ML20082M464

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Effluent & Waste Disposal Semiannual Rept for Second Quarter 1991
ML20082M464
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 06/30/1991
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20082M453 List:
References
NUDOCS 9109050217
Download: ML20082M464 (32)
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Text

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CALVERT CLIFFS NUCLEAR P0n'dR PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT  :

SUPPLEMENTA1_lNFORMATION Facility - Calvert Cliffs Nuclear Power Plant Licensee - Baltimore Gas & Electric Company i

I. REGULATORY LI. TILTS  !

l A. Fission and Activation Gases  ;

1. The instantaneous release rate of noble gases in gaseous effluents shall not result- in a site boundary dose rate greater than 500  ;

mrem / year to the whole body or greater than 3000 mrem / year to the ,

skin (Technical Specification 3/4.11.2.1). j

2. Gaseous Radwaste Treatment System and the Ventilation Exhaust 1reatment System shall be used to reduce gaseous emissions when i the calculated gamma dose due to gaseous effluents exceeds 1.20 t mrad or the calculated beta dose due to gaseous effluents exceeds I 2.40 mrad at the site boundary in a 92 day period (Technical  ;

Specification 3/4.11.2.4).  ;

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3. The air dose at the site boundary due to noble gases released in gaseous effluents shall not exceed (Technical Specification 3/4.11.2.2): 1 10 mrad /qtr, gamma air

[

20 mrad /qtr, beta air f

20 mrad / year, gamma air j 40 mrad / year, beta air f

4. All of the above parameters are calculated according to the methodology specified in the Offsite Dose Calculation Manual j (0DCM).

B. Iodines and Particulates with Half Lives Greater than Eiaht Days

1. The instantaneous release rate of iodines and particulates in  !

gaseous effluents shall not result in a site boundary dose in 't excess of 1500 mrem / year to any organ (Technical Specification 3/4.11.2.1). I

2. The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in ,

gaseous effluents when calculated doses exceed 1.8 mrem to any organ in a 92 day period at or beyond the site boundary (Technical Specification 3/4.11.2.4). '

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ADOCK 05000317 1 i PDR  !

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3. The dose to a member of the public at or beyond the site boundary from iodine-131 and particulates with hal f lives greater than eight days in gaseous effluents shall not exceed (Technical Specification 3/4.11.2.3):

15 mrem /qtr, any organ 30 mrem / year, any organ less than 0.1% of the above limits as a result of burning contaminated oil.

4. All of the above parameters are calculated according to the methodology specified in the ODCM.

C. Liould Effluents ,

1. The concentrations of radionuclides in liquid effluents from the plant shall not exceed the values specified in 10 CFR Part 20, Appendix B, for unrestricted areas (Technical Specification J/4.11.1.1). l
2. The liquid radwaste treatment system shall be used to reduce the concentration of radionuclides in liquid effluents from the plant when the calculated doses to unrestricted areas exceed 0.36 mre, to the whole body, or 1.20 mrem to any organ in a 92 day period (Technical Specification 3/4.11.1.3).
3. The dose to a member of the public in unrestricted areas shall not exceed (Technical Specifcation 3/4.11.1.2):

3 mrem /qtr, total body 10 mrem /qtr, any organ  :

6 mrem / year, total body 20 mrem / year, any organ

4. - All of the liquid dose parameters are calculated according to the methodology specified in the 00CM.

!!. MAXIMUM PERMISSIBLE CONCENTRATIONS A. Fission and Activation Gases Prior to the batch release of gaseous e f fl uents , a sample of the source is collected and analyzed by gamma spectroscopy for the principal gamma emitting radionuclides. The identified radionuclide concentrations are evaluated and an acceptable release rate is determined to ensure that the dose rate limits of Technical Specification- 3/4.11.2.1 are not exceeded.

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B. lodines anJ_hr.liculates with Half L]yn. Greater than Eiaht ~1a_y1 Compliance with the dose rate limitations for iodines and particulates is demonstrated by analysis of the charcoal and particulate sanples of the station main vents. The charcoal samples are analyzed by gamma spectroscopy for cuantification of any release of radiciodines. The particulate samfes are analyzed by gamma spectroscopy for quantification oi particulate radioactive material. Based on guidance provided in the ODCM, compliance with dose rate limits for the radiciodines and particulates may be based on a comparison of the actual measured release quantity over a sample period with h pre-established upper bound. '

C. Liquid Effluents The MPCs used for radioactive materials released in liquid effluents are in accordance with TLchnical Specification 3/4.11 l.1 and the values from 10 CFR 20, Appendix B, including applicable _ table notes.

In all cases, the more restrictive (lower) MPC found for each radionuclide is used regardless of solubility.

III.HCHNICAL. SEECIFICATION BIfLOKU10_.RDLUIREENTS Section 6.9.1.8)

A. Prevjpftn Calendar Year H210) Dose Assessment Summary During 1990 liquid releases from Calvert Cliffs resulted in a calculated maximum organ dose of 3.02E-01 mrem and a maximum whole

' ody dose of 2.19E-02 mrem.

o These doses are less than 4% of the .

Technical Specification vearly organ dose limit and less than 1% of the Technical Specification yearly dose limit for the whole body.

These doses were calculated using ODCM methodology. The controlling pathway was the fish and shellfish pathway with adul t as the controlling age group, and the Gastro-intestinal Tract representing the organ with the highest calculated dose.

Gaseous releases of noble gases resulted in a maximum air dose of 1.78E-02 mrad, gamma and 5.08E-02 mrad, beta lodine and particulate releases from Calvert Cliffs resulted in a maximum organ dose of  :

2.38E-03 mrem for the year via the milk-infant-thyroid pathway. These doses were calculated using ODCM methodology. For 1990, calculhted of fsite doses 'via the gaseous release pathways were below 3% of their allowable Technical Specification limits.

B. 40 CFR 190 Total Dose Compliaqqe Based upon all releases for 1990 and the ODCM calculations, the

'naximum exposed individual would receive less than 1% of the allowabie dose. During 1990, there were no'on-site sources of direct radiation l that would have contributed to a significant or measurable offsite dose. The direct radiation contribution is measured by both on-site and off-site thermoluminescent dosimeters (TLD). The results of these ,

measurements did not indicate any statistical increase in off-site radiation doses attributable to on site sources. There fore, no i increase in the off-site calculated doses is attributable to the ,

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direct exposure f rom on-site sources. A more detailed evaluation will I be reported in the Annual Radiological Environmental Monitoring '

Report.  :

C. 1911d.3mte Report Recujrementt During the first half of 1991, a types of radioactive solid waste f shipped from Calvert Cliffs 'e radioactive resin, which was  :

dewatered and shipped in high integrity containsrs and dry [

compressible waste, which was shipped as LSA waste in strong, tight  !

containers. Also, Appendix A provides a detailed breakdown of the -

waste shipments for 1991 per the categories as specified in Technir.al Specification 6.9.1.8.

D. ODCM and PCP Chana n Two changes were made to the ODCM during the first half of 1991. The changes were reviewed by p0SRC and approved by the Plant Ger.eral  !

j. Manager, Calvert Cliffs Nuclear Power Plant, prior to implementation. ,

The scope and insis for this change is discussed in Appendix B. In [

keeping with the requirement of the Technical Specification 6.17, a  ;

copy of the change to the CCNPP ODCM is enclosed in Attachment 1.

Vertical lines in the right margin of the text denote the above  !

referenced change with accompanying change number. j No changes were made in the PCP in the first half of 1991. j IV, 81L}LAjLGBD1 Not Applicable f

V. li[!LS11RBENTS AND APPROXJMATlDRS_AND TOTAL RAD 10AQTIVITY g A. Fission i nd Activittion Gases

1. Batch Releases l.

Prior to each batch release of gas from a pressurized gas decay  !

tank, a sample is collected and analyzed by gamma spectroscopy 4 using a Ge detector for the principal gamma emitting noble gas i radionuclides. The total activity released is based on the  ;

pressure / volume relationship (gas laws) of the tank.

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Prior to and after each containment purge, a gas sample is

,. collected and analyzed by gamma spectroscopy using a Ge detector ,

l. for the principal gamma emitting noble gas radionuclides. The  !

i total activity released is based on containment volume and purge  ;

l rate. Activity buildup while purging is also considered. i

2. Continuous Releases e

A gas sample is collected at least weekly from the main vents and l analyzed by gamma spectroscopy using a Ge detector for the i principal gamma emitting noble gas radionuclides. The total j activity released for the week is based on the total sample i r

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activity decay corrected to the midpoint of the sample period multiplied by the main vent flow for the week.

A monthly composite sample is collected from the main vents and analyzed by liquid scintillation for tritium. The total tritium release for the month is based on this sample analysis and the vent flow.

B. loJine and Particulatet

1. Batch Releases The total activities of radiciodines and particulates released from a pressurized gas decay tank, containment purges and containment vents are accounted for by the continuous samplers on the main vent.
2. Continuous Releases i During the release of gas f rom the ma-in_ vents, samples of iodines and particulates are cellected using a charcoal and particulate filter, respectively. The filters are removed weekly and are analyzed by gamma spectroscopy using a Ge detector for significant gamma emitting radionuclides. The total activity released for the week is based on the total sample activity decay corrected to the midpoint of the sample period multiplied by the main vent flow for the week. These weekly particulate filters are then composited to form monthly and quarterly composites for the gross alpha and strontium 89 and 90 analyses.

C. Linuid Effluent s

1. Batch Releases i

Prior to the release of liquid from a waste tank, a sample is collected and analyzed by gamma spectroscopy for the principal I gamma emitting radionuclides. To demonstrate compliance with the requirements addressed in Section 1.C.1 above, the measured radionuclide concentrations are compared with the allowable MPCs; dilution in the discharge conduit is considered, and an allowable release rate is verified.

The total activity released in each batch is determined by  :

multiplying the volume released by the concentration of each  ;

The actual is radionuclide. volume released based on -the difference in tank levels prior to and after the release. A l proportional composite sample is also withdrawn for each release and this is used in turn to prepara monthly and quarterly j composites for the gross alpha, strontium 89 and 90, and tritium l analyses, i 1

2. Continuous Releases [

Prior to discharge of any continuous releases, a sample is i collected and analyzed by gamma spectroscopy for the principal [

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gamma emitting radionuclides. The measured radionuclide concentrations are compared with the allowable MPC concentrations in the discharge conduit, and an allowable release rate is verified.

When steam generator blowdown is discharged to the circulating water conduits, it is sampled daily and these samples are used in turn to prepare a weekly blowdown composite based on each day's blowdown. The weekly composite is analyzed by gamma spectroscopy for the principal gamma emitting radionuclides. These results are multiplied by the actual quantity of blowdown to determine the total activity released. The weekly composite is also used to prepare monthly and quarterly composites for tritium, gross alpha, and strontium 89 and 90 analyses.

During primary to secondary leakage, the -secondary system becomes contaminated and subsequently, contaminates the turbine plant sumps. This low level activity (mostly tritium) water is released directly to the Chesapeake Bay, 1his water is sampled -at least-weekly and analyzed, by gamma spectroscopy, for the- principal gamma emitting radionuclides, and added to a composite based on the amount released during that week. This monthly composite is analyzed, by gamma spectroscopy, for the principal gamma emitting radionuclides. These results are multiplied by the actual quantity of liquid released to determine the total activity released. This monthly composite is also used to prepare monthly and quarterly composites for tritium, gross alpha, and strontium 89 and 90 analyses.

D. Estimation of Total Error Total error on all releases was estimated using as a minimum the random counting error associated with typical relenses. In addition to this random error the following systematic errors were also examined:

1. Liquid a) Error in volume of liquid released prior to dilution during batch releases.

l b) Error in volume of liquid released via steam generator blowdown.

c)- Error in amount of dilution water used during the reporting period, 1

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2. Gases a) Error in main sent release flow, b) Error in sample flow rate, c)- Error in containment purge relecee flow, d) Error in gas decay tank pressure.

Where errors could be estimated they are usually considered additive.

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I VI. BATCH RELEASES  :

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IST 2ND i QUARTER OVARJJJ ,

l A. 1.iouid j

1. Number of batch releases 5.40E+01 2.80E+01
2. Total time period for batch  !

releases (min) 1.22E+04 1.llE404 3 Maximum time period for a -

batch release (min) 9.19E402 1.30E+03

4. Average time period for batch releases (min) 2.26E402 3.96E402 i i
5. Minimum time period for a  !

batch release (min) 1.00E+00 2.30E401 '

6. Average stream flow during periods of effluent into a flowing stream ,

, (liters / min of dilution water) 3.84E406 4.16E+06 B. _ Gaseous

1. Number of batch releases 2.40E401 2.30E+01
2. Total time period for batch releases (min) 1.02E405 6.04E+04
3. Maximum time period for a br.tch release (min) 1.39E+04 1.02E+04
4. Average time period for batch  :

release (min) 4.26E+03 1.83E403

5. Minimum time period for a batch  !

release (min) 1.00E400 1.00E+00 i

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VII. ABNORMAL RELEASES 1991 IST 2ND QL!AJJJB OVARTER A. Ligu_id

1. Number of releases 2. Total activity released (Curies) I l

l B. Caseous j l

1. Number of releases '
2. Total activity releases (Curies) 9

I!AjlE 1A - REG GUJRLidl CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 GASLOUS (fBUEffTS - SUtit!bil0N OF ALLM1 EMES ,

l IST 2ND EST. TOTAL A. FIS$10N AND ACTIVATION GASES UNITS QUARTER QUARTER ERROR %

1. lotal Release Ci 7.07E402 6.95E+02 6.20E+00 i
2. Average relcase rate for period uCl/sec 8.99E401 8.84E401  ;
3. Percent of tech. spec. limit (1)  % 2.60E-03 3.20E-02 l
4. Percent of tech. spec. limit (2)  % l.07E-03 1.27E 03 l
5. Percent of tech. spec. limit (3)  % 2.10E-01 1.90E 01 i
6. Percent of tech. spec. limit (4)  % 1.05E 01 9.50E-02
7. Percent of tech. spec. limit (5)  % 2.70E 01 1.32E 02
8. Percent of tech. spec limit (6)  % 1.35E-01 6.60E-02 I B. 10 DINES  !

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1. Total Iodine - 131 Ci 4.79E 03 1.75E-02 16. f,0E4 00 l
2. Average release rate for period uC1/sec 6.09E 04 2.22E-03
3. Percent of tech. spec. limit (7)  % 5.82E-04 2.26E-03 {
4. Percent of tech. spec. limit (8)  % 3.20E-01 6.27E-01
5. Percent of tech. spec. limit (9)  % 1.60E-01 3.13E-01 C. PARTICULATES

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1. Particulates with half lives j greater than 8 days Ci 4.S G 07 (10) 12.80E401 l
2. Average release rate for period uCi/sec 6.33E-08 -
3. Percent of tech. spec. limit (7)  % 9.38E-07 -
4. Percent of tech. spec. limit (8)  % 9.15E-03 -
5. Percent of tech. spec. limit (9)  % 4.58E-03 -

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6. Gross alpha radioactivity Ci (10) -

6.54E+01 f i

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... s, IA01E 1 A - RLGJJRE_Lll_(G2 Dbl CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 l r

GASE035 EFFLUENTS - SUMMATION OF ALL.RE.LB1El  !

i IST 2ND EST. TOTAL f D. TRITIUM .. UNITS QUARTER OVARTER ERROR.% i

1. Total Release Ci 5296E400 2.PCE-01 1.32E401
2. Averaae release rate for period u Ci/sec 7.58E-01 3.64E 02 NOTES TO TABLE 1A  !

(1) Percent of I.A.1 whole body dose rate limit (500 mrem / year)

(2) Percent of 1.A.1 skin dose rate limit (3000 mrem / year) f (3) Percent of I.A.3 gamma quarterly dose limit (10 mrad) f (4) Percent of I.A.3 gamma yearly dose limit (20 mrad) j (5) Percent of I.A.3 beta quarterly dose limit (20 mrad)

(6) Percent of I.A.3 beta yearly dose limit (40 mrad)  ;

(7) Percent of 1.B.1 organ dose limit (1500 mrem / year) f r

(8) Percent of 1.B.3 quarterly dose limit (15 mrem) i (9) Percent of 1.B.3 yearly dose limit (30 mrem)  !

I (10) Less than minimum detectable activity which meets the LLD requirements of  !

Technical Specification Surveillance Requirement 4.11.2.1.2. }

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IABLE IC - PEG GUIDE 1.21 CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL .1.'ll-ANNUAL REPORT FIRST Half - 1991 DASE005.EEELUENTS - GROUND LEVELfLCLE&if]

CONTINUOUS MODE DATC!! H0DE IST 2ND IST 2ND 1 FISSION AND ACT1YAT_ Loll 0A1ES Ul(11$ OVARTER OVARIJR OVARTER OVAkTER Arnon -41 Ci (2) (2) (2) 1.49E-03 Krypton 85 C1 (2) (2) 2.10E+00 2.64E400 Krvoton -85m _Ci 1.98E401 9.30E-01 (2) 1.68E 04 Krvoton -87 Ci (2) (2) (2) (2)

Krvton -88 Ci (2) (2) (2) (2)

Xenon -131m Ci 6.56E400 (2) 9.95E-01 3.30E400 Xenon - l]) Ci 4.78E402 1.55E402 1.80E402 5.12E402 Xenon -133m Ci (2) (2) 4.95E 02 9.02E-01 Xenon -135 _JCi 2.72E401 1.50E401 4 12E-01 5.46E-02 Xenon -138 Ci (2) (2) (2) (2)

Total for Period Ci 5.13402 1.71E102 1.83E402 5.19E402

2. HALOGENS lodine -131 Ci 1.38E-03 3.26E;03 (1) (1)

Iodine -133 Ci 3.40E-03 1.42E-02 (1) (1)

Total for Period Ci 4.79E-03 1. 't 5E !>2 (1) (1) l 12 i

.* ., r lAXE 10 -RLG__fdllQE 1.21 LC.oJLQ CALVERT Cliffs NUCLEAR POWER PLANT i EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 MSIOU5lEELUENTS - GROUNQ_LE1(( JEj.EAS[$

CONTINU0US MODE BATCH N0DE ,

IST 2ND 1ST 2ND

3. PARTICULATES Ut])TS QVARTER OVARTLR QUARTER QUARTER i Mancanese -54 Ci (2) .R ) (1) (1) fron -59 .

Ci (2) (EJ (1) (1) ,

Cobalt -58 Ci (2) (2) (1) (1) i Cobalt 60 C1 (2) (2) (1) _ _ . _

(1) ,

Zinc. -65 Ci (2) (2) (1) (1)  ;

i- i j Strontium -89 Ci (1) (3) (1) (1)

Strontium -90 Ci 4.97E-07 (3) ll) (1)  !

Molybdenum -99 _ Ci (2) (2) (1) (1) i Cesium -134 Ci (2) (2) (1) (1)  :

, _,_ Cesium -137 Ci (2) (2) (1) (1)

  • Cerium -141 C_i (2) (2) (1) (1)  !

t i Cerium -144 Ci (2) (1) (J1 (1)

Total for Ppf.iod Ci 4.97E-07 -

(1) (1)  ;

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? l (1) Iodines and particulates in batch releases are accounted for with the main vent continuous samplers when the release is made through the plant main vent.

(2) Less than minimum detectable activity which meets the LLD requirements I of Technical Specification Surveillance Requirement 4.11.2.1.2. ,

(3) The second quarter strontium results will be available in a supplemental report as soon as : -

the activity values are available.

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IARLL2A - REGJUIDE 1.21 CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 LlQUJD EFFLUENTS - SUMMATIONE_All RELEASES IST 2ND EST. TOTAL A. FIS110N AND ACTIVATION PRODUCTS Ut{lTS QUARTER QUARTER ERROR.%

1. Total Release (not including t ri t ium.aa ses . alph a l Ci 5.40E 01 3 05E-01 1.93E401
2. Average diluted concentration durino period uti/ml -4.43E-09 2.04E-09 l
3. Percent of tech. spec. limit (l)  % 1.27E+00 3.90E-01 i
4. Percent of tech. spec limit (21  % 6.35E-01 1.95E-01

}. Percent of tech. spec. limit (3)  % 3.00(:.01 1.66E-01

6. Percent of tech. spec. limit (41  % 1.50E-01 8.30E-02 l l

B. TRITIUM f

1. Total Release Ci 5.24E+01 7.941401 dl.04E401 [
2. Average diluted concentration i durina period uti/ml 4.30E-07 5.29E 07 [

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3. Percent of aonlitable limit (5)  % 1.43E-02 1.76E-02 i C. DISSOLVED AND ENTRALNED GASES
1. Total Release Ci 2.44E-01 1.71E+01 4.80E+00 l 1
2. Average diluted concentration j durina period uti/ml 2.01E-09 1.14E-07 l l

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TABLE 2A - RG_ fit)lDE 1.21_IContJ CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 L10VID EFFLtJ!fNTS - SUMMATION OF ALL RELEASU IST 2ND EST. TOTAL D. GROSS ALPHA RADI0 ACTIVITY UNITS OU_ARIER OVARTER ERROR.% .,_

l. Total Release . Ci (6) 3.20E-05 3.34E401 E. VOLUME OF WASTE RELEASES (prior to dilution) liters 3.36E406 3.41E406 1.30E+00 l 1

F. VOLUME OF DILUTION WATER USED DURING PERIOD liters 1.43E+11 2.04E+11 1.64E+01 i

NOTES TO TABLE 2A (1) Percent of I.C.3 Quarterly Organ Dose Limit (10 mrem) l l

(2) Percent of 1.C.3 Yearly Organ Dose Limit (20 mrem)

(3) Percent of I.C.3 Quarterly Whole Body Dose Limit (3 mrem)  !

(4) Percent of I.C.3 Yearly Whole Body Dose Limit (6 mrem) l (5) Limit used is 3 x 10-3 uCi/ml  :

(6) Less than minimum detectable activity which meets the LLO requirements i of Technical Specification Surveillance Requirement 4.11.1.1.1. I i

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IMIE_JB - REG GU.lR L L21 CALVERT CLIFFS NUCLEAR POWER PLANT  :

EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT ,

FIRST HALF - 1991 .

LIQUID EFFLUENTS i

CONTINUOUS MODE BATCH H0DE IST 2ND IST 2ND NUCLIDES RELEASED UNITS 00ARTER OVARTER OVARTER OUARTER 19_dium -24 C1 (1) (1) 1.25E-O' (1)

[hromium - 5J Ci (11_ (1) 1.60E-02 U)

Manaanese -54 Ci (1) (1) 1.64E-03 2.11E-03 Cobalt -57 Ci (1) (1) (1) II) i Cobalt -58 Ci 5.72E-04 (1) 3.08E-02 2.57E 02 Iron -59 Ci (1) (1) (1) (1)

Cobalt -60 Ci (1) (1) 1.26E-02 7.07E-03 Zinc -65 Ci (1) _

(1) (1) (1)

Strontium -89 Ci (1) (2) (1) (2)

Strontium -90 Ci (1) (2) 3.95E-04 (2)

Strontium -92 Ci (1) (1) (1) (1)

Niobium -95 Ci (1) (1) 2.66E-03 4.30E-04 Niobium -97 Ci (1) (1) 3.46E-03 6.51E-03 Zirconium -95 Ci (1) (1) 2.46E-03 (1) r Molybdenum -99 Ci (1) (1) (1) (1)

Technetium -99m Ci (1) (1) 4.87E-04 1.63E 04 i Ruthenium -106 Ci (1) (1) 1.22E-04 8.77E-05  ;

i Silver -Il0m _ Ci 3.73E-03 (!) 4.71E-02 1.48E-02 _ .

l Tin -113 Ci (1). (1) 1.17E-04 (1)  :

Antimony -!?2 Ci 5.51E-03 (1) 2.80E-04 2.71E-05 l r Antimony -125 Ci (1) (1) Im33E-01 5.14E-02 _  !

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l Technetium -132 Ci (1) (1) 2.85E-06 (1) i 16 l F

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  • IAB]1.lS - REG GUIDE 1.21 (Cont 1 ,

CALVERT CLIFFS NUCLEAR POWER PLANT EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 L10VID EFlLUENTS f

CONTINUOUS MODE BATCH MODE IST 2ND IST 2ND HUCLIDESRELEASED_ UNITS OUARTER OUARTER OVARTER OVARTER i

lodine -1 11 C1 (1) (1) 2.43E-02 4.22E-02 l

.Lqdine -133 Ci (1) (1) 8.44E-03 2.58E-03 Iodine -135 Ci (1) (1) (1) (1) .

Cesium -134 Ci 2.12E-03 (1) 4.20E-02 2.49E-02

-Cesium -136 Ci (1) (1) 6.55E-05 7.17E 04 .

Cesium -137 Ci 8.09E-03 (1) (1) (1)

Barium -140 Ci (1) (1) (1) 1.76E 03 ,

Lanthanum _-140 Ci (1) (1) 1.44E-04 5.66E-03  !

i' Cerium -141 Ci 'll 11) (1) (1)

Ceriem -144 Ci (1) (1) 5.71E-03 4.16E-05 j Tunasten -187 Ci (1) (1) 5.85E-04 (1) j Total For Period C1 2,13E.02 (1) 4.85E-01 2.60E-01 Xenon -131m Ci (1) (1) (1) (1)

Xenon -133 C1 4.60E-03 (1) 2.33E-01 1.71E+01 i Xenon -133m Ci (1) (1) 3.76E-03 2.01E-03 JLenon -135 Ci (1) (1) 2.24E-03 3.35E-04  !

Xenon -135m Ci (1) (1) (1) (1) ,

t Total For Period Ci 4.60E-03 (1) 2.39E-01 1.71E+01 (1) .Less than minimum detectable a-:tivity which meets the LLD requirements of f Technical Specification Surveillance Requirement 4.11.1.1.1.

(2) The second quarter strontium results will be submitted in a supplemental report as soon as ;

the concentration values are available, f I

k 17  !

IMLLM CALVERT CLIFFS NUCLEAR POWER PLANT l EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT ,

FIRST HALF - 1991

{

SQLLD_WlSIE AND 1RRADJAllp FUELS!))PMENTS l

[

A. SOLID WASTE SHIPPED OFLSJ1E_EQILBURIAL OR DLSPOSAL (NOT IRRADIATED FUEll ,

6 MONTH EST. TOTAL

1. Type of Waste UNITS PERIOD (flROR%  !

3

a. Dewatered spent resin m . 93E400 Ci 9 76E401 2.00E+01 .
b. Dry Compressible Waste (Burial) m3 8.24E+01 Contaminated Equipment, etc. Cj 2.35E401 5.00E401 i (Prior to Compacticn) m 2.01E+02
c. Irradiated Components, m3 -

[

Control Rods, etc. Ci -

5.00E+01 j

d. Other (CVCS filters) m3 .

j Ci -

2.00E+01 >

f a

2. Estimate of Major Nuclides (By Type of Waste)* [
a. Iron -

55 4.50E400 %  !

Cobalt -

58 1.41E401 % [

Cobalt -60 4.30E400 %

Nickel -63 2.16E401 %

Antimony -

125 1.27E+00 % i Cesium - 134 1.23E+01 %

Cesium -

137 3.82E+01 %

b. Carbon - 14 1.33E400 %

fron -

55 3.94E+01 %

Cobalt -58 2.08E+00 %

Cobalt -60 1.21E+01 %

Ruthenium - 106 1.42E400 %

Nickel -63 2.37E401 %

Cesium -

137 1.llE+01 %

Antimony -

125 1.74E400 % >

Only nuclides greater than 1% are reported r

18

-4 g,,arp, -w.. e n.m , , - . . , ,..e- - , , ne.-v, n --w..- " ' - - -

. . ;,.y TABLE 3A

},g.[g , ,

,( r '- ,

yf CALVERT CLIFFS NUCLEAR POWER PLANT

':s9[

f p.4:; ' Ji' w..' EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT FIRST HALF - 1991 E

s y ,00llD WASTE ANLIRRADI ATEP FUEL SHIPMENTS (Cont.)

j,

, f" '

S01.10 :lAS_IE 61 PIED OFFSlJE FOR BURIAL OR DISPOSAL (NOT 1RRADIATED FUEQ 5$3 - }s ,

c v,- , ...r:,
3. h k 001 *.aste Disposition

~

~

ll amber of Shipments Mode of Transportation Destination j, y 4 Motor Surtace Transit Chem Nuclear S3 stems Inc.

Barnv:1 i , 3.'

3 Motor Surface Transit Scientific < .o.1y Group Oak is.nte, TN

[

a e

19 I

APPENDIX A SOLID RADWASTE SHIPMENT DATA FOR SEMI-ANNUAL EFFLUENT RELEASE REPOP. TING FIRST HALF - 1991 TYPE WASTE: D:sW 10 CFR 61 WASTE CLASS: A SOURCE OF WASTE: Radiologically ",ontrolled Areas SHIPPING CONTAINER: CPC B-25 Box TOTAL CURIE QUANTITY: 2.36 Ci HOW DETERMINED: Dose to curie content, conversion by volume based on generic dist .bution and scaling factors TOTAL BURIED WASTE VOLUME: 2911 ft 3 HOW DETERMINED: Container volume and number of containers shipped SOLIDIFICATION AGENT OR ABSORBENT: None TYPE WASTE: bewatered Resin j 10 CFR 61 'lASTE CLASS: A S0bdCE OF WASTE: Miscellaneous Liquid Radwaste Processing /CVCS Processing CHIPFING CONTAINER: High Integrity L-8-1201 me- (120.3 ft 3)

TDTAL CURIE QUANTITY: 3.85 Ci HOW DETERMINED: Gamma scan using sample of resin

' TOTAL WASTE VOLUME: 120.3 ft 3buried / Resin volume 99.3 f t 3 HOW DETERMINED: Weighed liner and calculated volume of resin SOLIDIFICATION AGENT OR ACSORBENT: None l

A-1

APPENDIX A SOLIO RADWASTE SHIPMENT DATA FOR SEMI-ANNUAL EFFLUENT RELEASE REPORTING FIRST HALF - 1991 TYPE WAS1E: Dewatered Resin 10 CFR 61 WASTE CLASS: C SOURCE OF WASTE: Miscellaneous Liquid Radwaste Processing /CVCS Processing SHIPPING CONTAINER: High Integrity L-8-120 liner (120,3 f t3)

TOTAL CURIE QUANTITY: 93.8 Ci HOW DETERMINED: Gamma scan using sample from resin TOTAL WASTE VOLUME: 360.9 ft3 buried / Resin volume 251.6 ft 3 HOW DETERMINED: Weighed liner and calculated volume of resin SOLIDIFICATION AGENT OR ABSORBENT: None h

L s

I k

?

t I

A-2  ;

,;: .e ,,

4

+

APPENDIX B 4

SUMMARY

OF CHANGES TO THE CCNPP ODCM FIRST HALF 1991 1

a l

I n

\

APPENDIX B CHANGES TO THE CCNPP ODCM IN FIRST HALF 1991

SUMMARY

AND BASIS The Calvert Cliffs Offsite Dose Calculation Manual (ODCH) is contained in the Chemistry Procedure CP-607. Changes to this document are controlled through the r.ormal procedure changa review and approval process. This process meets the requirements of CCNPP Technical Specification 6.17 Two changes were made to the ODCM during the first half of the year 1991. The changes were reviewed by the Plant Operations and Safety Review Committee (POSRC) and approved by the Plant General Manager, Calvert Cliffs Nuclear Power Plant, prior to implementation. Change bars are provided in the right margin of the ODCH text to identify the thanges made.

The first change was reviewed and approved on March 11,1991, and desig mted as change 91-029. This change was made to incorporate a revised conti olling radionuclide pathway and location as identified in the Annual Land Use Survey.

The second change was reviewed on June 10, 1991 and approved on June 11, 1991 and designated as change 91-124. The purpose of this change was to provide a better description of how ef fluent Radiation Monitor System (RMS) alarm and trip setpoints are calculated. The methodology for how the calculation is performed was not changed. The radionuclide mix for the calculation of RMS trip setpoints was updated to better reflect more recent process streams.

rinally, a statement was added denoting the simplified dose assessment calculations to be used for the calculation of releases only if the Effluent Management System is out of service.

B-1

A'ITACilhll?NT 1 OI) Chi (CI'-607) Tl?XT W1 l'11 Cil ANGl?S

CP-607:

Page 1 of 27 q

  • i 0FFSITE DOSE CALCULATION MANUAL lPMF j 89-191!

.1.0-RELEASE E1NITOR SETPOINTS - lPCR  !

91-124j

.l.1 Liquid Effluents l

1.1.1 In accordance with the requirements of' Technical specification i 3.3.3.10, alarm setpoints shall be established for the liquid l effluent monitoring instrumentation to ensure that the ,

concentration of radioactive material released in. liquid l effluents does not exceed the limits of 10CFR20, Appendix B, l Table II, Column 2. l

').l.2 The setpoint shall be established taking credit for operation of lPMF l:

the applicable number of circulating water pumps. l89-191l t

1.1.3 The maximum allowable Setpoint shall be determined as follows: lPMF -!

SP - I (Aj x RF )j x D_ilution Water Flow Rate + Bkgd l90-102I .

L Discharge Flow Rate l j Where: .

SP - Setpoint (cpm). This is the trip setpoint that is not routinely, lPCR _

adjusted and-is designed to trip shut the discharge control l91-124:

valves during a discharge. l l Aj - Specific Activity (gCi/ml) of radionuclide i. For liquid waste l  ;

this mixture shall include' Co-58 (25%), .Cs-137 (25%),1-131' l (16%), Ag-110m (3%), Cs-134 (13%), Co-60 (4%), I-133 (3%), Nb-95 l ,

(2%), and Sb-125 (9%). This mixture was derived from actual l 1986,1987, and 1988 liquid release data. More recent activity l [

values from 1989 and 1990 were :et chosen due to shutdown l l conditions during most of this poriod of time. The actual pCi/ml l [

shall be determined using ue appropriate HPC data as defined in '

10CFR20, Appendix B, Table-II, Column 2. f Rev. 1 [

- I

CP-607 Paga 2 cf 27  :

RFj = Monitor response for lPMF radionuclide i (cpm /pCi/ml) l90-102 Dilution Water Flow Rate - Dilution from 3 circulating water pumps lPMF-(6E+5 gpm). l90-102' l

Discharge Flow Rate

  • Maximum liquid effluent flow rate 'into l ,

circulating water (gpm). l 1

Bkgd - RMS background countrate (cpm) l l

1.1.4 If fewer than three(3) circulating water pumps are running in the l conduit being discharged to, then a setpoint change shall be l '

initiated for monitored releases. l i

1.1.5 Administrative controls shall be utilized to insure that 10CFR20 lPMF-limits are not exceeded for the restrictive case where lodine 89-191 131 represents 100% of the activity.

1.2 Gaseous Effluents 1.2.1 In accordance with the requirements of Technical Specification lPCR  ;

3.3.3.9, alann setpoints shall be established for the gaseous l91-124' effluent monitoring instrumentation to ensure that the dose rate l l due to radioactive materials released in gaseous effluents does l not exceed 500 mrem /yr (total body) or 3000 mrem /yr (skin). The l setpoint chosen for alarm is the 10CFR50.778 limit of 2 MPC at l

( the site boundary. This ensures appropriats dC notification l

! and is a more conservative notification that the Technical l Specification 3.11.2.1 may be exceeded. l 1.2.2 The maximum allowable Setpoint shall be determined as follows: lPMF l l 90.-102 SP = I (Ag X RFj ) - Bkgd  !

l (X/Q) X (F)

Rev. I s

m CP-607 Page 3 cf 27 Where:

SP - Setpoint (cpm). This setpoint is not routinely adjusted lPCR and will initiate a high alarm in the control room for l91-124, RE-5415 or RE-5416. For 0-RE-2191, a trip signal is given l to shut the waste gas discharge control valves. l Aj - Specific Activity (pCi/ml) of noble gas radionuclide i for l the radionuclide mixture of Xe 133 (94%), Xe 135 (4%) and l Kr 85 (2%). (Refer to Section 1.2.3.) This mixture was l derived from actual 1986, 1987, and 1988 gaseous batch and l continuous release data. More recent activity values from l .

1989 and 1990 were not chosen due to shutdown conditions l during most of this period of time. l RF j e monitor response for radionuclide 1 (cpm /pCi/ml) lPMF I

X/Q = Average annual meteorological dispersion based on averaged 90 102 meteorological data (2.2E-06 sec/m3 ) -

F = System flow rate for release point (m3/sec Skgd - RMS background level (cpm) lPMF <

90 102

  • 1.2.3 The total body dose rate conversion factors (Attachment 1) shall be used to calculate the nuclide concentrations since they are 1 i more restrictive than the skin dose factors for the radionuclide mixture of interest.

l.2.4 The main vent setpoints shall be set to ensure that l l simultaneous releases from both U-l and U-2 Main Venti do not -

exceed esta'lished o dose limits at the site boundary, i

1.3 The radiation monitor Alarm Setpoint for effluent releases, lPCR (gaseous and liquid), shall be predicted using pre-release data for l91-124 l each discharge as follows: l cpm - [(Aj X RF j )(Dilution Factor, if appropriate) + Bkgd] X 1.5 lPMF l90-102 Where: l Ag - Specific Activity ( Ci/ml) l '

RF j - monitor response for each isotope based on calibrated monitor lPCR response tables / graphs supplied by vendors or generated 91-124 in plant.

Rev. 1 4

+- .

  • . 1 ,

CP-607 .

Pcg3 4 cf 27 I

Bkgd - RMS background countrate lPMF 1.5 - Detector tolerance (150% of anticipated count rate) 90 102 L

An expected effluent monitor Alarm Setpoint is calculated for each lPCR discharge. The expected monitor release alarm is entered into the l91-124:

plant computer. If during a celease this expected response is l exceeded and causes an alarm, the discharge is manually terminated. l l 1.4 The liquid and gassous waste radiation monitor descriptions ,

including effluent controls and sample points are listed on j Attachment 2. .

1.5 The radiation monitors have remote visual / audible alarms and visual meter indications at the local detector locations. In addition, remote detector alarms, meters and recorders are located in the Control Room.

2.0 JBDJ0 ACTIVE LIQUID EFFLUENTS 2.1 frocessino Eouipment and flow Paths 2.1.1 The normal liquid waste processing equipment flow paths and monitoring systems are defined on Attachment 3. ,

2.1.2 Non routine lineups for the liquid waste processing system include tank recirculations, bypass options, extended filtration and ion exchange. These may be utilized as required by system chemical and operational activities.

2.2 l m jd Effluent. Concentration 2.2.1 Routine batch liquid waste releases to the Chesapeake Bay include:

a. Miscellaneous Waste
b. Reactor Coolant Waste
c. Steam Generator Blowdown (Unit I and 2).

Rev. 1

e- -

- . . 1 1.,

CP-607 Pag) 8 cf 27 In lieu of the individual radionuclide dose assessment as defined in Section 2.3.2, the following simplified dose calculational equatit.n may be used for verifying compliance with the dose limits of Technical Specification 3.11.1.2. Refer to Section 2.3.5 for the derivation and justification for this simplified method. This simplified method for calculating lPCR releases is used if the effluent management computer system is l91-124-out of service. l

. Total Body ,

Dtb - 2.22 E + O2 x V x I Ag l CW i

Maximum Oroan D

max = 1.70E + 03 x Y x I Ag '

CW Where:

I Dtb = conservatively evaluated total body dose (mrem) 2.22E+02 - conversion factor (hr/ min) and the conservative ' total body dose conversion factor (Cs-134 total body -

1.33+04 mrem /hr per pCi/ml) l V - volume of liquid effluent released (gal)  :

CW - average circulating water discharge rate during release  ;

period (gal / min) l A3 - average specific activity (pC1/ml) of radionuclide i in undiluted liquid effluent representative of the volume (V)

L L D max = Conservatively evaluated maximum organ dose (mrem) r 1.70E+03 = conversion factor (hr/ min) and the conservative .

maximum organ dose conversion factor (1-131), thyroid

-1.02E+05 mrem /yr per pCi/ml) h 1

Rev. 1 i

a, f.,

CP-607 Paga 16 of 27 DC = 3.17E-08 x X/Q (2 Mg x Qg)

DB = 3.17E-08 x X/Q (I Ng x Qj)

Where:

DS - air dose due to gama emissions for noble gas radionuclides (mrad) 3.17E coa /ersion factor (yr/sec)

X/Q = atmospheric dispersion to the controlling site boundary (P. 2E-06 sec/m3 )

Mg - air dose factor due to gamma emissions from noble gas radionuclide i (mrad /yr per pCi/m3) from Attachment (1)

Qg - cumulative release of noble gas radionuclide i over the period of interest (gCi)

D - air dose due to beta emissions from noble gas radionuclides (mrad)

Ng = air dose factor due to beta emissions from noble gas radionuclide i (mrad /yr per pCi/m3) Attachment (1) 3.4.2 Gaseous Dose - Noble Gases (Simplified Equation)

In lieu of the individual nodie gas radionuclide dose assessment as presented above, the following simplified dose calculation may be used for verifying compliance with the dose limits of Technical Specification 3.11.2.2. (Refer to Section 3.4.5 for the derivation and justification for this simplified method.) This simplified method for calculating l PCP.

releases is used if the effluent management computer system l91-124 is out of service. l Rev. I

& r . *,

CP-607 Paga 18 cf 21 W - atmospheric dispersion paiameter (X/Q or D/Q) to the controlling location. lPMr l89230 X/Q - atmospheric ditpersion for radiochemical particulates l via the inhalation pathway and for tritium via all l pathways (seconds /m3 ). l l

b/Q- atmospheric deposition for radiciodine and particulatec l (excluding tritium) via ingestion pathway. (m-2), l l

3 Rg - dose factor for radionuclide 1, (mrem /yr per yCi/m ) or l (m2 - mrem /yr per pCi/sec) from Attachment 7 for each age group a and the applicable pathway p. Values for R4 were {

derived in accordance with the methods described in NUREG-0133.

Og - cumulative release over the period of interest fer i radionuclide 1 131 or radioactive material in particulate form with half-life greater than 8 days (yCi).

The centrolling pathway and location for the CCNPP has lPCR been determined to be cow milk pathway at 4800 meters. l91029 3

SSW. At this location the X/Q - 1.9E-7 seconds /m lPMF and the D/Q - 8.63E-10m-2 The small amounts of tritium l90-102 discharged from the plant main vents are not considered lPHr in offsite particulate dose assessments. This is l89-230 justified because modeling evaluations show that the l maximum expected doses would be several orders of l magnitude below .he regulatory limits. l

. l l

l l

Rev. I

S t '. *,i CP-607  :

Paga -19 of 27_  :

I-3.4.4 Gaseous Dose - Radiciodine and Particulates _(Simplified Equation) l In lieu of-the individual radionuclide (I-131 and particulates)

{

dose assessment as presented above, the following simplified- l

- dose calculational equation may be used for verifying compliance l with the dose limits of T.S. 3.11.2.3.  !

I Dmax - 3.17E-08 x W x Ry.333 x I Qj i

W'-

re:  !

I

.D max = maximum organ dose (mrem) j 2

Rg 333 = 1.05E+12 m -mrem /yrper#Ci/sec,infantI-131 lPCR [

doseparameterfor_thethyroidforthecontrollingl91-029 {

cow milk pathway. -l t l

l~

W - D/Q.= 8.63E-10 m-2, l The ground plane exposure and inhalation pathways need not be considered when the above simplified calculation is used because of the.overall negligible contribution of these pathways to the . _ _ .

total thyroid dose. It is recognized that for some particulate L radionuclides (e.g., 00-60 and Cs-137), the ground exposure-pathway may represent a higher dose contribution than either the l

) -vegetation or milk pathway. However, use of the I-131 thyroid ,

dose parameter for all radionuclides will maximize the organ dose calculation,- especially considering that.no other radionuclide has'a higher dose parameter for any organ via-any pathway than I-131 for the thyroid via the applicable vegetation or milk pathway.

9 Rev. I'

_ i

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