ML17056A674

From kanterella
Jump to navigation Jump to search
Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989.
ML17056A674
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 12/31/1989
From:
NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17056A673 List:
References
NUDOCS 9003150337
Download: ML17056A674 (64)
v • d • e


Text

NINE MILE POINT NUCLEAR STATION UNIT 2 I

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT July through December 1989 DOCKET NO: 50-410 LICENSE NO.: NPF-69 NIAGARA MOHAHK POHER CORPORATION 9003150337 900302 PDR ADOCK 05000410 r vc

l 0

NINE NILE POINT NUCLEAR STATION UNIT 2 SEHI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Duly through December 1989 Facility: Nine Hile Point Unit ¹2 Licensee: Niagara Hohawk Power Corporation

1. Technical Specification Limits:

A) Fission and activation gases:

The dose rate limit of noble gases from the site to areas at and beyond the site boundary shall be less than or equal to 500 mrem/year to the whole body and less than or equal to 3000 mrem/year to the skin.

2. The air dose from noble gases released in gaseous effluents from the Nine Hile Point 2 Station to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and, during any calendar year to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

BEC) Tritium, Iodines and Particulates, half lives > 8 days:

The dose rate limit of Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days, released to the environs as part of gaseous effluents from the site to areas at or beyond the site boundary, shall be less than or equal to 1500 mrem/year to any organ.

2. The dose to a member of the public from Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half lives greater than 8 days as part of gaseous effluents released from the Nine Hile Point 2 Station to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 7.5 mrem to any organ and, during any calendar year to less than or equal to 15 mrem to any organ.

D) Liquid Effluents The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gas, the concentration shall be limited to 2E-04 microcuries/ml total activity.

D. Liquid Effluents (Cont'd)

2. The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from Nine Mile Point Unit 2 to unrestricted areas shall be limited during any calendar quarter to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ, and during any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.
2. Maximum Permissible Concentrations A) Fission and activation gases:

None specified B&C) Iodines and particulates, half lives > 8 days:

None specified D) Liquid Effluents:

Strontium-89 3E-6 Strontium-90 3E-7 Cesium-134 9E-6 Cesium-137 2E-5 Iodine-131 3E-7 Cobalt-58 9E-5 Cobalt-60 3E-5 Iron-59 5E-5 Zinc-65 1E-4 Manganese-54 lE-4 Chromium-51 2E-3 Zirconium-Niobium-95 6E-5'E-5 Molybdenum-99 Technetium-99m 3E-3 Barium-140 2E-5 Cerium-141 9E-5 Tungsten-187 6E-5 Arsenic-76 2E-5 Iodine-133 lE-6 Hydrogen-3 3E-3 Iron-55 8E-4 Manganese-56 1E-4 Sodium-24 3E-5 Xe-133 2E-4 Xe-135 2E-4

3. Average Energy (Fission and Activation gases Mev)

Third Quarter Fourth Quarter  :

4. Measurements and Approximations of Total Radioactivity Described below are the methods used to measure or approximate the total radioactivity and radionuclide composition in effluents.

A) Fission and Activation Gases: Noble gas effluent activity is determined by on-line gamma spectroscopic monitoring (intrinsic germanium crystal) of an isokinetic sample stream. During on-line monitoring system inoperability grab sampling and isotopic analyses are performed as specified by Technical Specifications.

B) Iodines: Iodine effluent activity is determined by gamma spectroscopic analysis (at least weekly) of 'charcoal cartridges manually or automatical,ly sampled from an isokinetic sample stream.

C) Particulates: Activity released is determined by gamma spectroscopic analysis (at least weekly) of particulate filters manually or automatically sampled from an isokinetic sample stream.

D) Tritium: Tritium effluent activity is measured by liquid scintillation or gas proportional counting of monthly samples taken with an air sparging/water trap apparatus.

E) Liquid Effluents: Isotopic Analysis of a representative sample of each batch.

F) Solid Effluents: Isotopic contents of waste shipments are determined by gamma spectroscopy analyses of a representative sample of each batch. Scaling factors established from primary composite sample analyses conducted off-site are applied, where appropriate, to find estimated concentration of non-gamma emitters. For low activity trash shipments, curie content is estimated by dose rate measurement and application of appropriate scaling factors.

5. Batch Releases The following information relates to batch releases of radioactive materials in liquid and gaseous effluents.

A) Liquid 1.

2.

Number of batch releases:

Total time period for batch releases: ~ ~

~ hours 44 min.

3.

4.

5.

6.

Maximum time period for a batch release:

Average time period for a batch release:

Minimum time period for a batch release:

Average stream flow during period of

~

~ ~

hours hours hours

~ min.

min.

min.

release of effluent into a flowing stream: Not Applicable 7.

8.

Total volume of water used to dilute the liquid effluent during release periods:

Total volume of water available to dilute

~~ liters the liquid effluent during reporting period: ~2K++9 i ters 1

-3

l 7 t

B) Gaseous (Primary Containment Purge)

1. Number of batch releases:
2. Total time period for batch releases: 9L5 hours
3. Maximum time period for a batch release: hours 4.

5.

Average time period for a batch release:

Minimum time period for a batch release: ~

12.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> hours 6, Abnormal Releases A. Liquids None.

B. Gaseous none

I P t

TABLE 1 A SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES ELEVATED AND GROUND LEVEL July through December 1989 3rd 4th EST.TOTAL QU69XEB MU~T XBBQH M A. Fi n

l. Total release Ci 2,96E+01 4.06E+01 5.0E+01
2. Average release rate for period pCi/sec 3.81E+00 5.22E+00
3. Percent of Technical Specification Limit B. XZU~
1. Total iodine-131,133,135 Ci 2. 07E-03 2. 23E-03 5. OE+01
2. Average release rate for period pCI/sec 2.60E-03 2.89E-05
3. Percent of Technical Specification Limit /.

1 Particulates with half-lives > 8 days¹ Ci 4.04E-03 3.07E-04 5.0E+01

2. Average release rate for period pCi/sec 5.08E-04 3.86E-05
3. Percent of Technical Specification Limit Gross alpha radio-activity Ci 1.89E-05 9.55E-06 5,0E+01 D. ~r
1. Total release Ci 5.83E+00 1.70E+00 5.0E+01
2. Average release rate for period pCi/sec 7.33E-01 2,14E-01
3. Percent of Technical Specification Limit
  • See Item E attached.

¹Include. Mo-99 also Revised alpha release data for 1988 and 1989 reports is as follows:

1st Quarter 1988 = 3.58E-06 Ci 2nd Quarter 1988 6.04E-06 Ci 3l d Quarter 1988 1.04E-05 Ci 4th Quarter 1988 = 1..08E-05 Ci 1st Quarter 1989 = 6.29E-06 Ci 2nd Quarter 1989 9.27E-06 Ci

)

k

TABLE 1A (Continued)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES ELEVATED AND GROUND LEVEL July through December 1989 3l d 4th QUKIKR ~ANTE i in n A iv n

1. Percent of (}uarterly Gamma Air Dose Limit ZJ2E~
2. Percent of (}uarterly 3.

Beta Air Dose Percent of Annual Limit Gamma 2 "N~

Air Dose Limit to Date Percent of Annual Beta j 'HKM1 F~l 4.

Air Dose Limit to Date

5. Percent of Nhole Body Dose Rate Limit LJ2~E-
6. Percent of Skin Dose Rate Limit i I n P wi h h lf-liv
1. Percent of Quarterly Dose Limit
2. Percent of Annual Dose Limit to Date kJ5FHQ
3. Percent of Organ Dose Rate Limit

TABLE 18 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 GASEOUS EFFLUENTS-ELEVATED (STACK)

July through December 1989 CONTINUOUS MODE BATCH MODE Nuclides Released Argon-41 Ci 7.05E-01 8.09E+00 Krypton-85m Ci 1.15E-01 Krypton-87 Ci 1.08E+00 Krypton-88 Ci 8.27E+00 5.75E+00 Xenon-133 Ci Xenon-135 Ci 4.44E-02 4,75E-01 Xenon-135m Ci 1.55E+00 1.94E+00 Xenon-137 Ci 1.10E+01 1.33E+01 Xenon-138 Ci 8.01E+00 9.83E+00 2.44E-01

2. I~au.

Iodine-131 Ci ~ 1.03E-04 3.10E-04 O.OOE+00 Iodine-133 Ci 1.97E-03 1.92E-03 0.00E+00 Iodine-135 Ci ** ** 0.00E+00 3.

Strontium-89 Ci 9.74E-05 1. 87E-04 Strontium-90 Ci 4.15E-05 Cesium-134 Ci Cesium-137 Ci 4.90E-07 Cobalt-60 Ci 3.80E-07 6.63E-06 Cobalt-58 Ci 4.30E-07 Manganese-54 Ci 4.21E-06 Barium-Lanthanum-140 Ci 1.02E-05 7.28E-06 Antimony-125 Ci Niobium-95 Ci Cerium-141 Ci Cerium-144 Ci Iron-59 Ci Cesium-136 Ci Chromium-51 Ci 2.84E-04 1.03E-04 Linc-65 Ci 4.68E-06 9.31E-06 Iron-55 Ci 1,92E-05 <1,87E-05 Molybdenum-99 Ci 4.90E-05 2.85E-06 4, Tri ~o1 Ci 4.75E+00 1.09E+00 1.58E-02 2.25E-02

    • Less than sensitivity of 1.00E-04 pCi/ml for noble gases, 1.00E-12 pCl/ml for iodines, 1.00E-11 pCi/ml for particulates and 1.00E-06 pCi/ml for tritium as applicable in Technical Specifications.

TABLE 1C SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 GASEOUS EFFLUENTS-COMBINED GROUND LEVEL-ELEVATED (REACTOR BUILDING VENT)

July through December 1989 CONTINUOUS MODE BATCH MODE Nuclides Released Argon-41 Ci 0.00E+00 0.00E+00 Krypton-85 Ci 0.00E+00 0.00E+00 Krypton-85m Ci O.OOE+00 0.00E+00 Krypton-87 Ci 0.00E+00 O.OOE+00 Krypton-88 Ci 0.00E+00 0.00E+00 Xenon-133 Ci 0.00E+00 O.OOE+00 Xenon-135 Ci 0.00E+00 O.OOE+00 Xenon-135m Ci 0.00E+00 O.OOE+00 Xenon-137 Ci O.OOE+00 O.OOE+00 Xenon-138 Ci 0.00E+00 O.OOE+00

2. ~Iline~

Iodine-131 Ci O,OOE+00 O.OOE+00 Iodine-133 Ci 0.00E+00 O.OOE+00 Iodine-135 Ci 0.00E+00 O.OOE+00 3.

Strontium-89 Ci <2. 36E-06 0.00E+00 O.OOE+00 Strontium-90 Ci <4.15E-07 0.00E+00 O.OOE+00 Cesium-134 Ci 0.00E+00 O.OOE+00 Cesium-137 Ci 9.40E-07 0.00E+00 O.OOE+00 Cobalt-60 3.28E-06 O.OOE+00 O.OOE+00 Cobalt-58 Ci 2.42E-06 0.00E+00 O.OOE+00 Manganese-54 Ci 1.62E-06 0.00E+00 O.OOE+00 Barium-Lanthan um-140 Ci O.OOE+00 O.OOE+00 Antimony-125 Ci O.OOE+00 O.OOE+00 Niobium-95 Ci 0.00E+00 0.00E+00 Cerium-141 Ci 0.00E+00 O.OOE+00 Cerium-144 Ci 0.00E+00 O.OOE+00 Iron-59 Ci O.OOE+00 O.OOE+00 Cesium-136 Ci O.OOE+00 O.OOE+00 Chromium-51 Ci 3.47E-03 1.51E-04 0.00E+00 O.OOE+00 Zinc-65 Ci 7.35E-05 9.45E-06 0.00E+00 O.OOE+00 Iron-55 Ci ** 0.00E+00 O.OOE+00 Molybdenum-99 Ci 2.60E-05 7.23E-06 0.00E+00 O.OOE+00 j'rishi~ Ci 1.07E+00 5.87E-01 0.00E+00 0.00E+00 There were no batch releases from this release point.

Less than the sensitivity of 1.00E-04 pCi/ml for noble gases, 1.00E-12 pCi/ml for iodines, 1.00E-11 pCi/ml for particulates and 1.00E-06 pCl/ml for tritium as applicable in Technical Specifications.

TABLE 2A SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES July through December 1989 3rd 4th Est. Total Qm~ Qm~ Erron A. Fi iv i r

1. Total release (not including tritium, gases, alpha) 3. OOE-03 1. 92E-01 5. 00E+1
2. Average diluted con-centration during reporting period pCi/ml 2.05E-10 1.57E-08
3. 'ercent of applicable 1 imi t 3.19E-04 2.05E-02 B. ~Tr 1~1
1. Total release Ci 7.30E-01 5.86E+00 5.00E+1
2. Average diluted con-centration during reporting period pCi/ml 4.97E-08 4.77E-07
3. Percent of applicable limit '/ 1.66E-03 1.59E-02 C. D v
l. Total release Ci 9.08E-05 1.27E-04 5.00E+1
2. Average diluted con-centration during reporting period pCi/ml 6.18E-12 1.04E-11
3. Percent of applicable limit 3.09E-06 5.19E-06 D.
1. Total release Ci E. %clem.
l. Prior to dilution liters 8.50E+05 6.81E+06 5.00E+1
2. Volume of dilution water used during release period liters 2.08E+08 1.39E+09 5.00E+1 3 ~ Volume of dilution water used during reporting period liters 1,47E+10 1.23E+10 5,00E+1 Unit Shutdown none detected.
    • <1.00E-07 pCi/ml sensitivity for gross alpha radioactivity.

-9

l TABLE 2A (Continued)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES July through December 1989 3l d 4th Quadr Gum~

F. P r n fT hni Percent of quarterly Whole Body Dose Limit 1. 5.69E-02 4.74E+00

2. Percent of quarterly Organ Dose Limit (Liver) 3. 78E-02, 3.15E+00
3. Percent of Annual Whole Body Dose Limit to Date 2.56E-01 2.63E+00 4, Percent of Annual Organ Dose Limit to Date (Liver) '/ 1.71E-01 1.75E+00
5. Percent of 10CFR20 Concentration Limit 3.19E-04 2.05E-02 Percent of Dissolved or Entrained Noble Gas Limit 1. 3.09E-06 5.19E-06

-10

TABLE 2B RADIOACTIVE EFFLUENT RELEASE SEMI-ANNUAL REPORT NINE MILE POINT NUCLEAR STATION ¹2 LIQUID EFFLUENTS RELEASED July through December 1989 anni 3 CONTINUOUS MODE J4'4K BATCH MODE Nuclides Released Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Iodine-131 Ci Cobalt-58 Ci 7.59E-05 1.79E-02 Cobalt-60 Ci 7.15E-04 3.42E-02 Iron-59 Ci 6.42E-03 Zinc-65 CI 9.70E-04 6.48E-02 Manganese-54 Ci 4.41E-04 2.01E-02 Chromium-51 Ci 8.04E-04 4.89E-02 Zirconium-Niobium-5 Ci Molybdenum-99 Ci 2.25E-04 Technetium-99m Ci 1.96E-04 Barium-lanthanum-140Ci Cerium-144 Ci 1.87E-04 Hydrogen-3 Ci 7.30E-01 5.86E+00 Sodium-24 Ci 2.18E-04 Iron-55 Ci Manganese-56 Ci Nickel-65 Ci Arsenic-76 Ci Iodine-133 Ci Tungsten-187 Ci 1.78E-04 Xenon-133 Ci 5.50E-05

  • Xenon-135 Ci 3.58E-05 1.27E-04
  • Less than sensitivity of 5.00E-07 pCi/ml for gamma emitting nuclides, 1.00E-05 pC1/ml for dissolved and entrained noble gases and tritium, 5.00E-08 pCi/ml for Sr89 and Sr90 and 1,00E-06 pCi/ml for Fe55 as applicable in Technical Specifications.

-11

l~

TABLE 3A SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS July through December 1989 A ~ 1 W ff-i fr 8 1 rDi N irr EST. TOTAL iJKQf~

a.* Spent resins, filter sludge bottoms, etc. m'. 26E+02 3.80E+02 5.00E+01 5.00E+01

b. Dry compressible waste, contaminated equip., etc. 3.06E+01 5.00E+01 1.23E-01 5.00E+01 Irradiated components, control rods, etc. m'one None
d. Other ** 8,19E-01¹ m'i 1,12E-01¹
  • All were solidified in cement as Class A waste in strong, tight containers. All was shipped as radioactive LSA.
    • Noncompacted commingled trash shipped to Oakridge, TN for processing before burial in Barnwell, S.C. (buried during second half of 1989).

¹Preliminary data supplied by processor.

-12

~

~

TABLE 3A (Continued)

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹2 (1989)

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS

2. r Nuc 1 i
a. Spent Resins, filter sludges, evaporator bottoms, etc Zinc-65

~Pr ~

1.29E+01 Est. Total

~rr r 5.00E+01 Iron-55 5.29E+00 5.00E+01 Cobalt-60 4.83E+00 5,00E+01 Manganese-54 2.44E+00 5.00E+01 Cobalt-58 2.26E+00 5.00E+01 Chromium-51 7.14E+01 5.00E+01 Others 8.80E-01 5.00E+01 b, Dry Compressible Waste, Contaminated Equips, Etc.

Est. Total

~rr r Zinc-65 5.00E+01 Iron-55 Cobalt-60 II'.96E+01 Manganese-54 9.87E+00 1.89E+01 1.31E+01 5.00E+01 5.00E+01 5.00E+01 Cobalt-58 2.30E+00 5.00E+01 Chromium-51 3.46E+01 5.00E+01 Others 1.63E+00 5.00E+01 c ~ Irradiated components, control rods, etc.

NONE

d. Other (Noncompacted Est. Total commingled trash shipped to ~pmi r Oakridge, TN for processing prior to burial)

Cobalt-60 3.08E+01 5.00E+01 Manganese-54 2.91E+01 5.00E+01 Zinc-65 2.37E+01 5.00E+01 Cobalt-58 1.33E+Ol 5.00E+01 Iron-55 2.42E+00 5.00E+01 Others <1.00E+00 5.00E+01 W Di iin hi mn f Tr D in in 16 Truck Burial in Barnwell SC B. IRRADIATED'UEL SHIPMENTS (DISPOSITION)

N r f hi en D i i n None

-13

~

~

\

(~

TABLE 4 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE MILE POINT NUCLEAR STATION ¹ 2 RECENT CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL There have been no changes to the Offsite Dose Calculation Manual during this reporting period.

-14

l)

TABLE 5 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT NINE HILE POINT NUCLEAR STATION ¹2 RECENT CHANGES TO PROCESS CONTROL PROGRAH Revision 3 of the PCP was issued in September 1989, reformatting this document to comply with the general rewrite of Administrative Procedures. The basic program was unchanged, except for the reformat. A copy of Rev. 3 and approval documentation is attached to this report.

r TABLE 6 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 EXPLANATION OF INSTRUMENTATION INOPERABILITY DATE OUT DATE RETURNED Offgas Hp-Train August 1987 08/28/89 Design Problems.

Offgas H2-Train August 1987 08/28/89 Design Problems.

Offgas Hp-Common August 1987 08/28/89 Design Problems.

Reactor/Radwaste January 10, 02/23/90 Difficulty in Vent Noble Gas 1989 trouble-shooting and obtaining parts.

Reactor/Radwaste January 10, 02/23/90 Difficulty in Vent System Flow 1989 trouble-shooting and obtaining parts.

Reactor/Radwaste January 10, 02/23/90 Difficulty in Vent Effluent Flow 1989 trouble-shooting and obtaining parts.

Reactor/Radwaste January 10, 02/23/90 Difficulty in Vent Iodine Sampler 1989 trouble-shooting and obtaining parts.

Reactor Radwaste January 10, 02/23/90 Difficulty in Vent Particulate Sampler 1989 trouble-shooting and obtaining parts.

Liquid Radwaste November 1987 Not Returned Procedure changes are Effluent Radiation to Service needed to restore operability.

Tank Level Indicating None on site Devices Cooling Tower October 8, 12/30/89 Lost power to monitor Blowdown Radiation 1989 during voltage spike.

Evaluating repair via special task force.

Service Hater A June 1989 Not Returned to Replacement pump is not Radiation Service available. Expected repair is August 1990.

-16

TABLE 7 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 HOURS AT EACH HIND SPEED AND DIRECTION January through December 1989 Consistent with Amendment 94 of the Technical Specifications, a summary of hourly meteorological data collected for the year 1989 is not provided with this report; however, it is available upon request.

M

-17

II ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY JANUARY DECEMBER Doses to members of the public (as defined by the Technical Specifications) from the operation of the NMP2 facility as a result of activity 1nside the site boundary is controlled by activities at the Energy Center. Th1s facility 1s open to the public and offers educational 1nformation, summer picnicking activities and fishing, Any possible doses received by a member of the public by utilizing the private road that transverses the east and west s1te boundaries are not considered here since travel the distance.

it takes a matter of minutes to The activity at the Energy Center that is used for the dose analysis is fishing because it is the most time consuming. Although there is no specific survey informat1on available, many of the same indiv1duals have been observed to return again and again because of the access to salmonid and lake trout populations. Dose pathways cons1dered for this act1v1ty 1nclude direct radiation, inhalation and ground dose (shoreline sed1ment or soil). Other pathways, such as ingestion pathways, are not considered because they are either not applicable or are ins1gnificant. Only releases from the NHP2 stack and vent were evaluated for the inhalation pathway.

The direct rad1ation pathway is evaluated in accordance with the methodology found in the Offsite Dose Calculation Manual (ODCH). This pathway considers radiation from the generating facilities, any possible overhead plume, and direct radiation from plume submersion. The direct radiation pathway is evaluated by the use of high sensit1v1ty environmental TLDs. Since any signif1cant fishing activity near the Energy Center occurs between April through December, environmental TLD data for the approximate period of April l December 31, 1989 was considered. Data from two environmental TLDs from the approximate area where the fishing occurs were compared to three control environmental TLD locations for the same time period. The average f1shing area TLD dose rate was 7.72E-03 mRem per hour for the period. The average control TLD dose rate was 6.76E-03 mRem per hour for the period (approximate second, third and fourth calendar quarters of the year). The average 1ncrease in dose as a result of fishing in this area at a conservative frequency at eight hours per day is 3.00E-01 mRem from d1rect radiat1on for the per1od in question. The ma]ority of the dose from th1s pathway is from the NHPl facility because of its proximity to the fishing area. A small portion may be due to the NHP2 facility.

The inhalation dose pathway is evaluated by utilizing the inhalation equation 1n the Offs1te Dose Calculation Manual, as adapted from Regulatory Guide 1.109. The equation basically gives a total inhalation dose in mRem for,. the t1me period in question (April December). The total dose equals the sum, for all applicable rad1onuclides, of the NMP2 stack and vent release concentration, times the average NMP2 stack and vent flowrate, times the applicable five year average calculated X/Q, times the inhalation dose factors from Regulatory Guide 1:109, Table E-7 times the Regulatory Guide 1.109 annual air intake, t1mes the fractional port1on of the year in question. In order to be slightly conservative, no rad1ological decay is assumed.

-18

ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY (Continued)

The 1989 calculation utilized the following information:

NMP2 stack:

Unit 2 average stack flowrate 47.33 m3/sec X/Q value 9.6E-07 (annual historical average)

Inhalation dose factor - Table E-7 of Regulatory Guide 1.109 Annual air intake 8000 m3 per year (adult)

Fractional portion of the year - 0.0356 (312,hours)

I-131 4.05 E-Ol pCi/m3 I-133 3.93 E-00 pCi/m3 Mn-54 4.35 E-03 pCi/m3

- 3.22 E-01 pCi/m3 'r-89 Sr-90 1.18 E-03 pCi/m3 Co-60 7.07 E-03 pCi/m3 Cs-137 4.36.E-04 pCi/m3 Co 3.82 E-04 pCi/m3 Na-24 1.43 E-00 pCi/m3 Cr-51 3.61 E-Ol pCi/m3 Zn 1.38 E-02 pCi/m3 Tc-99m 3.67E-02 pCi/m3 Mo-99 4.64 E-02 pCi/m3 Ba-140 1.55 E-02 pCi/m3 La-140 7.87 E-02 pCi/m3 Fe-55 4.58 E-02 pCi/m3 H-3 6.32 E+03 pCi/m3 NMP2 Vent:

Unit 2 average vent flowrate - 110.8 m3/sec X/Q value. 2.8E-06 (annual historical average)

Inhalation dose factor - Table E-7 of Regulatory Guide 1.109 Annual air intake 8000 m3 per year (adult)

Fractional portion of the year 0.0356 (312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br />)

Co 9. 19 E-04 pCi/m3 Cr 1.37 E-00 pCi/m3 Co 1.25 E-03 pCi/m3 Mn-54 6.15 E-04 pCi/m3 Zn 3.15 E-02 pCi/m3 Mo-99 1.26 E-02 pCi/m3 Tc-99m 1.12 E-02 pCi/m3 Na 4.05 E-01 pCi/m3 Cs-137 3.57 E-04 pCi/m3 Sr-89 2.49 E-03 pCi/m3 Sr-90 4.57 E-04 pCi/m3 Fe-55 3.88E-02 pCi/m3 H 1.22 E+03 pCi/m3 The inhalation dose to a member of the public as a result of activities inside the site boundary is S.'l4E-05 mRem to the thyroid (maximum organ dose) and 3.01E-05 mRem to the whole body.

-19

II ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY (Continued)

The dose from standing on the shoreline to fish is based on the methodology in the Offsite Dose Calculation Manual as adapted from Regulatory Guide l.l09.

Dur1ng l989, 1t was noted that fishing was performed from the shoreline on many occasions although waders were also utilized. In order to be conservative, it is assumed that the maximum exposed individual fished from the shoreline at all times. The use of waders, of course, would result in a dose of zero from this pathway. The shoreline sediment doses are not taken 1nto consideration by env1ronmental TLD data.

The Offsite Dose Calculation Manual equat1on basically gives the total dose to the whole body and skin from the sum of all plant related radionuclides detected in shorel1ne sed1ment samples. The plant related radionuclide concentration is ad)usted for background sample results, as appl1cable. The equation, therefore, yields the whole body and sk1n dose by multiply1ng the radionuclide concentration ad)usted for any background data (as applicable),

times a usage factor, times the sediment or soil density 1n grams per square meter (to a depth of one centimeter) times the applicable shore w1dth factor, times the regulatory guide dose factor, times the fractional portion of the year over wh1ch the dose is appl1cable. In order to be conservative and to simplify the equat1on, no radiological decay is assumed since the applicable rad1onuclides are usually long lived.

The calculation utilized the following information:

- Usage factor - 312 hours.

- Density in grams per meter 40,000.

- Shore width factor 0.3.

Hhole body and skin dose factor for each radionuclide Regulatory Guide 1.109, Table E-6.

- Fractional portion of the year 1, (used average radionucl'ide concentration over total time period).

Average Cs-137 concentration 0.49 pC1/g.

Average Co-60 concentration 0.06 pC1/g.

The total whole body and skin dose from standing on the shoreline to fish is 1.15 E-02 mRem whole body and 1.35 E-02 mRem skin dose for the period.

Doses to members of the public relative to activit1es inside the site boundary from aquatic pathways other than ground dose from shoreline sed1ment/so1l are not applicable.

-20

ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY (Continued)

In summary, the total dose to a member of the public as a result of activ1t1es ins1de the site boundary from direct radiation, inhalat1on and shoreline dose pathways is 3.00 E-01 mRem to the whole body and 5.14 E-05 mRem to the maximum exposed internal organ (thyroid). The dose to the skin of an adult 1s 1.35 E-02 mRem. These doses are generally a result of the operation of NMP2. A portion of these doses for the direct radiation pathway are attr1butable to the NMPl facility.

-21

ATTACHMENT 3 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 RADIATION DOSES TO THE LIKELY MOST EXPOSED, MEMBER OF THE PUBLIC OUTSIDE THE SITE BOUNDARY JANUARY - DECEMBER ill Radiation doses to the likely most exposed member of the public are evaluated relative to 40CFR190 requirements. The dose limits of 40CFR190 are 25 mRem (whole body or organ) per calendar year and 75 mRem (thyroid) per calendar year. The intent of 40CFR190 also requires that the effluents of NMP2 as well as other nearby uranium fuel cycle facilities be considered. In this case, the effluents of NMPl, NMP2 and the James A. FitzPatrick (JAF) facilities must be considered.

Doses to the likely most exposed member of the public as a result of effluents from the site can be evaluated by using calculated dose modeling based on the accepted methodologies of the facilities'ffsite Dose Calculation Manuals or may, in some cases, be calculated from the analysis results of actual environmental samples. Acceptable methods for calculating doses from environmental samples are also found in the facilities'ffsite Dose Calculation Manuals. These methods are based on Regulatory Guide 1.109 methodology.

Dose calculations from actual environmental samples are, at times, difficult to perform for some pathways. Some pathway doses should be estimated using calculational dose modeling. These pathways include noble gas air dose, inhalation dose, etc. Other pathway doses may be calculated directly from environmental sample concentrations using Regulatory Guide 1.109 methodology.

Since the effluents from the generating facilities are low, the resultant gaseous and liquid effluent doses are anticipated to be low. In view of this, doses can be based on calculated data. Doses are not based on actual environmental data for 1989 with the exception of doses from direct radiation, fish consumption and shoreline sediment. In addition, in order to be conservative and for the sake of simplicity, it is assumed in the dose calculations that the likely most exposed member of the public is positioned in the maximum receptor location for each pathway at the same time. This approach is utilized because the doses are very low and the computations are greatly simplified.

The following pathways are considered:

1. The inhalation dose is calculated at the critical residence because of'he high occupancy factor. In order to be conservative, the maximum whole body and organ dose assumes no correction for residing inside a residence.
2. The milk ingestion dose is calculated utilizing the maximum milk cow location. As noted previously, in order to be conservative and for the sake of simplicity, the likely most exposed member of the public is assumed to be at all critical receptors at one time. In this case, the member of the public at the critical residence is assumed to consume milk from the critical mi'lk location.

-22

l ATTACHMENT 3 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 RADIATION DOSES TO THE LIKELY MOST EXPOSED MEMBER OF THE PUBLIC OUTSIDE THE SITE BOUNDARY (Continued)

The maximum dose from the m1lk 1ngest1on pathway as a result of consuming goat's milk is based on the same criteria established for item 2 above (ingestion of cow's milk).

The maximum dose associated from consuming meat is based on the critical meat animal. The likely most exposed member at the critical residence is assumed to consume meat from the critical meat animal location.

The maximum s1te dose associated with the consumption of vegetables is calculated from the crit1cal vegetable garden location. As noted previously, the likely most exposed member of the public is assumed to be located at the critical residence and is assumed to consume vegetables from the critical garden location.

The dose as a result of direct gamma radiation from the s1te encompasses doses from direct "shine" from the generat1ng facilities, d1rect radiation from. any over head gaseous plumes, plume submersion and from ground deposition. This total dose is measured, by env1ronmental TLD. The crit1cal location 1s based on the closest year round residence from the generating facilities as well as the closest residence in the cr1t1cal downwind sector in order to evaluate both direct radiat1on from the generating fac111ties and gaseous plumes as determined by the local meteorology. During 1989, the closest res1dence and the cr1tical downwind residence are at the same location.

The measured average dose for l989 at the critical residence was 62.6 mRem. The average control dose (average of three locations) was 57.5 mRem. The average dose at the cr1tical res1dence is greater than the average control locat1on dose. A ma)or port1on of this net dose is due to the differences between doses from naturally occurr1ng rad1onucl1des in the soil and rock at the different locations. This difference in dose rate can be demonstrated by observing the 1989 average dose for an environmental TLD located near the cr1tical residence TLD but approximately 700 feet closer to the generat1ng facilit1es. The annual average dose for this TLD location was 59.7 mRem. The dose for this location is lower than the critical residence location even though they are close to one another and even though the TLD location with the lowest dose is closer to the generating facil1ties.

The dose, as a result of f1sh consumption, is considered as part of the aguatic pathway. The dose for 1989 is calculated from actual results of the analysis of environmental fish samples. For the sake of being conservative, the average plant related radionuclide concentrations were utilized from fish samples taken near the site discharge points. The average concentrat1on was ad)usted to account for any background concentrations using average control sample data. Only Cs-137 was detected during 1989 (net concentration was 6.30E-03 pC1/g wet). The calculated maximum adult organ dose was 1.44E-2 mRem to the liver. The maximum whole body dose is 9.40E-03 mRem to an adult.

-23

I ATTACHMENT 3 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1989)

NINE MILE POINT NUCLEAR STATION ¹2 RADIATION DOSES TO THE LIKELY MOST EXPOSED MEMBER OF THE PUBLIC OUTSIDE THE SITE BOUNDARY (Continued)

8. The only other pathway considered is the shoreline sediment pathway relative to recreational activities. The dose due to recreational activities from shoreline sediment is based on the methodology in the Offsite Dose Calculation Manual as adapted from Regulatory Guide 1.109.

The Offsite Dose Calculation Manual gives the total dose to the whole body and skin from the sum of plant related radionuclides detected in shoreline sediment samples. The plant related radionuclide concentration is ad]usted for background sample results, as applicable. The total whole body and skin dose from shoreline recreational activities is 1.01 E-03 mRem whole body and 1.17 E-03 mRem skin dose for the period.

In summary, the maximum dose to the most likely exposed member of the public is 4.86E-02 mRem to the child liver (maximum organ dose) and 3.61 E-02 mRem to the whole body (child). The maximum organ and whole body doses were a result of gaseous effluents. Doses as a result of liquid effluents were secondary.

The total whole body, maximum organ and skin dose from shoreline recreational activities and fish consumption are 1.04 E-02 mRem whole body, 1.44 E-02 mRem to the liver and 1.17 E-03 mRem skin dose for the period. The direct radiation dose to the critical residence from the generating facilities was insignificant or zero. These maximum total doses are a result of operations at the Nine Mile Point Unit 1, Nine Mile Point Unit 2 and the James A.

Fitzpatrick facilities. The maximum organ dose and whole body dose are below the 40CFR190 criteria of 25 mRem per calendar year to the maximum exposed organ or the whole body, and below 75 mRem per calendar year to the thyroid.

-24

'+1

$ 01

\I 1

~

TECHHICBL HEVEEW HllQ CQBTHOL.

5WlZKhlRF MCIMSHTlllW" 2 . N 'P d R . NC Cl 1 I dc.~.'s, pen% I Pro r'uthor

Sage DOSCf'Xp C10% OC ChRSIS(Lteaiee the aetace/cwsoa oE ILeneral chaagee) iC ~ e BC'. rc 'r'" ""< Al re A.

a cS c ws MODIFICATION

" IF PERICOIC REVIEW WITHNO CHANGES (Prd Rsv, NC). USE THE LAST PUBLISHED REVISION NEER ANO CONTINUE REVIEW PROCESS..

IRIRhl)ISCIPI.IMLRYlLEVIEV (minimum ol one oerion reoaired)

DEPT. NAME TITLE SIGNATURE DATE CROSS DISCIPLIELRY REVKT OC aot required. uee liaee for )uetification etxtemeat)

DEPT. NAME TITLE SIGNATURE DATE QP veau s IP Nor IN coNCURRENCE. Do Nor SIGNILUT RETLNNoocurxNT ro THE AUTHOR WITH ccnrmrs Routecf to Quality Aeeureace d'or reriew: Yea Q, No Q. IC No.

I Q. A. Representative ate ~ Eccommeatcareattmched.Q. Ai Roaual co A~JUL. for retied Yes /No G. If No. rasa A~Jt&. Repreeenaxttve ~ ~ are II " Sc comment! are attached. Q.

>>>>>>>>>>vu>>>>>>>>>>>>>>>> Roate to AUTHOR I UNIT SUPV.

SAFIZV ANAJ VSIS aZavmZn: NO g. VZS a(SZZATrACHZD)

REVtrr OF THE Saa~ DLXmmrrmS IS RECOMMENDED. (Approrera It's CODEX mme AID hPPROVAI.

chILL ciyufy approval'on tho procedure corer shoot l .. 0 DocmlKIT azm FOR soRC Oazrnrs ~ ..........). ~HLovED. ms a. Ro a.

OWRKRSEIP DEPT S zpr NC F161AA SHIT i OF@

AP-2.0 -32 April 1989 P

a*

If I

~ II

k TKCHSIChL RISKIER hRD CONTROL EVALUATIONOF NEED FOR SAFEIY ANALYSIS IN ACCORDANCE VITH i0 CFR 50.59 (Documents that require General Supt. approvai per'ech Spec. 6.8)

FOR DOHIMENTNO. ~ ~EV. DATE The Author (A) and four SORC Members (Minimum - 2 regular members, 2 alternates) are to respond to each of the questions below.

'I NO YES~

Does the document/r evision r esuit in a change to the facility A g CI or procedures described in the FSAR ? 1 8 CI 2

3 ~Ef Cl Ci Does the document/r evision deviate from compliance to Tech Specs. or is the mar gia of safety defined in the basis A

1 gCf Cl Cl reduced  ? 2 K Cl 3 69- 0 Cl Does the document/revision increase the probability of occurreace, ot the consequences of an accident, or malfunction of equipment impor tant to safety (Class i) evaluated in the

'SAR incr eased? 3 B- O 4 gf Cl Does the document/revision create the possibility for an accident or malfunction of a differ eat type than aay evaluated in the FSAR?

3 G- Cl

  • A "MAYBE constitutes a YES response.

4 g Cl SORC MEMBERS RECOMMENDATIONS TO GENERAL SUPERINTENDENT Recommended Nuclear Engineer ing or Tech Services perfor'm a safety ANALYSIS to pr esent to SORC (noted by a "YES" response to any of the above questions)

Recommended full SORC committee review this Evaluation l 2 3 4 of need for Safety Analysis. Cl a Cl O Recommended approvai - This document does not involve an unreviewed safety questioa. m SORC Member Name SO C Member Signatures Date 1

2 SORC meeting 3 number (if Required) 4 ~

Figaro 2.DA SH 2 OF 4 AP-2.0 ~3 APril 1989

TECHllICHLREVIEW MQ CDllTHDL.

REFERENCE DOCUMENTS The items entered below have been included in the preparation and/or reviev of the attached reference document and are presented in place of a specific check sheet for the document.;.

The following persons were Procedure is in compliance MENTARY vith consulted about this procedure the folio<<ing Technical Specifications MAME Tl ACTION~MEND 8 a i/ co~

@ 'g, /, 3'~/ Ab)~uzi M~M Proc,~sr'~ A/

Compliance with: CFR / US-NRC QI/g P~g~

Compliance <<ith REGULATORY GUIDES(s) DATED BY ANSI STANDARD(s) DATED BY

~<8 Ad

~~/

<<Ith: ASME Boiler and is consistent with the following Station

'ompliance Pressure Vessel Code(s) or Site procedures:

SE ION DATE DD DU R 8 OTHER INFORMATION SOURCES CONSULTED BY AOTEOR Q~WAZ..............DATE................/.... pz REVIEWED BY................DATE...F 4 .Eg.'.......

COMMENTS FIGURE 2.0A SHEET 3 OF 4 AP-2.0 -34 April 1989

~

Cr 1

~g

TECHIIICALHEVIEW Allo COllTRGL'..

REVIEW CHECK LIST TO BE PREPARED BY AUTHOR CHECK LIST FOR DOCUMENT NO../ I  !.2!../......... REV....D ......... DATE, .f.........../

A" ONLY BOZES THhT hPPLY YES NA All references needed to implement the procedure are clearly identified and available........... 5 0 The procedure contains adequate equiyment lists, precautions and limitations, prerequisites, graphs, diagrams or data sheets as required.................~..............

Surveillance and Maintenance Procedure utilizes PLANT IMPACT statement associated with approval/permission for use.............,....................,.....,......................,...,.... AH OHO ~0~ 0~ 0~ ~0 0

, As appropriate, procedure addresses use of MARK - UPs...................................,...,.................. 0 If appropriate, procedure requires use of fire protection measures.

e, burning permits etcoo ~ ~ ~ ~ 0 ~ 00 ' '0 0000000000000000 '00000000000 ~ 00000000000 ~ 0 F 0000 If leads are lifted. jumpers placed or blocks used In the procedure, the PLANT IMPACT statement acknowledges such use....................................................... ~ 000 \ 0000000000 0000 '

~ ~ ~ 0000 F 0 ' lg As appropriate, procedure notifies other affected departments such as Q.C., Operations, IAC, Maintenance, Rad Protection etc 0 If Technical Specification ls exceeded. appropriate action is Identified......->> -. -- - 8 0 The procedure references valve numbers, motor control numbers. your suyplies.

Instrumentation identification is clear and correct...........,............,...................,...,....... I!I When encountered. EQ. related equipment is identified as such.......................... 0 steps are clear and accurate. They are not unnecmsarily difficult to Implement.... 9 0 Q'rocedure The procedure reflects the latest system or component configuration--------.

The procedure reflects work as it is to be done at the station.................................................... 9 0 Procedure removes any Jumpers or blocks and restores lifted leads used to effect the VOrk00000000000000000000000000000000000000 0000000000000000000000000000000 000000000000000000000000000000000000000000000000000000000000000-000 0 6 RETURN TO SERVICE" uses double verification and'identifies specifics being verified.......H. 0 For maintenance procedures. RETURN TO SERVICE either performs a POST MAINTENANCE TEST or references a required tes t............................................................................................. 0 Xl MARK RK UPs UP are cleared or surrendered...................................,.............................................. 0 Zi ACCEPTANCE CRITERIA" identifies accomplishment of specific goals................. ..... b 0 FORM PREPARED B ..............

FIGURE 2.0.4 SHEET 4 OF 4 AP-2.0 -3 APri1 1,9Q9

I l.o PURPOSE 1.1 To describe the methods for processing p a cka g ing an d transportation oof v ow>>level low>> radioactive waste and provide assuran assurance of complete solidification, of various radioactive "wet e w t es " in accordance was with applicable NRC regulations and guidelines.

1.2 To satisfy the Nuclear Regulatory Commission's Low-Level Waste and Uranium Recovery Pro)ects Branch (FLU) requir equ rement andd establish parameters within which the Werner & Pfleid e erer. Corporation Volume'rocess o ume Reduction System (WPC-VRS) must be operated to ensure complete solidification.

Pf'c,c.c+5 ( r a(~Q g grog~cia.cog pe field Ps 4w iWc'Zoic onionhLce vAta FLU requirements provides assurance that the requirements identified in 10CFR61, Sub Part D, Technical Requirements for Land Disposal Facilities and Fi 1 assification and Waste Form Technical Position P ayers, whi ch lass

~u<l that a Cl lass A waste shall be a free standing monolith with

%&free standing .erat~~ WraaKt.

Iieui g 2.0 SCOPE 2.1 A licabilit This procedure applies to the Unit 2 Radwasteas e P rocess Control Pro r rogram for solidification and transportation of radioactive waste.

2.2.1 To satisfy the Nuclear, Regulatory Commission's Low-Level Waste and Uranium Recovery Prospects Branch (FLU) requirements.

2.2.2 To establish process parameters within which the Merner and Pfleiderer Corporation Volume Reduction System (WPC-VRS) shall operate to ensure complete solidification of waste.

2.2.3 To ensure compliance with Federal regulations for the packaging and transportation of radioactive materials.

2.2.4 To ensure Radwaste Operators are trained and qualified in the operation of radioactive waste processing equipment.

2.2.5 To ensure Chemistry Technicians are trained and qualified in the sampling and analysis of wet radioactive waste.

2.2.6 To ensure Radiation Protection Technicians are trained and qualified in radiological controls monitoring of radioactive waste shipments.

AP-3.7.1 Rev. 02

I 4>>

Nuclear Regulatory Commission $ 61.56 column. Example: A waste contains Sr- {6) Waste must not. be pyrophoric.

90 in a concentration of 50 Ci/m'nd Pyrophoric materials contained in Cs-137 in a concentration oi 22 waste shall be t,reated, prepared. and the concentrations bot.h exceed packaged to be nonfLammable.

Ci/m'ince the values in Column 1, Table 2, they {7) V aste in a gaseous form must, be must be compared to Column 2 values. packaged at a pressure that does not For Sr-90 fraction 50/150=0.33: 'or Cs- exceed 1.5 atmospheres at, 20 C. Total 137 fraction, 22/44=0.5; the sum of activity must not exceed 100 curies per the fractions=0.83. Since the sum is container.

less than 1.0, the waste is C!ass B. '8) Waste containing hazardous, bio-(8) Determination of coucentrntious logical. pathogenic. or infectious mate-in zoastes. The concentration of a radi- rial must be treated to reduce to the onuclide may be determined by irdt- maximum extent practicable the po-rect. methods such as use of scaling tential hazard from the non-radiologi-factors which relate the inferred con- cal materials.

centration of one radionuclide to an- (b) The requirements in this section other that is measured, or radionu- are intended to provide stability of the clide material accountability, if there waste. Stability is intended to ensure is reasonable assurance that the indi- that the waste does not structurally rect methods can be correlated with degrade and affect overall stability of actual measurements. The concentra- the site through slumping. collapse, or tion of a radionuclide may be averaged other failure of the disposal unit and over the volume of the waste, or thereby lead to water!nfiltration. Sta-weight of the waste ii the units are ex- bility is also a factor in limiting expo-pressed as nanocuries per gram. sure to mi inadvertent intruder, since characteristics.

it provides a recognizable and nondis-0 61.66 %aside persible waste.

(a) The f'ollowing requirements are (1) Waste must'have structural sta-minimum requirements for all classes bility. A structurally stable waste form of waste and are intendea to facilitate will generally maintain its physical di-handling at the disposai site and p.o- mensions and its form, under the ex-vide protection of health and safety of oected disposal conditions such as personnel at the disposal site. weight; of overburden and compaction (1) Waste must not be packaged for equipment,. the presence of moisture.

disposaL in cardboard or fiberboard and microbial activity, and internal boxes. factors such as radiation effects and (2) Liquid waste must be solidified or chemical change&. Structural stability packaged in sufficient absorbent mate- can be provided by the waste form rial to absorb twice the volume of the itself, processing the waste to a stable liquid. form, or placing the waste in a dispos.

(3) Solid waste containing liquid al container or structure that provides shall contain as little free standing stability after disposal.

and noncorrosive liquid as is reason- (2) Notwithstanding th provisions ably achievable. but in no case shall in 5 61.56(a) (2) and (3). liquid wastes, the liquid exceed 1% of the volume. or. wastes containing liquid, must be (4) Waste must, not be readily capa- converted into a form that contains as ble of detonation or of explosive de- little free standing and noncorrosive composition or reaction at normal liquid as is reasonably achievable, but pressures and temperatures, or of ex- in no case shall the liquid exceed 1%

plosive reaction with water. of the volume of the waste when the

{5) Waste must not contain, or be ca- waste is in a disposal container de-pable of generating, quantities of toxic signed to ensure stability, or 0.5'// of gases, vapors, or fumes harmful to per- the volume of the waste for wite sons transporting, handling, or dispos- processed to a stable form.

ing of the waste. This does not apply (3) Void spaces within the waste.and to radioactive gaseous waste packaged between the" waste and its package ln accordance with paragraph (a)(7) of 'must. be reduced to:the extent, practi-this section. cable.

109

C A

Technical Position on Waste. Fo~.

Rev 0 May 1983 demonstrate that the product is a free standing monolith with no more than 0.5 percent of the waste volume as free liquid.

An alternative to processing some Class 8 and C waste streams, particularly ion exchange resins and filter sludges, is the use of a high integrity container. The high integrity container would be used to provide the long-term stability required to meet the stability requirements in 10 CFR Part 61. The design of the high integrity container should be based on its specific intended use in aorder to ensure that the waste contents, as wel,l as interim storage and ultimate disposal environments, will not compromise its integrity over the long-term. As with waste solidification, a process control program for dewatering wet solids should be developed and utilized to ensure that the free liquid requirements in 10 CFR Part 61 are being met.

C. Reoulatorv Position

1. Solidified Class A Waste Products
a. Solidified Class A waste products which are segregated from-Class 8 and C wastes should be free standing monoliths and have no more than 0.5 percent of the Waste volume as free liquids as, using the method described in ANS 55.1. 'easured
b. Solidified Class A waste products which are nut segregated from Class 8 and C wastes should meet the stability guidance for Class 8 and C wastes provided below.
2. Stability Guidance for Processed (i.e., Solidified) Class 8 and C Wastes
a. The stability guidance in this technical posi tion for processed wastes should be implemented through the qualification of the individual licensee's process control program. Generic test data may be used for qualifying process control programs.

Through the use of a well designed and implemented process control program, frequent requalification to demonstrate stabi'lity is expected to be unnecessary. However, process control programs should include provisions to periodically demonstrate that the solidification system is functioning properly and waste products continue to. meet the 10 CFR Part 61 stability requirements. Waste specimens should be prepared 117

1

~ V

'J

Retrieved from "https://kanterella.com/w/index.php?title=ML17056A674&oldid=2440278"