 4.3E-1++ 4.30E-1 4 '0E-1++ 2.22E-1 3. 10E-1 4.0E-5+

C-14/Co-60 6.21E-4 6.4E-5 6.2E-4++++ 1 ~ 64E-4 2.90E-2 8.81E-5 1.3E-2+

Tc-99/Cs-137 4.6E-4+ 1.1E-4+ 9.3E-5+ 9.3E-5+ 9.3E-5+ 9.3E-5+ 4.2E-5+

I-129/Cs-137 2.6E-4+ 1.0E-5+ 3.9E-5+ 3.9E-5+ 3.9E-S+ 3.9E-5+ 6.3E-S+

Ni-63/Co-60 4.27E-2 7.74E-3 2.4E-2 4.53E-2 2.4E-Z 1 . 5E-2+++ 1.2EO+

Fe-55/Co-60 7.06E-1 2.6E-l 3.4E-l 3.06E-1 1.06E-1 4. 10E-1 1. 5EO+

Sr-90/Cs-137 2.6E-3+ 1.2E-2+ 1.6E-2+ 5.00E-3 5.91E-3 2.67E-S 3.3E-l+

Pu-239/Ce-144 4.5E-3+ 5.8E-3+ 9 'E-3+ 9.7E-3+ 9.7E-3+ 8.7E-4+ 1. 1E-2+

Pu-238/Pu-239 1.5EO+ 8. OE-1+ 1.7EO+ 1.7EO+ 1.7EO+ 1.7EO+ 1.6EO+

Pu-241/Pu-239 1. 1E2+ 9.4El+ 9.6El+ 9.6El+ 9.6E1+ 9. 1E1+ 1.2E2+

Am-241/Pu-239 9. 1E-1+ 3. 9E-1+ 6.6E-1+ 6.6E-l+ 6.6E-l+ 1. 7EO+ 4.7E-l+

Cm-242/Pu-239 9. 5E-1+ 7. OE-1+ 9.7E-1+ 9.7E-1+ 9.7E-1+ S. 7E-1+ 3. 1E-1+

Cm-244/Pu-239 7.2E-l+ 3. OE-1+ 7.6E-1+ 7.6E-1+ 7.6E-1+ 7.8E-l+ 2. 9E-1+

Scaling or scaled nuclide was not present in enough concentration to make determination of scaling factor. In these cases the scaling factor was obtained from the "Updated Scaling Factors in Low-Level Radwaste" EPRI NP-5077 Final Harch 1987.

stitiall Outside laboratory (SAIC) analysis of the H-3 concentration in RWCU 8 EDR/FDR resins identified H-3 concentrations higher than those indicated for reactor coolant. This is not consistent with the data presented in the EPRI reports and has no logical basis since H-3 is not concentrated in resin and is a function only of the amount of inter-water trapped in the resin. The water content of dewatered powdered resin was identified as 551. in EPRI NP-4037 and would be expected to be even less for the drying system in use at WNP-2 which utilizes a humidity gauge to define the drying cycle end point. For these reasons the H-3/RX coolant ratio reported for CFD powdered resin will be used for all three powdered resin waste streams. The CFD H-3/RX coolant ratio of 0.43 is consistent with the EPRI reports and the drying system in use at WNP-2.

Independent laboratory analysis showed the Ni-63 concentration of sludges at 4.03E-3 uCi/gm which compares to the Co-60 concentration of 3.52E-2 uCi/gm. This comparison would yield a Scaling Factor of 1. 14E-l. The above mentioned EPRI Report recommends a Scaling Factor of 1.5E-2. Because of the long period of time between the generation of the waste and the counting of the sample (approximately 1 year) the EPRI Number is considered more accurate.

The independent laboratory analysis showed the C-14 concentration in CFD of 3.62E-3 uCi/gm which compares to the Co-60 concentration of 5 '6E-3 uCi/gm. This comparison would yield a Scaling Factor of 6.07E-l. The above mentioned EPRI report recommends a Scaling Factor of 6.2E-4. It is felt that there was cross contamination of the sample at the lab resulting in high concentration of C-14. The recomnended EPRI number will be used.

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Table 4-3 HNP-2 SOLID HASTE SHIPMENTS January dune 1990 A. SOLID HASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL

1. Type of Haste Waste Stream Unit 6-month

a. Spent resins, filter sludges, m 117

b. Dry active waste, contaminated m 183

c. Irradiated components, control m No Ship-

d. Other, (absorbed aqueous liquid) m No Ship-

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

a. Dewatered Spent Resins 2

"Indicates scaled nuclide 22

b. Dry Active Hastes (DAH) 4

c. Irradiated Components None

d. Other Absorbed Liquids None

3. Solid Haste Disposition Tr i 32 Flat bed trailer (6) US Ecology 10-142 Cask (2) Richland, WA 14-170 Cask (13) 14-D2.0 Cask (11)

B. IRRADIATED FUEL SHIPMENTS (Disposition)

None "Indicates scaled nuclide The top ten radionuclides are listed in descending curie quantities.

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5.0 E L D The meteorological data for the first half of calendar year 1990 will be included in the Semiannual Effluent Report due 60 days after january 1, 1991 and will include data covering the full calendar year 1990. An extended outage of the main meteorological tower occurred during the 1st Quarter 1990. A discussion of its effect on the yearly data collection will be included with the data in the above report.

6 0 0 The dose impact on man for the calendar year 1990 will be included in the Semiannual Effluent Report due 60 days after January 1, 1991.

Ho amendments were made to the NNP-2 Offsite Dose Calculations Manual (ODCM) during this reporting period.

8.0 V No changes were made to the Process Control Program (PCP) during this reporting period which required POC approval.

E E 9.1 Locations where GASPAR II dose calculations were performed for the first and/or second quarters of 1990:

9.1.1 4.8 miles southeast (SE) for the highest organ dose using ground, inhalation and vegetation pathways.

9.1.2 6.4 miles southeast (SE) for the highest organ dose using ground, vegetables, cow milk inhalation and meat pathways.

9.1.3 4.2 miles east southeast (ESE) for the highest organ dose using ground, inhalation and vegetable pathways.

9.1.4 7.2 miles east southeast (ESE) for the highest organ dose using ground, inhalation and cow milk pathways.

9.1.5 4.3 miles northeast (NE) for the highest organ dose using ground and inhalation pathways.

9. 1.6 4.1 miles east northeast (ENE) for the highest organ dose using ground, inhalation and vegetable pathways.

9.2 A new milk sampling location was established during the second quarter at 10.9 miles southeast (SE). This location was the only available milk sampling site in the region, other than the ones already sampled. Since this location is not within the 9.9 miles required by the Technical Specifications, this deviation from the Technical Specifications requirements will be reported in the annual "Radiological Environmental Monitoring Program" (REMP) report until a closer milk sampling site is located.

9.3 Milk sampling was discontinued at 6.4 miles southeast (SE) due to the unavailability of milk samples there.

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No major changes were made to the radioactive waste systems (liquid, gaseous, or solid) during this reporting period.

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