ML18152A022

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Semiannual Radioactive Effluent Release Rept,Jul-Dec 1988. W/890302 Ltr
ML18152A022
Person / Time
Site: Surry  Dominion icon.png
Issue date: 12/31/1988
From: Blount P, Cartwright W
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
89-141, NUDOCS 8903090333
Download: ML18152A022 (56)
v • d • e


Text

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e e VIRGINIA ELECTRIC AND POWER COMPANY 10 CFR 72.33(d)(3)

RICHMOND, VIRGINIA 23261 W. R. CA:RTWRIOHT VICE PRESIDENT March 2, 1989 NUCLEAR United States Nuclear Regulatory Commission Serial No.89-141 Attention: Document Control Desk NO/RPC:vlh Washington, D. C. 20555 Docket Nos. 50-280 50-281 72-2 License Nos. DPR-32 DPR-37 SNM-2501 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Enclosed is the Surry Power Station Semi-Annual Radioactive Effluent Release Report for July 1, 1988 through December 31, 1988. The report, submitted pursuant to Surry Power Station Technical Specification 6.6.B.3 and ISFSI Technical Specification 1.4.1 of Appendix C, 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.

Enclosure cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Mr. Robert M. Bernero, Director Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555

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POW 24 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SURRY POWER STATION (July 1. 1988 Through December 31. 1988)

Prepared~~=-* ~

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Assistant Supervisor. Health Physics Count Room and Environmental Reviewed By: .4,.... r~ 4"..L..

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  • Health Physics Technical Services Reviewed By: :a . * *' ~

Superinteflent.HealthP

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'r I j FORWARD This report is submitted as required by Appendix A to Operating License No.'s DPR-32 and DPR-37, Technical Specifications for Surry Power Station, Units 1 and 2, Virginia Electric and Power Company, Docket No.'s 50-280, 50-281, Section 6.6.B.3. and as required by Appendix C to Materials License No. SNM-2501, Technical Specifications for Environmental Protection for Surry Independent Spent Fuel Storage, Docket No. 72-2, Section 1.4.1.

  • RADIOACTIVE EFFLUENT RELEASE REPORT FOR THE SURRY POWER STATION (July 1. 1988 Through December 31. 1988)

Index Section No. Subject Page 1

1 Purpose and Scope 1&2 2 Discussion 2

3 Supplemental Information Attachment 1 Effluent Release Data Attachment 2 Annual and Quarterly Doses Attachment 3 Revisions to Offsite Dose Calculation Manual (ODCM)

Attachment 4 Revisions to Process Control Program (PCP)

Attachment 5 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems Attachment 6 Inoperability of Radioactive Liquid and Gaseous Effluent Instrumentation Attachment 7 Unplanned Releases Attachment 8 Lower Level of Detection (LLD) for Effluent Analysis

Page 1 of 2

1.0 Purpose and Scope

The Radioactive Effluent Release Report includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste as outlined in Regulato-ry Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", Revision 1, June 1974, with data summarized on a quarterly basis following the format of 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 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 also includes a list of unplanned releases during the reporting period.

As required by Technical Specification 6.8B, changes 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 by Technical Specification 6.8A.

Major changes to the radioactive liquid, gaseous and solid waste treatment systems are reported as required by Technical Specification 6.9. Information to support the reason for the change and a summary of the 1 OCFR50.59 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. 7E.2, a list and explanation for the inoperability of radioactive liquid and/or gaseous effluent monitors are provided in this report.

2.0 Discussion The basis for the calculation of the percent of technical specification for the critical organ in Table IA of Attachment 1, is Technical Specification 3.llB.l.a (ii). Technical Specification 3.1 lB. l.a (ii) 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 IA of Attachment 1, is Technical Specification 3.llB.l.a (i). This Technical Specification 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 l, is Technical Specification 3.1 lA. l.a. Technical Specification 3.1 lA. l.a states that the concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in 1 OCFR20, 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 10-4 microcuries/ml.

Page 2 of 2 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 bounda-ry. The maximum exposed member of the public from the release 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 grass-cow-milk pathway, with the critical organ being the thyroid. The maximum exposed memb-er of the public from radioactive materials in liquid effluents in unrestricted areas is defined as an adult, exposed by either the invertebrate or fish pathway, with the critical organ being either the thyroid or GI-LLI. The total body dose was also determined for this individual.

Unplanned releases presented in Attachment 7 are defined in Technical Specification 6.6.B.3 as those gaseous releases exceeding Technical Specification 3.11.B. l.a and those liquid releases exceeding Technical Specification 3.11.A.1.a.

The typical lower level of detection (LLD) capabilities of the radioactive effluent analysis instrumentation are presented in Attachment 8. These LLD values are based upon conservative conditions (i.e., minimum sample volumes and maximum delay time prior to analysis). Actual LLD values may be lower. If an isotope is not detected when analyzing effluent samples, then the activity of that isotope will be reported as Not Detected (N/D) on Attachment 1 of this report. When all isotopes listed on Attachment 1 for a particular quarter and release mode are less than the lower level of detection, then the totals for this period will be designated as Not Applicable (N / A).

3.0 Supplemental Information Technical Specification 3.11.D. l.d requires the identification of the cause for the unavailability of milk or leafy vegetation samples, required by Technical Specification, Table 4.9-3, and identification for obtaining replacement samples.

All milk and leafy vegetation samples required by Table 4.9-3 were available for collection during the period of July 1, 1988 through December 31, 1988.

Technical Specification 3.11.D.2.b requires the identification of new sample locations determined with the Land Use Census as yielding a calculated dose or dose commitment greater than the values being calculated in Technical Specif-ication 4.9.C. No new sample location(s) that may yield a greater dose or dose commitment that are currently used in Technical Specification 4.9.C, were identified in the Land Use Census.

Dry Cask Independent Spent Fuel Storage Installation Technical Specification Appendix C, 1.4.1 requires reporting the quantity of each of the principal radionuclides released from the installation to the environment in liquid and gaseous effluents during the previous 6 months of operations. There were no liquid or gaseous effluent releases from the Dry Cask Independent Spent Fuel Storage Installation during the period of July 1, 1988 through December 31, 1988.

Attachment 1 EFFLUENT RELEASE DATA (July 1, 1988 Through December 31, 1988)

Attachment 1 TABLE IA Page 1 of 7 EFFLUENT ANO WASTE OISPOSAL SEMIANNUAL REPORT PERIOO: 7/ 1/88 TO 12/31/88 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES SURRY POWER STATION UNITS 1&2 A. FISSION & ACTIVATION GASES UNIT THIRO QTR. FOURTH QTR. J'6' EST. ERROR

1. TOTAL RELEASE CI 1.51E 02 9.65E 00 2. SOE 01
2. AVG RELEASE RATE FOR PERIOO UC~/SEC 1. 90E 01 1.21E 00 B. IOOINE
1. TOTAL I-131 CI 4. 37E-03 1.85E-04 2.50E 01
2. AVG RELEASE RATE FOR PERIOO UCI/SEC 5.50E-04 2.33E-05 C. PARTICULATE
1. HALF-LIVES )8 OAYS CI 6. 74E-03 9.07E-04 2.50E 01
2. AVG RELEASE RATE FOR PERIOO UCI/SEC 8. 48E-04 1.14E-04
3. GROSS ALPHA RAOIOACTIVITY CI J.85E-06 2.25E-06
0. TRITIUM
1. TOTAL RELEASE CI 6.12E 00 7. 27E 00 2. 50E 01
2. AVG RELEASE RATE FOR PERIOO UCI/SEC 7.69E-01 9.15E-01 J.,.'CENTAGE OF T. S. LIMITS CRITICAL ORGAN OOSE RA TE 3.87E-02 5.60E-03 TOTAL BOOY OOSE RATE 5.64E-02 6. 73E-05 SJ{IN OOSE RATE 2.23E-02 2.80E-05

J I TABLE lB Attachment Page 2 of 7 EFFLUENT AN]) WASTE J)ISPOSAL SEMIANNUAL REPORT PERIOJ): 7/ 1/88 TO 12/31/88 GASEOUS EFFLUENTS-MIXEJ)-MOJ)E RELEASES CONTINUOUS MOJ)E BATCH MOJ)E SURRY POWER STATION UNITS 1&2 UNIT THIRJ) FOURTH THIRJ) FOURTH QUARTER QUARTER QUARTER QUARTER

1. FISSION & ACTIVATION GASES J{R-85 CI N/J) N/J) 2.54E-01 N/J)

KR-85M CI N/J) N/J) 1.15E-02 N/J)

J{R-87 CI N/J) N/J) N/J) N/J)

KR-88 CI N/J) N/J) N/J) N/J)

XE-133 CI 1. 34E 01 8. 73E 00 9.00E 00 1.13E-Ol XE-135 CI 5.28E-02 N/J) 9. 94E-01 N/J)

XE-135M CI N/J) N/J) N/J) N/J)

XE-138 CI N/J) N/J) N/J) N/J)

XE-133M CI N/J) 7.15E-03

  • 1.52E-01 N/J)

AR-41 CI N/J) N/J) N/J) N/J)

XE-131M CI N/J) 7. 71E~Ol 2.99E-02 3.0lE-02 TOTAL FOR PERIOJ) CI 1. 34E 01 9.50E 00 1. 04E 01 1. 43E-01

2. IO.DINES I-131 CI 7.98E-06 7. 64E-07 5.lOE-05 l.30E-06 I-133 CI 4.09E-07 N/.D 5. 47E-06 N/J)

I-135 CI N/.D N/.D N/.D N/.D I-132 CI N/.D N/.D N/J) N/J)

TOTAL FOR PERIO.D CI 8.39E-06 7. 64E-07 5.64E-05 1.30E-06

3. PARTICULATES SR-89 CI N/.D N/.D N/J) N/J)

SR-90 CI N/J) N/.D N/.D N/.D CS-134 CI N/J) N/J) N/.D N/.D CS-137 CI 4.13E-08 5.53E-08 7.73E-09 5.52E-09 BA-140 CI N/J) N/J) N/J) N/.D LA-140 CI N/.D N/J) N/.D N/.D C0-60 CI 4.83E-08 8. 67E-08 2. 57E-10 N/J)

CR-51 CI N/.D N/.D N/J) N/.D MN-54 CI N/.D N/J) N/J) N/.D C0-57 CI N/.D N/.D N/.D N/J)

C0-58 CI N/.D N/J) 8. 57E-09 3.44E-09 SB-125 CI N/.D N/.D N/.D N/.D NB-95 CI N/J) N/J) N/J) N/.D SE-75 CI 1. 97E-08 N/.D N/.D N/J)

RB-88 CI N/.D N/P l .16E-04 N/P AG-llOM CI N/.D N/J) N/.D N/.D CS-138 CI N/J) N/J) N/J) N/J)

J I TABLE IC Attachment 1 Page 3 of 7 EFFLUENT ANO WASTE OISPOSAL SEMIANNUAL REPORT PERIOO: 7/ 1/88 TO 12/31/88 GASEOUS EFFLUENTS-GROUNO LEVEL RELEASES CONTINUOUS MOOE BATCH MOOE SURRY .POWER STATION UNITS 1&2 UNIT THIRO FOURTH THIRO FOURTH QUARTER QUARTER QUARTER QUARTER I. FISSION & ACTIVATION GASES KR-85 CI N/0 N/0 N/0 N/0 J{R-85M CI 4. 73E-05 N/0 N/0 N/0 KR-87 CI N/0 N/0 N/0 N/0 J{R-88 CI N/0 N/0 N/0 N/0 XE-133 CI 4.17E 01 N/0 8. 51E 01 N/0 XE-135 CI l .18E-03 N/0 l .19E-04 N/0 XE-135M CI 2.95E-04 N/0 N/0 N/0 XE-138 CI N/1) N/1) N/1) N/0 XE-133M CI N/1) N/1) 4.86E-06 N/0 AR-41 CI N/1) N/1) l.26E-04 N/1)

XE-131M CI N/0 N/0 N/1) N/0 TOTAL FOR PERIOJ) CI 4.17E 01 N/A 8.51E 01 N/A

2. IOOINES I-131 CI 3.09E-03 l.83E-04 l.22E-03 N/0 I-133 CI 2.64E-03 N/1) l.29E-04 N/0 I-135 CI 9.43E-05 N/0 N/0 N/0 I-132 CI N/1) N/1) 4.35E-06 N/0 TOTAL* FOR PERIOO CI 5.83E-03 1. 83E-04 l.35E-03 N/A
3. PARTICULATES SR-89 CI J.20E-06 N/0 N/1) N/0 SR-90 CI 2.59E-07 N/1) N/0 N/0 CS-134 CI 5.95E-04 6. 70E-05 5.84E-06 N/0 CS-137 CI l.25E-03 2.44E-04 3.17E-05 N/1)

BA-140 CI N/0 N/1) N/0 N/0 LA-140 CI N/1) N/1) N/0 N/1)

C0-60 CI l.15E-03 2.99E-04 1. lOE-05 N/1)

CR-51 CI 1.45E-04 N/1) N/0 N/1)

MN-54 CI 8.96E-05 9.02E-06 3. 78E-11 N/0 C0-57 CI 5.76E-06 N/0 N/0 N/1)

C0-58 CI 3.41E-03 2.88E-04 l.55E-05 N/1)

SB-125 CI 2.04E-05 N/0 2.26E-10 N/1)

NB-95 CI 7. 56E-06 N/0 N/0 N/D SE-75 CI N/1) N/0 N/0 N/0 RB-88 CI 6. 37E-07 N/0 l. lOE-09 N/0 AG-llOM CI N/0 N/0 2.63E-10 N/0 CS-138 CI l.98E-06 N/0 N/0 N/0

J \ TABLE 2A Attachment 1 Page 4 of 7 EFFLUENT ANlJ WASTE lJISPOSAL SEMIANNUAL REPORT PERIOlJ: 7/ 1/88 TO 12/31/88 LIQUilJ EFFLUENTS-SUMMATION OF ALL RELEASES SURRY POWER STATION UNITS 1&2 UNIT THIRlJ QTR. FOURTH QTR.  % EST. ERROR A. FISSION ANlJ ACTIVATION PROlJUCTS

1. TOTAL RELEASE (NOT INCLUlJING TRITIUM, GASES, ALPHA) CI 5. 28E-01 6. 48E-01 2.50E 01
2. AVG lJIL. CONC. lJURING PERIOlJ UCI/ML 9.62E-10 7.23E-09
3. PERCENT OF APPLICABLE LIMIT J'6" 1.53E-02 6.90E-02 B. TRITIUM
1. TOTAL RELEASE CI 1.02E 02 1.05E 02 2.50E 01
2. AVG lJIL. CONC. lJURING PERIOlJ UCI/ML 1. 86E-07 1.17E-06
3. PERCENT OF APPLICABLE LIMIT 0/
  • '0 6.19E-03 3.90E-02 C. lJISSOLVElJ ANlJ ENTRAINElJ GASES
1. TOTAL RELEASE CI 5.13E-Ol 7.34E-02 2. SOE 01
2. A VG lJIL. CONC. lJURING PERIOlJ UCI/ML 9.35E-10 8.19E-10
3. PERCENT OF APPLICABLE LIMIT .%" 4.68E-04 4. lOE-04 ROSS ALPHA RAlJIOACTIVITY

. TOTAL RELEASE CI O.OOE-01 8.00E-05 2.50E 01 E. VOLUME OF WASTE RELEASElJ (PRIOR TO lJILUTION) LITERS 5.45E 07 5.63E 07 3.50E 00 F. VOLUME OF lJILUTION WATER USElJ lJURING PERIOlJ LITERS 5. 49E 11 8. 95E 10 3.50E 00

' I TABLE 2B Attachment 1 Page 5 of 7 EFFLUENT ANIJ WASTE IJISPOSAL SENIANNUAL REPORT PERIOIJ: 7/ 1/88 TO 12/31/88 LIQUI.D EFFLUENTS CONTINUOUS Jl10IJE BATCH Jl10IJE SURRY POWER STATION UNITS 1&2 UNIT THIRIJ FOURTH THIRIJ FOURTH QUARTER QUARTER QUARTER QUARTER SR-89 CI N/IJ N/IJ N/IJ N/IJ SR-90 CI N/IJ N/IJ N/IJ N/IJ CS-134 CI 7.59E-03 3.96E-03 8.0lE-03 5.17E-02 CS-137 CI 6.96E-02 4.68E-02 2.76E-02 1.35E-01 I-131 CI N/IJ N/IJ 2.07E-02 1.20E-02 C0-58 CI N/IJ N/IJ 7.53E-02 5.44E-02 C0-60 CI 2.25E-03 5.46E-04 1. 26E-01 1.14E-01 FE-59 CI N/IJ N/IJ 5.34E-04 4.46E-04 ZN-65 CI N/IJ N/IJ 1. 87E-05 1.19E-05 JlfN-54 CI N/IJ N/IJ 3.58E-03 2. 79E-03 CR-51 CI N/IJ N/IJ 2. 47E-02 1. 76E-02 ZR-95 CI N/IJ N/IJ 1.52E-03 7.15E-04 NB-95 CI N/IJ N/IJ 3.18E-03 2.45E-03 M0-99 CI N/IJ N/IJ 4.28E-05 2.38E-05 TC-99/11 CI N/IJ N/IJ 1.06E-04 5.39E-06 BA-140 CI N/IJ N/IJ 8.82E-03 N/IJ LA-140 CI N/IJ N/IJ N/IJ N/IJ CE-141 CI N/IJ N/IJ N/IJ N/IJ SB-124 CI N/IJ N/IJ 8. 09E-03 1.00E-02

  • - SB-125 CI N/IJ N/IJ 9.69E-02 1.22E-01 AG-llOM CI N/IJ N/IJ 1.55E-03 2.BlE-03 RU-103 CI N/IJ N/IJ 5.77E-04 1. 36E-04 NA-24 CI N/IJ N/IJ 1.55E-04 9.67E-05 TE-132 CI N/IJ N/IJ 2.04E-05 6.80E-05 NA-22 CI N/IJ N/IJ 1.53E-06 N/IJ C0-57 CI N/IJ N/IJ 2.lOE-04 2.58E-04 I-133 CI N/IJ N/IJ 8.63E-04 3.14E-06 I-132 CI N/IJ N/IJ 1. 74E-04 6.32E-05 I-135 CI N/IJ N/IJ 2.85E-05 N/IJ FE-55 CI N/IJ N/IJ 3.98E-02 6.98E-02 I-134 CI N/IJ N/IJ N/IJ 1.53E-05 CE-144 CI N/IJ N/IJ N/IJ 7. 74E-06 SR-92 CI N/IJ N/IJ N/IJ 5.87E-06 TOTAL FOR PERIOIJ CI 7.95E-02 5.13E-02 4.48E-01 5. 96E-01 XE-133 CI N/D N/IJ 4. 87E-01 7.33E-02 XE-135 CI N/IJ N/IJ 2.04E-02 N/IJ AR-41 CI N/IJ N/IJ 5.29E-05 1.94E-05 XE-135/11 CI N/IJ N/IJ 1.29E-04 N/IJ XE-133/11 CI N/IJ N/IJ 1.24E-03 N/IJ KR-85/11 CI N/IJ N/IJ 3.58E-06 N/JJ XE-131/11 CI N/IJ N/IJ 4.60E-03 N/IJ

., l Attachment 1 Page 6 of 7 TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS PERIOD 07/0l/88-12/31/88 SURRY POWER STATION A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL{Not irradiated fuel)

I. Type of waste Unit 6-month Est. Total Period Error,%

a. Spent resins, filter sludges, evaporator m3 2.24E+Ol 5.00E+OO bottoms, etc. Ci 2.05E+Ol 2.50E+Ol
b. Dry compressible waste, contaminated m3 3.58E+02* 5.00E+OO equip., etc. Ci 7.57E+OO 2.50E+O 1
c. Irradiated components, control m3 O.OOE+OO O.OOE+OO rods, etc. Ci O.OOE+OO O.OOE+OO
d. Organic waste(i.e. oil and scintillation ma O.OOE+OO O.OOE+OO fluid) Ci O.OOE+OO O.OOE+OO
2. Estimate of major nuclide composi.tion(by type of waste)
a. Cs-137  % 4.36E+Ol Co-60  % 2.26E+Ol Cs-134  % l.38E+Ol Ni-63  % 6.88E+OO Co-58  % 6.59E+OO Fe-55  % 3.82E+OO Mn-54  % l.13E+OO
b. Co-60  % 4.82E+Ol Fe-55  % 3.00E+Ol Cs-137  % 8.93E+OO Ni-63  % 8.44E+OO Cs-134  % 2.82E+OO Co-58  % l.28E+OO
c.  %
d. - - -  %
3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 6 Truck Barnwell, SC 12 Truck Oak Ridge, TN

J **

Attachment 1 Page 7 of 7 TABLE 3 (Cont)

EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT SOLID WASTE AND IRR ADIATED FUEL SHIPMENTS PERIOD 07/01/88-12/31/88 B. IRRADIATED FUEL SHIPMENT(Disposition)

Number of Shipments Mode of Transportation Destination 0

  • 12 shipments of dry compressible waste were shipped from Surry to a licensed waste processor for volume reduction. Therefore, the volume as listed for this waste type is not representative of actual volume buried. The total volume of dry compressible waste disposed of at Barnwell, S.C., for this reporting period, was 2.97E+2 m3

Attachment 2 Page 1 of 1

  • LIQUID ANNUAL AND QUARTERLY DOSES GASEOUS Total Body Thyroid GI-LLI Gamma Beta Thyroid (mRem) (mRem) (mRem) (mRad) (mRad) (mRem) 1st Quarter l.33E-03 3.19E-03 3.64E-03 8.37E-02 2.22E-01 7.2SE-02 2nd Quarter S.68E-03 l. l 2E-03 3.82E-02 S.27E-02 l.08E-01 2.70E-02 3rd Quarter 1.91E-03 4.09E-03 l.08E-02 8.SSE-02 2.54E-01 8.63E-02 4th Quarter 8.90E-02 3.86E-03 3.61E-01
  • I.03E-04 3.23E-04 3.82E-03 9.79E-02 1.23E-02 4.14E-01 2.22E-01 S.82E-01 l.90E-01

.nnual

Attachment 3 Page 1 of 1

There were no changes to the Offsite Dose Calculation Manual (ODCM) during the period of July 1, 1988 through December 31, 1988.

Attachment 4 Page 1 of 1

In December 1988, the Process Control Program was deleted as a separate entity and incorporated into the Health Physics Department procedures. The procedure was rewritten to conform with the requirements and implementation of the Virginia Power Radiation Protection Plan. Criteria for the processing and shipping of solid radioactive waste have not changed.

/

OCT 1 f 19e8 "A\,WU-R& V Page I of 4

~,,___ _ J/DM ti/-t;S"' -19 8 PROCEDURE CHANGE REQUEST VIRGINIA POWER SURRY POWER STATION AOM21

1. DEPARThlENT:

N/A - ~rocess Control ?rogram

~ND. 3.1.NIT No. 4.RfV. DATE:

Process Contr. ?rag. 1 &2 04/29/86 5.TITU:

RADIOACTTVE l*!A8TE ?ROCESS CONTROL P. or:

a.owa~

Delete procedure A 11u1~ Kc.vu I. DATE:\?-~ ~'1) t.REFe:ea.s:

Radiation Protec Radiatio *on "!?lan*

11

  • SAFETY~YSS REOUAED? 'tS I OCFR SO* H YES IC) I OCR.72
  • 31 YU NO 0 REVlEV RECJJlRED? 0 0 REVIE'W REQUIRED{] [El SAFETY AHALYSIS IS REOUIRED FOR >>rt YES ANSWER. SAFETY AHALY'SIS N#J 1oa:RSO.Si' 10CFR72.35 REV1EW IS REQUIRED FORM<< YES

>>$HfJ'. N OOESTlCNS 1 nfi:U>>,t 3.

a 1.The facility, as described In the UFSAR or Technical Specfications, wm be changed.

NO 0 @ Sections Reviewed: UfSAR 11; Tech Specs 5.0 1 6,J

2. The procedures or methods of operation. as desen'bed In the UFSAR or Technical 0 Specifications, wiD be cha!'QGd.

Section, Reviewed: UF.S'AR 11,12; Tech Specs 3.11, 4.9, 6.1, 6.4 WIS NO 3. A test or experiment, which Is not desenbed In the UFSAR or Technical Specifications, la 0 El proposed.

WIS NO 4. Safety-Related structures, systems, equtptment or components*. not described In ht UFSAR D [] or Technlcal Specifications. wftl be changed.

't9I NO 5. The faclrrty, procedure, or method of operation which eould affect equlptment Important to 0 a safely wlD be changed.

12. REVIEWED: PEER SUB-COMMITTE D VERIFICATICW D VALIDATION 15..RE<XMMENOED~norY'l.\r" lt::J' N't-nvwcu
  • 0 DISAPPROVED 18.REV\EWS> B Y ~ tfJClEAR WETf ~ OPEAATNGCOMMTTEI: 1tl bd N>PROVED D APPROVED /.S ~FED
23. DATE:

J,;;;i. -/ -8' g-

(1)

PROCEDURE Process NUMBER: Contr. Prog.

SURRY POWER STATION (2)

DATE : APR 2 9 1986 (3) (4 TYPE PROCEDURE: PROCESS CONTROL PROGRAM UNIT: 1 & 2 (5)

TITLE: RADIOACTIVE WASTE PROCESS CONTROL PROGRAM (6)

CONTENTS SECTION

1. Scope and Purpose
2. Definitions
3. System and Process Descriptions
4. Waste Sources and Characteristics
5. Sampling, Analysis and Process Surveillance
6. Contractor Services/Station Interface and Requirements
7. Station Records
8. References Think ALARA RECOMME~ --~~____..;==------===:......:::==~-=:;..._-=a;,;,(7,,......~~~--------~""'"""

DATE: ¢~/!,

!TY ASSURANCE REVIEW: (9) (10)

DATE: ~~,1~.c APPROVED STATION NUCLEAR S FETY AND OPERATING COMMITTEE REVIEW: (11) (12)

DATE: APR 2 9 1986 APPROVED (13) (14)

DATE:

SPS

' . PCl'-1,0 APR 2 9 1986 Surry Power Station Radioactive Waste Process Control Program Table of Contents 1,0 Scope and Purpose 2.0 Definitions 3,0 System and Process Descriptions 4,0 Waste Sources and Characteristics 5,0 Sampling, Analysis and Process Surveillance 6.0 Contractor Services/Station Interface and Requirements 7.0 Station Records 8.0 References

,, PCP-2.0 APR 2 9 1986 SURRY POWER STATION RADIOACTIVE WASTE PROCESS CONTROL PROGRAM 1.0 Scope and Purpose The Process Control Program (PCP) for Surry Power Station provides the minimum requirements and guidelines to be followed to assure that "wet" radioactive wastes are processed and packaged in accordance with all Federal and State regulations.

The program encompasses those forms of radioactive wastes as defined in section 2.0 and establishes the processing parameters through sampling, testing and determinations, to ensure an acceptable product for trans-portation and burial.

Methods will be provided to assure that waste solidification/dewatering systems are operating properly and to substantiate an acceptable solidi-fied/dewatered and/or absorption process product.

The Process Control Program shall be implemented by all personnel who operate dewatering and solidification equipment and/or control absorp-tion processing, collect and process samples used to establish process parameters and those who prepare documentation for shipping of radioac-tive wastes.

PCP-3.0 APR 2 9 1986 2.0 Definitions 2.1 Batch - A discrete quantity of waste material exhibiting certain chemical and physical properties which may be considered a homogeneous mixture for the purposes of sampling, testing and processing.

2.2 Sample - A reasonably representative aliquot of a batch, as defined above, to be obtained for the purpose of deter-mining/verifying solidification parameters, absorption processing or radioactive constituents.

2.3 Direct Measurements/Analysis - The methodology used to quantify the radioactive concentrations existing in the waste.

Direct gamma spectroscopy measurements/analysis of the waste form itself and/or the acquisition of samples for subsequent measurement/analysis, constitutes an accept-able means of quantifying the radioactive content of the waste.

2.4 Indirect Measurements/Analysis - The methodology used to estimate the radioactive concentrations existing in the radioac-tive waste. A representative composite of the waste stream influent and effluent concentrations may be considered a reasonably representative sample for the purpose of performing radiological analysis. For radio-isotopes not readily measured, the establishment of an inferential measurement program to allow the ratios of readily measurable isotopes to those not easily measured, may also constitute an acceptable means of performing radiological content analysis.

  • I PCP-4.0 APR 2 9 1986 2.0 Definitions [continued]

2.5 Wet Wastes - Those forms of radioactive materials normally gener-ated as by-products from the liquid waste processing systems, including spent bead resins and filter elements which contain greater than one half percent by volume of liquids per container.

In addition, wet wastes will also include radioactively contaminated oily wastes that require disposal at a licensed facility.

2. 6 Solidification - Solidification is the conversion of radioactive wet waste t'o a homogeneous, monolithic, immobilized free standing solid, with a definite volume and shape.

2.7 Absorption process - The use of absorbent material to eliminate the presence of free standing liquid in wet wastes.

2.8 High Integrity Container - A container that has been certified to meet the requirements of burial site criteria for waste form stabilization, which may be used in lieu of solidi-fication.

PCP-5.0

\ APR 2 9 1986 3.0 System and Process Descriptions

  • Burial ground requirements for waste stabilization, and elimination of free standing liquids are complied with by processing the wet wastes via the following systems and/or methods:

3.1 Dewatering System The Dewatering System consists of a dewatering container (usually a shipping container), a dewatering pump and associated hose and piping. Station personnel use the following procedures, approved by the Station Nuclear Safety Operating Committee, to transfer and dewater spent resins in preparation for shipment or solidifica-tion:

- Station Procedure OP-20.1.1, Resin Waste System - Transfer-ring spent resin from primary demineralizer to shipping container

- Station Procedure OP-22,9.5, Liquid Waste receiving, filling and dewatering a spent media receiving container The above procedures specify (or reference) the minimum time periods for dewatering pump operation, settling time and subse-quent verification that the waste contains "no detectable free standing liquids" as required by applicable radioactive waste burial ground license conditions.

These procedures also require that each step be initialed, thus insuring that the operating personnel comply with the dewatering procedures and document their completion.

PCP-6.0 APR 2 9 1986 3.0 System Description [continued]

  • 3.2 Solidification System Wet wastes, as defined in Section 2.0, which are classified, pur-suant to 10CFR61.55, as Class B or Class C wastes, shall be solidified prior to burial or disposed of in approved high inte-grity containers.

Surry Power Station currently has no installed solidification system onsite to process wet wastes. In the event that such processing is required, an outside contractor will provide a solidification system to the Station. The contractor shall also furnish to the Station a system description, system operating procedures that specify the process control parameters, and a copy of the topical report or equivalent documentation indicating compliance with 10CFR Part 61 and burial site requirements.

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. 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.

PCP-7.0 APR 2 9 1986 3.0 System Description [continued]

  • 3.2 Solidification System [continued]

The system operating procedures for solidification and a copy of the topical report or equivalent documentation will be submitted to the Station for determination of compliance with the Process Control Program. Upon approval by the Station Nuclear Safety and Operating Committee, the procedures will be incorporated into the appropriate Operations and/or Health Physics Procedures as an attachment or by reference.

The solidification system will not be considered operable until the operating procedures are approved by the Station.

The topical report or equivalent document must be reviewed by the Station prior to the performance of any solidification processing.

3.3 Absorption Processing Wet wastes, including oily wastes, as defined in Section 2.0, meeting the criteria for Class A wastes, as per 10CFR 61.55, may be processed by absorption to eliminate free standing liquids.

Sufficient absorbent shall be used to ensure that the waste contains "no detectable free standing liquids" prior to sealing the container.

Processing of such wastes with absorbents shall be performed in compliance with procedures approved by the Station Nuclear Safety Operating Committee, and only with materials approved by the burial site.

PCP-8.0 APR 2 9 1986 4.0 Waste Sources and Characteristics

  • 4.1 Bead Resin Four (4) systems currently exist processing radioactive liquids.

at Surry Power Station for 4 .1.1 The Liquid Waste Processing System processes radioactive liquids normally associated with primary and secondary leakage which are collected from various sumps/tanks and transferred to the high and low level waste tanks for processing. Liquids collected in these waste tanks are pumped through one or more trains of filters and/or ion exchange resins and transferred to the liquid waste test tank. The liquid waste tank is isolated, recirculated, sampled and analyzed for radioactive content. The tank may then be released to the discharge canal provided the radioactivity is within the release limits.

4.1. 2 The Primary Coolant System is purified by processing the letdown flow through the letdown filters and one of two mixed bed demineralizers. The mixed bed is a combination of anion and cation bead resin, used for removal of fission and corrosion products. A cation bed is used intermittently to adjust primary coolant pH by removal of Li-7 and a deborating demineralizer (anion resin) used to reduce boron concentrations. Primary coolant is passed through a post resin filter, then returned to the system through the volume control tank *

.. PCP-9.0 APR 2 9 1986 4.0 Waste Sources and Characteristics [continued]

4.1 Bead Resin [continued]

4.1.3 The Boron Recovery System processes primary vents and drains for the purpose of recycling or disposal of boric acid and removal of gaseous products. Liquids are pumped from the primary drains tanks to the Gas Stripper System, through the Cesium Removal Ion Exchangers (cation resin),

the Boron Recovery Filters and finally to the Boron Recovery Tanks.

4.1.4 The Spent Fuel Pit Purification System provides a means to cool the Spent Fuel Pit water. The purification system is a closed system, circulating spent fuel pit water through a heat exchanger, a filter element and ion exchanger (anion/cation resin), and discharging back into the fuel pit.

Resins from these various systems are considered "spent" when decontamination factors indicate a significant decrease or when activity levels reach a pre-determined level. Spent resins are transferred to shipping contain-ers using primary grade water.

Resins will remain in the shipping container while the sluice water passes t,hrough a retaining element, dis-charging into the liquid waste system.

Bead resins are normally processed by solidification or dewatering. When solidified, the normal solidification agent is cement. When alternate solidification agents are used, the description shall be provided in the contractors solidification system operating procedures.

' . PCP-10.0 APR 2 9 1986 4.0 Waste Sources and Characteristics [continued]

4.2 Filter Elements Mechanical filters with wound fiber cartridges are used for removing particulate matter from liquid systems. Spent filter elements are removed from systems and placed in storage bunkers to await processing and shipment. The filter elements are encapsu-lated in cement to eliminate free standing liquid prior to place-ment in high integrity containers for stabilization and shipment, as required by 10CFR part 61 and burial site criteria.

4.3 Organic Waste Oil used in systems for cooling and lubrication which comes into contact with radioactive contamination, must be processed to

. ensure compliance with Burial site requirements. Oily waste is generated by system leakage or from normal system replacement and stored until the waste can be processed. Contaminated oily waste shall meet the criteria for Class A, as per 10CFR 61.55 and may be processed, using approved procedures, by absorption or solidifi-cation and sealed in a container meeting burial site requirements.

Solidification equipment, materials and procedures will normally be supplied by a contractor.

('

PCP-11,0

,986

~l>R 2 9 5.0 Sampling, Analysis, and Process Surveillance

  • 5.1 Collection of Samples At least one representative sample from at least every tenth batch of wet radioactive waste, shall be tested to verify solidification and/or absorption.

For collection and handling of high activity wastes, where han-dling of samples could result in personnel radiation exposures inconsistent with ALARA practices, representative nonradioactive samples may be used for solidification testing, Nonradioactive samples shall be as representative as possible to the actual waste with regard to chemical properties.

Samples of actual waste are also used for direct radiological analysis to determine waste classification in accordance with 10CFR 61.55, Samples should be drawn reasonably close to date of shipment or date of processing.

5.2 Analysis Analysis shall be performed on radioactive waste to define process control parameters, prior to waste solidification, The type of analysis to be performed on waste samples will be dependent on the specific solidification operation procedures submitted by contrac-tor services. Results of analysis will be recorded on Waste Solidification Data Sheets, contained in solidification operating procedures, Samples of wet waste, taken to determine isotopic content, shall be analyzed for major gamma emmitters by radioassay techniques in accordance with approved Health Physics Procedures.

PCP-12.0 APR 2 9 1986 5.0 Sampling, Analysis, and Process Surveillance [continued]

  • 5.3 Process Surveillance 5.3.1 If any sample, taken to verify solidification, failsJto solidify, the solidification batch under test shall be suspended until such time as additional samples can be obtained, alternative solidification parameters can be determined in accordance with the Process Control Program or its supporting procedures, and a subsequent test verifies solidification. Solidification of the batch may then be resumed using the alternative solidification parameters determined.

If the initial sample from a batch of waste fails to verify solidification, then representative samples shall be collected from each consecutive batch of the same type of wastes until three consecutive initial test specimens demonstrate solidification. The Process Control Program and/or its supporting procedures shall be modified as required to ensure solidification of subsequent batches of waste.

5.3.2 If the waste is processed by absorption, a test sample shall be obtained from at least one of every ten batches to verify ratios of absorbent to wet waste volumes are adequate to ensure complete elimination of free liquids.

If a test sample used to verify absorption shows free standing liquid, then the batch under test shall be suspended until such time that the procedures can be modified to ensure sufficient absorption to eliminate any free standing liquids.

PCP-13.0 APR 2 9 1986 5.0 Sampling, Analysis, and Process Surveillance (continuedJ 5,4 Acceptance Criteria 5.4.1 Solidification Acceptability The following criteria define an acceptable solidification process and process parameters:

Sample solidifications used to verify batch solidifica-tion will be considered acceptable if there are no visible or drainable free liquids.

Sample solidifications are considered acceptable if upon inspection the waste will retain its shape when removed from the test container.

Solidified containers of waste shall be inspected to insure containers are at least 85% full as required by South Carolina license condition 39c and for proper curing (i.e. Hardness) prior to closing.

5.4.2 Radioassay Acceptability -

The results of the radioassay are considered acceptable, when it has been verified and documented that the waste material is packaged in a container which is acceptable for transportation and burial, considering the radioactivity concentrations which exist in the waste.

Strong-tight containers are acceptable for waste meeting Class A criteria of 10CFR 61,55 with no free standing liquid. High integrity containers or solidification is required for Class Band Class C wastes.

PCP-14.0 APR 2 9 1986 5.0 Sampling, Analysis, and Process Surveillance [continued]

5.4 Acceptance Criteria [continued]

5.4.3 Dewatering Acceptability Procedures for dewatering containers are specified in Section 3 .1. These proce-dures specify the minimum time periods for dewatering pump operation, settling time and subsequent verification that the waste contains "No detectable free standing liquids" (less than 1% by waste volume).

5.4.4 Absorption Acceptability - Samples used to verify proper absorption will be considered acceptable if there are no visible or drainable free liquids.

PCP-15.0 APR 2 9 1986 6.0 Contractor Services/Station Interface and Requirements Currently there is no installed solidification system onsite to process wastes that require solidification. In the event such services are required, a contractor will be requested to submit, for approval, solidification system operating procedures, a list of physical inter-faces, materials required, and a list of expected utility/contractor responsibilities.

The solidification system operating procedures will be reviewed in accordance with the Process Control Program to determine adequate station control and Quality Assurance criteria are met. Once approved by the Station Nuclear Safety Operating Committee, the system operating procedures will be incorporated into, or referenced by existing Health Physics and/or Operations Controlling Procedures.

A copy of the topical report or equivalent documentation indicating compliance with 10CFR Part 61 and burial site requirements must be reviewed prior to performing solidification processing.

PCP-16.0 APR f 9 1986 7.0 Station Records 7.1 Process Documentation Station records shall be maintained, to document that the dewater-ing and solidification process was carried out in accordance with the Process Control Program. Applicable procedures specified in Section 3.0, a copy of the topical report or equivalent documenta-tion provided by the contractor, and the contractor's operating procedures utilized during solidification, will be retained.

Data sheets shall be used to record test sample solidification data. The data sheets may include but are not limited to, type of waste to be solidified, major constituents, pH, waste/liquid/so-lidification agent/catalyst ratios, waste oil content, mixing and curing times.

Data sheets should also include batch number, batch volume, and date processed for each batch solidification and/or absorption.

Station records shall be maintained for any Direct or Indirect Measurements/Analysis performed on the waste material.

PCP-17.0 APR 2 9 1986 7.0 Station Records [continued]

7.2 Changes to Process Control Program Changes to the Process Control Program shall be submitted in the Semi-annual Radioactive Effluent Report. The report shall include detailed information that totally supports the rationale for the change, determination that the change did not reduce the overall conformance of the solidified waste product to the existing criteria for solid wastes, and documentation that it has been reviewed by the Station Nuclear Safety Operating Committee.

Changes to the Process Control Program shall become effective upon review and acceptance by the Station Nuclear Safety Operating Committee

  • PCP-18.0 APR 2 9 1986 8.0 References
  • 8.1 8.2 Surry Power Station Technical Specifications State of South Carolina Radioactive Materials License No. 097, and amendments 8.3 Surry Power Station Operating Procedures 8.4 Branch Technical Position - ETSB 11-3 8.5 State of Washington Radioactive Materials License No. WN-1019-2, and amendments 8.6 10CFR Part 20, Paragraph 311 8.7 10CFR Part 61, Paragraphs 55 and 56

'. ADM-60 OCT 1 1 1988 ATTACHMENT 8 Pagel of 4

    • VIRGINIA POWER PROCEDURE CHANGE REQUEST SURRY POl--1/ER STATION A l)IU  :;.. I- o5 -01 AOM21
1. DEPARTMENT:

Health 'Physics 2.PP.OCSX.ff Na. 3.~ITNo. l &.

2 4.REV. DATE:

HP-7.2.20 New 5.Tlll.E:

PROCESS CONTROL PROr,RAM new rocedure

7. OW'3E REOlJESTED BY: I. DATE:

R Radiation Protection Plan

10. ~ FOA CHANGE; Implementation of Cha ter 7 of the Radiat 11 *SAf'ETYAt-W.YSISREO\MED? S NO IOCFR5D,511 YES NJ IOCFR.72,311 YU NO 0 @ REVIEW RECJJIRED7 0 [) REVIE"lt REQUIREOa lli]

SAFETY ANALYSIS IS REOUIRED FOR >Hf YES ANSWER. 5"FETY AH,l,L.YSIS ~ 10CFR50.5W 10CFR72.35 REVIEW IS REOUIRED FOfl >Hf YES

~ NOOESTICNS 1 nff:03H3,_

"f& NO 1.The facllity, as described in the UFSAR or Technical Specfteations, will be changed.

0 ~ Sections Reviewed: UFSAR 11; Tech. Spec. 5.0, 6.1

'IS NO 2. The procedures or methods of operation, as descnbed in the UFSAR or Technical 0 Ii) Specifications, wiD be changed.

Section, Reviewed: UFSAR 11 & 12; Tech. Spec, 3. 11, 4, 9, 6, 1 , 6. 4 WB NO 3. A test or experiment. which Is not descn'bed In the UFSAR or Technical Specifications, Is D

'11B mNO proposed.

4. Safety-Related structures, systems, equlptment or components, not descn'bed In 1he UFSAR D liJ or Technlcal Specifications, wDI be changed. * '

WIS NO 5.The facility, procedure, or method of operation which could affect equlptment lmportan1 to D lfil safety will be changed.

12. RmfWED: PEER SUB-OOMMITTE 14. DATE:

D VERIFICATa. D VALIDATION /2- ~ '&!

15JIECQ,(MENOED ACTIOtt

@ APPRO\IEO D DISAPPROVED 1a.REVIEWED BY STATIOO IIJClfAA SAFETY At<<, OPEAATNO OOMJ.IT1U:

(Z) APPROVED D APPROVED >S MOOFEO

21. N E W ~ RFv'ISON OATI::

(1)

PROCEDURE NUMBER: HP-7.2.20 SURRY POWER STATION (2)

DATE: DEC 0 8 1988 (3) (4)

TYPE PROCEDURE: UNIT:

Health "Physics 1 & 2 (5)

TITLE: PROCESS CQNIEQI, "PEQGE~M (6)

CONTENTS SECTION

1. PURPOSE
2. REFERENCES
3. PRECAUTIONS AND LIMITATIONS
4. INSTRUCTIONS
  • 5.

6.

ATTACRME't-.TTS FOR"'1S T HINK ALAR A' RECOMMEND APPROVAL: (7) (8)

  • G-f?~ DATE: ;l-/6cf'/?Y APPROVED STATION NUCLEAR SAFETY ~~ERATING COMMITTEE: (9) (10) 1j t

// ./'1,~ .

1- '

'--~

I/

..l-1 DATE: liEC O 8 19" fl APPROVED (MANAGER) (If Required): (11) (12)

  • AJtA DATE:

j -

i HP-7.2.20 Page l of 6 SURRY POWER ST ATION UNITS I & 2 PROCESS CONTROL PROGRAM 1.0 PURPOSE I. I 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 Plan. Chapter VII, "Radioactive Material and Effluents Control," Section 2, "Solid Radioactive Waste Control."

2.2 IOCFR6I, Licensing Requirements for Land Disposal of Radioactive Waste."

2.3 Code of Federal Regulations, Title 49, Transportation. Parts 100 to 199.

2.4 Surry Power Station Technical Specifications; 1.0.P, 3.11.E, basis for Solid Radioactive Waste, 4.9.K, 6.6.B.3, and 6.8.A.

2.5 US NRC Low:Level Waste Licensing Branch, "Technical Position on Radioactive Waste Classification" and "Technical Position on Waste Form," May 1983, Rev 0.

2.6 US NRC, Standard Review Plan 11.4, "Solid Waste Management Systems" Rev 2, July 1981 (NUREG-0800).

3.0 PRECAUTIONS AND LIMIT ATIO NS 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 6.8.A, including changes being submitted to the NRC as part of the Semiannual Radioactive Effluent Release Report. -

3.3 The PCP consists of instruction 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 ensure 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.

HP-7.2.20 Page 2 of 6 DEC O 8 1988 3.5 Attachment 3, "Definitions and Additional Descriptions" provides definitions of terms and other description or consideration specific to the PCP. Procedure users should be familiar with ,he contents of this attachment.

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
e. 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 used 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 Classification of Radioactive Waste 4.2.1 Processed wet waste shall be classified in accordance with IOCFR61.55 .

  • 4.2.2 Classification of wet waste shall be in accordance with procedure HP-7.2.21, "Sampling, Analyzing and Classifying Solid Radioactive Waste."

HP-7.2.20 Page 3 of 6 DEG O 8 i:l38 4.3 Processing Wet Radioactive Waste 4.3.l Ion excbange 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.

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 solidif icatiori 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 Implement the Process Control Program

  • 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."

4.4.4 If solidification of wet waste is applicable, it shall be performed in accordance with Attachment I, "Solidification of Wet Waste."

4.4.5 If filter elements are to be stabilized by encapsulation, it shall be consistent with the procedure provided in Attachment 2, "Dewatering and Encapsulation of Filter Elements."

HP-7.2.20 Page 4 of 6 DEC O S :~OS 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 I, "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 the Station Nuclear Safety Operating Co~mittee 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.

i_

.. HP-7.2.20 Page 5 of 6 5.0 ATTACHMENTS 5.1 Attachment !;_"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 6.1 None

HP-7.2.20 Attachment 1 Page 1 of 3 DEC O 8 198~

SOLIDIFICATION OF WET WASTE NOTE: Surry 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 Program Solidification Parameters 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.

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 Sampling. Analysis. and Process 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 waste 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.

a. 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 samples can be obtained, alternative solidification parameters can be determined.
b. Subsequent test(s) must 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.
1. Obtaining test specimen shall continue until three consecutive initial test specimen demonstrate solidification.

2.2.4 If required, the Process Control Program and/or applicable procedures shall be revised to ensure solidification of subsequent batches of waste.

.. HP-7.2.20 Attachment I Page 2 of 3 2.3 Samples of waste shall be obtained in accordance with procedure HP-7.~'.tI.°"~ariigtng, Analysis and Classification of Solid Radioactive Waste" as required to determine 10CFR6 l 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 number, batch volume, and date processed for each batch solidification and/or absorption. -

4.0 Solidification Acceptance 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 as 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. -

4.3 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 with NRC BTP C.2 (stability guidance) or other evaluation criteria specifically approved by the _NRC or site license.

a. If evaluation is required, refer to applicable documents as indicated.

l HP-7.2.20 Attachment 1 Page 3 of 3 DEC 0 3 1SS8 4.3.2 Solidified Class A liquids shall meet the requirements of NRC BTP C.l:

a. If Class 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.

HP-7.2.20

' Attachmeni. 2 Page 1 of 2 DEC D 3 1S6a

  • DEWA TERING AND ENCAPSULATION OF FILTER ELEMENTS NOTE: Filter elements are normally mechanical filters with wound fiber cartridges used for removing particulates 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.

NOTF.: 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 n uclides with half -

lives greater than 5 years which have a total specific activity of 1 uCi/cc or greater, it may be disposed of as Class A was~e.

  • 2.2 If filter media is classified as Class *B or Class C waste (per 10CFR6 l.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

  • Filter Elements to be Disposed of as Class 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:

3.3.1

  • There shall be no indication of moisture on the filter in the form of drops or surf ace 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 he equal to at least one-fourth the volume of the filter.

3.4 Ensur.e the documentation indication package contents describes the presence of the filters.

4.0 Filter Elements to be Disposed 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 wire is recommended to hold the filter(s) in a fixed geometry.

(

.' HP-7.2.20 Attachment 2 Page 2 of 2 DEC O 8 1388 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 Introduce the solidification media into the container and fill the container to completely cover the f ilter(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, "Packa.g~ng 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-encapsulated 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 held 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 Dewater the container as applicable in accordance with procedure HP-7.2. 70, "Packaging Radioactive Waste."

'I HP-7.2.20

' Attachment 3 Page I of 3 DEC O 8 i999

  • DEFINITIONS AND ADDITIONAL DESCRIPTIONS 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 Program (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 IOCFR20, IOCFR 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste (Tech Spec definition).

Stabilization or Stabflity 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 dispos:i I (10Cf R61.56(b).

Solidification Solidification shall be the conversion of wet waste into a form that meets shipping and burial ground req*uirements (Tech Spec definition).

Free Liquid 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 11.4, ETSB 11-3).

Hf-7.2.20 Attachment 3 Page 2 of 3 DEC O 8 1988

  • A quantity of waste which has been mixed or may be mixed to produce a homogeneous mixture for the purposes of sampling, testing, and processing. Different samples of the homogeneous mixtui;e 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 which fills one disposable liner or drum.

Test Specimen 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.

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.

Spent Ion Exchange Resi_ns Resins are considered spent when decontamination factors indicate a significant decrease or when activity levels reach a pre-determined level.

Non-Corrosive Liquid 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).

High Integrity 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 C.4 for more details if desired. High integrity containers must be approved by the appropriate agency.

Primary Coolant and Chemical and Volume Control Systems Reactor coolant is purified by processing the letdown flow through a letdown filter, a mixed bed demineralizer, a cation demineralizer (optional), deborating demineralizer (optional), and a reactor coolant filter to the volume con i rol tank. From the volume contr_ol tank, part of the coolant is recycled to the reactor via charging pumps. Part of the coolant is rou_ted 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 ion 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 Spent Fuel Pit Purification System DEC O 3 i988 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.

Liquid Waste Processing 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.

Spent Resin Transfer System Spent resins are transferred (flushed) to intermediate holding tanks or shipping containers for processing as radioactive waste. Resins will remain in the shipping container while the sluice water passes through a retaining element, discharging into the liquid waste system.

Dewatering System The Dewatering System consists of a dewatering container (usually a shipping container), a dewatering pump and associated hose and piping. Station personnel use procedures to transfer and dewater spent resins in preparation for shipment and/or solidification. The procedures specify (or reference) the minimum time periods for dewatering pump operation, settling time and subsequent verification that the waste contains "no detectable free standing liquids" as required by applicable disposal site criteria .

Attachment 5 Page 1 of 1 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS There were no major changes to Surry's Radioactive Liquid, Gaseous or Solid Waste Treatment Systems during the period of July l, 1988 through December 31, 1988.

Attachment 6 Page 1 of 1

  • INOPERAJBILITY OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENT MONITORING INSTRUMENTATION Technical Specification 3. 7.E.2 requires the Semi-Annual Report to include an explanation why monitors required by Technical Specification Tables 3. 7-5(a) and 3. 7-5(b) which were determined inoperable, were not returned to operable status within 30 days.

Two monitors require explanation under this criteria for the period of July 1, 1988 through December 31, 1988. They are the Component Cooling Service Water Monitor and the Waste Gas Holdup System Explosive Gas Oxygen Monitor.

1. The Component Cooling Service Water Monitor (RM-SW-107) continues to be inoperable as described in the previous Semi-Annual Report.

A study was performed during the fall of 1988 to determine if grab sampling and monitoring the Component Cooling Surge Tank level would be an acceptable permanent alternative to returning this monitor to service. The study concluded that this approach is not desirable.

The merits of in-line, on-line and off-line radiation monitors were discussed by System Engineering and Power Engineering Services with two vendors. The in-line and on-line monitoring concepts were abandoned on the basis that contamination on the tube side of the heat exchanger might be mistaken for Service Water Contamination. The off-line monitor, as in the original design, is thought to be the most viable option. Failure of the original design may have been due to improper sizing of sample lines. A special test will be performed with larger sample lines. Due to the current double outage and the lack of required materials, the test is not likely to begin before this reporting period closes on June 30, 1989. Once installation is complete, the test should simulate normal operating conditions for 60 days followed by a test evaluation.

Grab sampling, as required by Table 3. 7-5(a) when the monitor is out of service, has been performed since the monitor became inoperable. The grab sampling will continue until this matter is resolved.

2. The Waste Gas Holdup System Oxygen Monitor continues to be inoperable. As described in the previous Semi-Annual Report, the monitor cannot be calibrated within the tolerance of Technical Specifications. A new calibration method has been supplied by the monitor manufacturer, however, a calibration has not been performed due to maintenance associated with the Waste Gas Surge Drum. When maintenance is completed, the monitor will be calibrated and returned to service.

Grab sampling as required by Technical Specification Table 3. 7-5(b) when the monitor is out of service, has been performed since the monitor became inoperable. Grab sampling will continue until the monitor is returned to service. ~

Attachment 7 Page 1 of 1

  • UNPLANNED RELEASES There has been no Unplanned Liquid or Gaseous Releases that exceeded Technical Specification 3.11.A. l.a and 3.11.B. l.a during the period July 1, 1988 through December 31, 1988 .

't Attachment 8 Page 1 of 1 LOWER LEVEL OF DETECTION FOR EFFLUENT SAMPLE ANALYSIS GASEOUS: Isotope Required LLD Typical LLD (uCi/ml) (uCi/ml)

Kr-87 1.00 E-4 4.31 E-7 - 6.37 E-7 Kr-88 1.00 E-4 3.79 E-7 - 4.46 E-7 Xe-133 1.00 E-4 2.39 E-7 - 2.58 E-7 Xe-133m 1.00 E-4 1.03 E-6 - 1.58 E-6 Xe-135 1.00 E-4 1.30 E-7 - 1.52 E-7 Xe-135m 1.00 E-4 2.63 E-6 - 3.05 E-6 Xe-138 1.00 E-4 6.27 E-6 - 1.05 E-5 I-131 1.00 E-12 I. 9 2 E 3. 6 8 E-13 Sr-89 1.00 E-11 9.00 E 8.00 E-13 Sr-90 1.00 E-11 1.00 E 9.00 E-14 Cs-134 1.00 E-11 1.46 E 2.45 E-13 Cs-137 1.00 E-11 2.87 E 1.09 E-12 Mn-54 1.00 E-11 1.29 E 2.29 E-13 Fe-59 1.00 E-11 4.58 E 6.10 E-13 Co-58 1.00 E-11 1.45 E 2.72 E-13 Co-60 1.00 E-11 3.45 E 4.16 E-13 Zn-65 1.00 E-11 3.09 E 4.38 E-13 Mo-99 1.00 E-11 1.13 E I. 7 4 E-12 Ce-141 1.00 E-11 1.26 E 1.66 E-13 Ce-144 1.00 E-11 4.97 E 7.47 E-13 Alpha 1.00 E-11 1.53 E 5.57 E-14 Tritium 1.00 E-6 8.18 E-8 - 1.03 E-7 LIQUID: Sr-89 5.00 E-8 3.00 E-8 - 5.00 E-8 Sr-90 5.00 E-8 2.00 E-9 - 1.00 E-8 Cs-134 5.00 E-7 4.53 E-8 - 6.22 E-8 Cs-137 5.00 E-7 5.61 E-8 - 2.06 E-7 I-131 1.00 E-6 3.09 E-8 - 5.17 E-8 Co-58 5.00 E-7 3.01 E-8 - 5.00 E-8 Co-60 5.00 E-7 5.98 E-8 - 7.81 E-8 Fe-59 5.00 E-7 7.89 E-8 - 1.07 E-7 Zn-65 5.00 E-7 5.39 E-8 - 8.77 E-8 Mn-54 5.00 E-7 2.4 7 E-8 - 4.16 E-8 Mo-99 5.00 E-7 2.00 E-7 - 3.23 E-7 Ce-141 5.00 E-7 3.33 E-8 - 3.87 E-8

  • Ce-144 5.00 E-7 1.42 E-7 - 1.77 E-7 Fe-55 1.00 E-6 1.00 E-6 - 1.00 E-6 Alpha 1.00 E-7 8.81 E-9 - 1.37 E-8 Tritium 1.00 E-5 2.25 E-6 - 2.83 E-6 Xe-133 1.00 E-5 6.51 E-8 - 7.98 E-8 Xe-135 1.00 E-5 3.10 E-8 - 3.46 E-8 Xe-133m 1.00 E-5 2.76 E-7 - 3.23 E-7 Xe-135m 1.00 E-5 4.89 E-7 - 6.40 E-7 Xe-138 1.00 E-5 1.40 E-6 - 2.14 E-6 Kr-87 1.00 E-5 8.72 E-8 - 1.22 E-7 Kr-88 1.00 E-5 8.36 E-8 - 1.16 E-7
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