W3P88-1268, Semiannual Radioactive Effluent Release Rept for Jan-June 1988
| ML20153D034 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/30/1988 |
| From: | Burski R LOUISIANA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| W3P88-1268, NUDOCS 8809020019 | |
| Download: ML20153D034 (37) | |
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Semiannual Radioactive Effluent Release Report January 1, 1988 - June 30, 1988 Waterford 3 SES Louisiana Power & Light 1
r W310526HP 8809020019 880630 PDR ADOCK 05000382 /
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, i TABLE OF CONTENTS 1.0 SCOPE 2.0 SUPPLEMENTAL INFORMATION 2.1 Regulatory Limits 2.2 Maximum Permissible Concentrations 2.3 Average Energy 2.4 Measurements and Approximations of Total Radioactivity 2.5 Batch Releases 2.6 Abnormal Releases 3.0 GASEOUS EFFLUENTS 4.0 LIQUID EFFLUENTS 5.0 SOLID WASTES 6.0 METEOROLOGICAL CATA 7.0 ASSESSMENT OF DOSES 8.0 RELATED INFORMATION 8.1 Changes to the Process Control Program 8.2 Changes to the Offsite Dose Calculation Manual 8.3 Unavailability of REMP Milk Sampling 8.4 Report of Technical Specification Required Instrument Inoperability 8.5 Hissed Effluent Samples 8.6 Corrections to Previous Semiannual Radioactive Effluent Release Reports 9.0 TABLES 10.0 ATTACHMENTS U310526HP 1
l 1.0 SCOPE This Semiannual Radioactive Effluent Release Report is submitted as required by Louisiana Power and Light's Waterford 3 Technical Specification !
6.9.1.8. It covers the period from January 1, 1988 through June 30, 1988.
Information in this report is presented in the format outlined in Appendix l B of Regulatory Guide 1.21. l The information contained in this report includes:
i (1) A summary of the quantities of radioactive liquid and gaseous effluents j and solid wastes released from the plant during the reporting period; (2) Explanation of why certain instrumentation was not ree ored to !
operable status within the time specified in the ACTION Statement, as !
per Waterford 3 SES Technical Specification 3.3.3.10 and 3.3.3.11; (3) A summsry of missed samples required by Waterford 3 SES Technical Specification 4.11.2.1.2; and t
(4) A sumary and correction of errors identified in previous Semiannual Radioactive Release Reports. l The summary of meteorological data and results from the assessment of radioactive doses due to the release of liquid and gaseous radioactive effluents will be included in the Semiannual Radioacti'te Effluent Release Report to be submitted within 60 days after January 1,1989.
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2.0 SUPPLEMENTAL INFORMATION 2.1 Regulatory Limits _
The Technical Specification Limits applicable to the release of radioactive material in liquid and gaseous effluents are described in the following sections.
2.1.1 Fission and Activation Gases (Noble Gases)
The dose rate due to radioactive noble gases released in gaseous effluents from the site to areas at and beyond the site boundary sha*1 be limited co 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 air dose due to noble gases released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and, l
- b. During any calendar year: Less than or equal to 10 mrad ;
I for gamma radiation and less than or equal to 20 mrad for
! beta radiation.
2.1.2 Iodines; Particulates, Half Lives > 8 Days; and Tritium The dose rate due to Iodine-131 and 133, tritium, and all radionuclides in particulate form with half lives greater than eight (8) days, released in gaseous effluents from the site to areas at and beyond the site boundary, shall be limited to less than or equal to 1500 mrce/yr to .$ny organ.
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The dose to a member of the public from Iodine 131 and 133, tritium, and all radionuclides in particulate form with half lives greater than eight (8) days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,
- b. During any calendar year: Less than or equal to 15 mrem to any organ.
2.1.3 Liquid Effluents The concentration of radioactive ma'.erial 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. Fe.r dissolved or entrained noble gases, the concentration shall be limited to 2.0E-4 pCi/ml total activity.
The dose or dose comitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas shall be limited to the following:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and
- b. 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.
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2.1.4 Uranium Fuel Cycle Sources The dose or dcie commitment to any member of the public due to releases of radioactivity and radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except t.he thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.
2.2 Maximum Permissible Concentrations 2.2.1 Fission and Activation Gases; Iodines; and Particulates,llalf Lives > 8 Days For gaseous effluents, maximum permissible concentrations are not directly used in release rate calculations since the '
applicable limits are expressed in terms of dose rate at the site boundary.
2.2.2 Liquid Effluents The maximum permissible concentration (MPC) valu(s specified in 10 CFR Part 20, Appendix B, Table II, Column 2 are used as the permissible concentrations of liquid radioactive eff'uents at the unrestricted area boundary. A value of 2.0E-4 pCi/ml is useU as the MPC for dissolved and entrained noble gases in liquid effluents.
2.3 Average Energy This is not applicable to Waterford 3 SES's radiological effluent !
technical specifications.
2.4 Measurements and Approximations of_ Total Radioactivity The quantification of vadioactivity in liquid and gaseous effluents was accomplished t,y , n; forming the sampling and radiological analysis of effluents in accordance with the requirements of Tables 4.11-1 and f 4.11-2 of the Waterford 3 SES Plant Technical Specifications.
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2.4.1 Finsion and Activation Gases (Noble Gases)
For continuous releases, a gas grab sample was analyzed monthly fcr noble gases. Each week a Gas Ratio (GR) vas calculated according to the following equation:
GR = Average Weekly Noble Gas Monitor Reading Monitor Reading During Noble Gas Sampling The monthly sample analysis and weekly Gas Ratio were then used to determine noble gases discharged continuously for the I
previous week. For gas decay tank and containment purge batch releases, a gas grab sample was analyzed prior to release to determine neble gas concentrations in the batch.
In all cases the total radioactivity in gaseous effluents was determined from measured concentrations of each radionuclide present and the total volume discharged.
2.4.2 Iodines and Particulates .
Iodines and particulates discharged were sampled using a continuous sampler which contained a charcoal cartridge and a particulate filter. Zach week the charcoal cartridge and particulate filter were analyzed for gamma em!tters using gamma spectroscopy. The determined radicauclide concentrations and effluent volume discharged were used to calculate the previous week's activity releasna.
The particulate samples vere composited and analyzed quarterly for Sr-89 and Sr-90 by a contract laboratory (Teledyne Isotopes). Particu'. ate gross alpha activity was measured weekly using alpha scintillation counting techniques. The determined activities were used to estimate effluent concentrations in subsequent releases until the next scheduled analysis was performed, f
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e Greb samples of continuous and batch releases were analyzed monthly for tritium. The determined concentrations were used to estimate tritium activity in subsequent releases until the next scheduled analysis was performed.
2.4.3 Liquid Effluents '
t For continuous releases, samples were collected weekly and analyzed using gamma spectroscopy. The measured concentra-tions were uced to determine radionuclide concentrations in i the previous week's releases. For batch releases, gamma analysis was performed on the sample prior to release.
For both continuous and batch releases, composite samples ,
were analyzed quarterly by a contract laboratory (Teledyne '
Isotopes) for Sr-89, Sr-90, and Fe-55. Samples were composited and analyzed monthly for tritium and gross alpha using liquid scintillation and gas flow proportional counting techniques, respectively. For radionucliden measured in the composite ;
samples, the measured concentrat ions in the composite samples from the previous month or quart.er were used to estimate ;
released quantities of these is stopes in liquid effluents during the current month or quarter.
The total radioactivity in liquid effluent releases was determined from the measured and est'. mated concentrations of (
each radionuclide prosent and the total volume of tha effluent l discharged, 2.5 Batch Releases I l
The summarization of information for gaseous and liquid batch releases !
is included in Table 1.
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- i 2.6 Abnorms1 R31 eases 2.6.1 Abnormal Release on April 3, 1988.
On April 3, 1988, an unplanned, unmonitored and uncontrolled
. release of radioactivity occurred during removal of the outside door of the containment equipment hatch. At no time were any Technical Specifications dose limits exceeded.
Description of Event:
On the evening of April 2, 1988 the steps necessary to open the equipment hatch commenced. At approximately 0025 on April 3, 1988, leak rate testing on the equipment hatch was completed and removal of the outer door began, Just before 0200, the Personnel Conte.mination Monitors (PCM-l's) and friskers located outside of the equipment hatch on the Q-Deck began alarming. A noble gas sample obtained at 0200 indicated the presence of Xe-133 in the area just outside of the equipment hatch at a concentration of 1.6E-06 uCi/ce. A subsequent sample pulled at 0311 showed that the Xe-133 concentration had increased to 2.7E-06 uCi/ce. The source of the activity was investigated. It was determined that the activity was ori-ginatinp from the cuntainment annulus. The activity in the l annulus came from earlier operation of Containment Atmosphere Removal System (CARS). Therefore, shield building ventilation was resumed at 0340. Q-Deck samples collected at 0357 showed that Xe-133 actinity had increased to a maximum of 5.5E-06 uCi/cc. At 0455 a noble gas sample collected just outside the equipment hatch indicated no detectable levels of activity.
Opening of the equipment hatch was completed at approximately 0900.
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Cause of Event:
The root cause of the release was related to shield building ventilation not being run while the outer hatch wcs being removed. After the seals on the outer hatch were deflated, activity present in the containnent annulus was allowed to escape.
Corrective Actions:
In order to prevent a recurrence of this type of release, procedures are being modified to require shield building ventilation to be run continuously while the seal on the outer door is deflated. If required, shield building ventilation would only be secured long enough to move th9 door to the open position. Shielding building ventilation would not be required as long as containment purge is operating. Having shield building ventilation operating will help prevent any release of activity from the containment annulus or from leakage past the inner door seal.
Radiological Consequences of the Release:
A total amount of 3800 uCi of Xe-133 were estimated to have been released during this event. The gamma and beta doses in air from this release were calculated to be SE-07 and IE-06 mrad, respectively. These doses are SE-06 and 7E-06 percent of the respective annual ganma and beta dose limits (approximately IE-05 percent of the gamma and beta quarterly limits) allowed by Technical Specifications. Therefore, the doses resulting from this release were deemed to be insignificant.
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Since a release rate from the Q-Deck area or Containment annulus could not be reliably calculated due to low flow rates, the instantaneous dose rates could not be calculated directly. However, calculations indicate that to exceed the instantaneous dose rate limits, an exit velocity of 280 miles per hour would have to be attained. At no time was this exit velocity possible. Therefore, the instantaneous release rate limits could not have beca exceeded.
2.6.2 Abnormal Release on May 23, 1988 On May 23, 1988, a small amount of radioactivity (Co-58 and I-131) was released through an abnormal release pathway.
This monitored and controlled release occurred during Integrated Leak Rate Testing (ILRT) depressurization.
Initial sampling of the containment atmosphere prior to the release indicated that the effluent did not contain radioactive concentrations above the appropriate lower limits of detection (LLD's) (i.e., no activity was detected).
However, continuous sa ples collected during the release and later analyded indicated the presence of activity. The post-release measured radioactive concentrations of I-131 and Co 58 in the effluent were determined to be at and below the pre-release lower limits of detection (i.e., activity was present at levels lower than could be reliably detected in the pre-release samples). The reason this activity was detected in the release samples and not in the pre-release samples is due to the fact that a much larger sample volume was collected during the release. With this increased sample volume, much lower detection levels were attained and activity was detected. The samples taken prior to release satisfied the appropriate Technical Specification LLD's and although the release occurred through an abnormal pathway, the pathway was monitored and continuous samples were collected. Based on the results of these continuous samples, at no time were any Technical Specification dose limits exceeded.
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Description of Event:
During Integrated Leak Rate Testing of containment, the containment is pressurized to a maximum pressure of 44 psig.
After obtaining the required measurements, containm-nt must be depressurized by discharging the excess air added during the pressurization phase. Current wording in Table 4.11-2 of the Waterford 3 Technical Specifications does not specifically ;
address sampling requirements associated with ILRT depressuri- ;
zation. While ILRT depressurization is not technically a containment purge (by Definition 1.23 in the Technical Speci-fications), it was decided that pathway restrictions and sampling requirements associated with purging containment of airborne radioactivity were applicable. Therefore, if '
the containment atmosphere contained radioactivity, it would t be released via the plant stack. ;
1 In order to depressurize containment through the plant stack the normal purge pathway could not be used due to the damage that would occur to the duct work from the calculated exit ;
velocities and pressures. Therefore, it would be necessary l to depressurize into the Reactor Auxiliary Building. Because l of RAB Ventilation System operating limitations, depressuri- ,
zation through the RAB would require thirty-six to forty-eight !
hours to complete. The possibility of depressurizing directly f to atmosphere to reduce this depressurizatien time was evaluated. (
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After reviewing the current Waterford 3 Technical Specificatiom i regarding gaseous radioactive effluents with the Licensing Department, it was concluded that if the containment atmosphere l s mpled prior to release was radioactive, the release could not be made directly to the atmosphere. It would have to be released through the plant stack via the RAB Normal Ventilation i System. However, if the atmosphere sampled did not contain any radioactivity, the pathway restrictions in the Technical Specifications did not apply as long as adequate precautions ;
were taken to identify changing conditions during the release (i.e., radioactivity in the release pathway) that would I warrant termination of the release.
A safety evaluation was performed on the release pathway to examine the radiological consequences of an accidental release of radioactivity during ILRT depressurization. As a result of the safety evaluation, a portable radiation monitor (with l alarming capability) would be used to monitor the release f pathway for changing conditions that would warrant (i.e.,
greater than two times background) termination of the release. t The evaluation concluded that plant safety would not be decreased by utilizing this release pathway. :
i l Prior to 'LRT depressurization, instructions were issued ,
describira the sampling and monitoring requirements for ILRT l depressurization. The instructions specifically 6tated that
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) reactor produced radioactivity was detected in the pre-release !
grab samples. In addition, the instructions specified that !
the samples collected were to be treated as effluent release
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samples; that is, the same detection limits and sampling [
criteria used for routine radioactive effluent samples were !
applicable. The instructions also specified that the release be immediately terminated in the event that increasing l radiation levels (i.e., greater than two times background) !
{ vere detected by the portable radiation monitor. The radiation monitor selected was also capable of collecting (
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On May 23, two separate sets of gas, iodine, and particulate samples were collected. These samples were analyzed as effluent release samples in accordance with Health Physics L Department procedures. In addition, chemistry obtained and analyzed a sample for tritium in accordance with Chemistry Department procedures. All analysis results were below the ,
lower limits of detection. The air to be released was there-fore treated as a non-radioactive effluent and discharged directly to the atmosphere. Depressurization to the atmosphere began on May 23 at 09:00 and lasted until 21:39 the same day.
After completing depressurization, the continuous iodine and particulate samples collected by the temporary radiation j monitor from the depressurization pathway were analyzed. Low i r
levels of I-131 and Co-58 were found in the samples with a ,
calculated average concentration in the discharge stream of 1.5 E-11 uCi/cc I-131 and 3.2 E-13 uti/cc Co-58. The a posteriori lower limits of detection on the pre-release r samples ranged between 1.1 E-11 to 1.9 E-11 uCi/cc for I-131 and 3.2 E-12 to 6.8 E-12 uCi/cc for Co-58. Therefore, the activity detected after the release was on the order of or below the detection limits of the pre-release samples.
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O Cause of the Release:
The root cause of radioactivity being release through an l abnormal pathway was related to the fact that the concentrations of radioactivity in the release stream were at or below the lower limits of detection of the pre-release samples. Tech- !
i nical Specification Table 4.11-2 states that the a priori ;
lower limits of detection for the principle particulate gamma emitters should be 1 E-11 uCi/cc. The a priori limit for weekly I-131 samples should be 1 E-12 uCi/cc with footnote g allowing this limit to be increased by a factor of 10 for 4 daily samples. An evaluation of the a priori lower limit of detection for various radionuclides on each gamma spectroscopy [
system indicate that the weekly limits can be satisfied with a minimum count time of 2000 seconds and a sample volume of 1 E+07 cc for the weekly samples. The minimwn sample volume ;
required to meet the daily limit was calculated to be !
8.0 E+05 cc. Based on the sample volumes and count times of L the pre-release samples, these daily a priori limits were ,
satisfied. Hovever, as sample volume increases the sensitivity of the analysis increases. The larger the sample volume, the ;
better the sensitivity and subsequently the lower the limit of -
detection. With respect to the continuous samples collected, the sample volume was almost two orders of magnitude larger than the pre-release sample volume and the analysis sensitivity ,
i increased accordingly. Since the level of activity detected l in the continuous samples was near or below the lower limit of l detection of the pre-release samples, the possibility of I detecting activity at these levels would be only by statistical ;
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Consequences of the Releaset, Although the release of radioactivity occurred through an abnormal pathway, it ves monitored and adequate sampling was performed to assess the radiological impact of the release.
The radiological impact resulting from the release was evaluated in accordance with Offsite Dose Calculation Manual methodologies.
At no time vere any release limits specified in Technical Specifications 3.11.2.1 and 3.11.2.3 exceeded. The total amount of I-131 and Co-58 released were 4.4 and 0.094 uci, respectively. The instantaneous dose rate to a receptor at I the site boundary was calculated to be 0.023 mrem /yr or 0.002 % -
of the Technical Specification Limit. The total projected l maximum organ dose resulting from this release was 0.0034 mrem
- or 0.023 % of the allowable annual limit (0.046 % of the allowable quarterly limit).
Corrective Actions: !
l Although the release occurred through an abnormal pathway, it was monitored and adequate provisions for sampling were taken.
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3.0 CASEOUS EFFLUENTS i
j The quantities of radioactive material released in gaseous effluents are !
summarized in Tables IA, IB, and 1C. Note that there were no elevated I releases, since all Waterford 3 SES releases are considered to be at !
I ground level. h 4.0 LIQUID EFFLUENTS i
The quantities of radioactive material released in liquid effluente are !
summarized in Tables 2A and 28. !
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6.0 METEOROLOGICAL DATA j
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The summary of the hourly meteorological data for this reporting period
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will be included in the Semiannual Effluent Release Report to be l submitted within 60 days after January 1, 1989. I I
7.0 ASSESSMENT OF_ DOSES 7.1 The summary of doses due to gaseous and liquid effluents for this i reporting period will be included in the Semiannual Effluent Release i
Report to be submitted within 60 days after January 1, 1989, 1
8.0 RELATED IhTORMATION l I
8.1 Changes to the Process Control program f l I i There were no changes to the Process Control Program for the period i
] covered by this report.
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t 8.2 Changes to the Offsite Dore Calculation Manual
- There were no changes to the Offsite Dose Calculation Manual for the period this report covers. ;
8.3 Unavailability of REMp Hilk Samples !
Due to the unavailability of three milk sampling locations within ;
five kilometers of the plant, Broad Leaf sampling is performed in :
accordance with Technical Specification Table 3.12-1. Hilk is collected, when available, from the control location and three ;
identified sampling locations as indicated in Waterford 3 Offsite '
Dose Calculation Manual, Table 2 and Table 3. '
8.4 Report if Technical Specification Required Instrument Inoperability Technical Specification, Limiting Condition for Operation (LCO),
3.3.3.10 and 3.3.3.11 requires the reporting in the Semiannual' Radioactive Effluent Releast Report of why designated incperable instrumentation was not restored to operability within the time specified in the ACTION Statement. During the reporting period, there were four separate cases when instrumentation was not restored to opercbility within the time specified. These cases are described in the following sections.
8.4.1 Monitor
Waste Gas Holdup System Hydrogen and Oxygen Moni?. ors Period of Inoperability: 3/21/85 - 06/30/88 (At end of reporting period monitors were still innperable)
Time Required by T(chnical Specifications to Restore Operability: 30 days Cause of Inoperability:
Due to initial design problems excess amounts of moisture were allowed to leak into both the Beckman 02 and Delphi H2 and 02 analyzer systems. Repiacemant of the analyzers and modification of the sample system was implemented during this period.
While this was being done the system remained out of service.
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Reason Operability N2t Restored Within Allotted Time:
Extensive hours were spent attempting to restore these analyzers to operable status. Several analyzer cells were replaced, the solenoid and regulator were repaired, and the sample pump was both repaired and replaced. After these e
ef' orts failed to return the monitor to service, a station modification was initiated to replace the analyzer cells with less moisture sensitive models and to completely redesign the sample line condensate drain system. ,
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This modification entailed work in several areas of the plant and on four different systems. All existing piping and 91ectronics associated with the Waste Gas Holdup System lydrogen and Oxygen Monitors was essentially scrapped and ledesigned.
Work included re-routing all sample lines in the Laundry Room, modifying the existing drain header, and fabrication of a new
] drain header to tie into the Vent Gas Collection Header. i t
ie-routing of the Gas Surge Header Sample line and fabrication i of its drain was performed in Safeguarda Room B. On Gas Decay l Tanr A a new separator and drain line on the Waste Gas Collec- !
tion Discharge Header was added to the second low point.
f Actual work on the Gas Analyzer Panel consisted of (1) adding l 12 new holenoid valves in the sample inlets; (2) modifying the f panel to secommcJate the new exo-sensor units; (3) installing f f
a new pump and its associated tubing; and (4) wiring of all t I
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While testing the system, the Gas Decay Tank "C" sample line
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was found to be crossed with the Gas Surge Tank sample line.
Due to greater pressure in the Gas Decay Tank than that of i
, the Gas Surge Tank the sampling pump diaphragm was blown. !
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This cvent also identified other problems with the system j
- which required correction, i
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Design changes have been made to uncross the lines and install a pressure switch to prevent any subsequert overpressurization.
The pump was replaced and new wiring was installed. Additional moisture traps, pressure regulators, valves and pressure indicators have been installed, a
l A detailed test procedure was prepared and approved. Testing i of the system was completed in early December of 1986 and the system was placed "in-service". Operational checks were performed on the system after the plant returned to the operating mode following a refueling outage. The operational checks demonstrated that while the piping and sample tubing changes proved beneficial, the analyzers themselves proved unreliable due to inherent design problems. Therefore, the Waste Gas }{ydrogen and Oxygen Monitoring System could not be returned to operable status.
Based on the problems encountered with the analyzers and sample / analyzer rystem, a new sample / analyzer system has been designed and installed. The improved system was designed with emphasis on sample conditioning, use of proven analyzers for the application and simple design. Initial testing has j been completed. Initial operation began on June 29, 1988. As j soon as final testing is complete and operating procedures are in place the analyzer system will be returned to cervice, which should be within two months.
8.4.2 Monitor
Waste Gas lloldup System Noble Gas Activity Monitor Period of Inoperability: 4/17/88 to 6/28/88 Time Required by Technical Specifications to Restore Operability: 30 days Cause of Inoterability:
The old Nuclear Measurement Ccrporation type monitor was replaced with a Sorrento Electronics (formerly known as General Atomics Corporation) type monitor.
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Reason Operability Not Restored Within Allntted Time:
Replacement of the Gaseous Waste Management (GWM) system l effluent monitor began on April 17, 1988. The process involved
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removing the old Nuclear Measurement Corporation monitor; '
forming new rkids for the monitor; rerouting sample lines, j power lines, and communications lines; *nd installing the new [
Soerento Electronics monitor. A primary calibration was i performed. Before the calibration could be accomplished a new l c=11bration procedure had to be written and approved in order to evaluate detector energy response, linearity, and response to actual gaseous sources. Once the procedure was approved i and the monitor installed and energized, the calibration was !
l' performed using solid and gaseous calibration sources. The primary calibration and functinnal testing of the monitor was completed on June 28, 198A at which time the monitor was l declared operable and placed back in service.
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8.4.3 Monitor
Boron Waste Management System Effluent Monitor
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Period of Inoperability: 4/13/88 to 6/7/88 Time Required by Technical Specifications to Restore Operability: 30 days i
Cause of Inoperability:
f The old Nuclear Measurement Corporation type monitor
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was replaced with a Sorrento Electronics (formerly kn vn as i General Atomics Corporation) t>Te monitor.
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. l Reason Operability Not Restored Within Allotted Time: ,
l Replacement of the Boron Waste Management (BW) system effluent monitor began by taking the monitor out of service on April 13, 1988. The involved and very time consuming process of removing the old Nuclear Measurement Corporation monitor !
and installing t.he new Sorrento Electronics monitor began. A primary calibration was to be performed, but before this could be accomplished a new calibration procedure had to be written and approved in order to evaluate detector energy response and [
linearity. Once the procedure was approved and the monitor instslled and energized, a primary calibration was performed using solid calibration sources. ,
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The primary calibration and functional t 4 ting of the monitor was completed on June 7,1988 at which t..ae the monitor was '
declared operable and placed back in service, i
8.4.4 Monitor
Liquid Waste Management System Effluent Monitor "
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Period of Inoperability: 4/17/88 to 6/7/88 !
! Time Required by Technical Specifications to Restore i, operability: 30 days Cause of Inoperability:
The old Nuclear Measurement Corporation type monitor was replaced with a Sorrento Electronics (formerly known as I General Atomics Corporation) type monitor.
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c Reason Operability Not Restored Within Allotted Time:
Replacement of the Liquid Waste Management (LWM) system effluent monitor began by taking the monitor out of service on ..pril 17, 1988. The involved and very time consuming proces; af removing the old Nuclear Measurement Corporation monitor and installing the new Sorrento Electronics monitor began. A primary calibration was to be performed but before this could be accomplished a new calibration procedure had to be written and approved in order to evaluate detecter energy response and linearity. Once the procedure was approved and the monitor installed and energized, a primary calibration was performed using solid calibration sources.
The primary calibration and functional testing of the monitor was completed on June 7, 198b 0 which time the monitor was declared operable and placed back in service.
8.5 Missed Effluent Samples On April 13, 1988, it was discovered that a monthly Technical Speci-fication (TS) sampling requirement was missed. TS Surveillance Requirement 4.11.2.1.2 requires a plant stack tritium sample to be taken and analyzed monthly. The surveillance was last performed on March 3, 1988 The tickler card reminding the Chemistry Technician to collect and analyze the tritium sampla had not been placed in the 31 day file. The tickler card remained in the April monthly file.
The root cause of this event was cognitive personnel error due to not filing the surveillance tickler card. This resulted fu the sample not being scheduled. A plant stack tritium sample was taken and analyzed on April 13, 1988. As a result, several Chemistry procedures are be.ing revised to ensure the tickler card file receives supervisory review and to provide clearer instructions for performing plant stack tritium sampling. The plant stack tritium sample is now scheduled by the Station Information Management System (SIMS) computer program.
A complete description of this event and the subsequent corrective actions was reported to the NRC in I.ER 88-007-00.
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8.6 Corrections to Previous Semiannual Radioactive Effluent Release Reports While reviewing the effluent release data covering the period from July 1, 1986 through Dece.nber 31, 1986, a typographical error was found on page 28 of that report. The curies of Co-58 in spent resin solidified with cenient was incorrectly reported as 1.5E+00. The correct value should have been 1.5E+01 curies. The corrected data is included in Attachment 1 of this report.
9.0 TABLES 1 Batch Release Summary 14 Semiannual Summation of all Releases by Quarter - All Airborne Effluents 1B Semiaruual Airborne Continuous Elevated and Ground Level Releases 1C Semiannual Airborne Batch Elevated and Ground Level Releases 2A Semiannual Summation of All Releases by Quarter - All Liquid Effluents 2B Semiannual Liquid Continuous and Batch Releases 3 Solid Waste Shipped Offsite for Disposal 10.0 ATTACHMENTS
- 1. Corrections to the Semiannual D ioactive Effluent Release Report for the period of July 1 to December 31, 1986 l
l l
l W31052611P 23 l
t .
j TABII 1 (1 of 1) ll h
d I
MPORT CATEGORY 84T01 M11 AGE SlMmf IELEASE P0lWT t 4.1.
TYPE F MLIAGE MTQt LIGJID Ale OAGEOUS Pet 100 ST18tf TIM i 0:00 Hts = 12:00AM JAltsty 1,1999 PGt!0D De TIE 4367:59 IftS = II:59PM JLAE 30 1998 LIGJ1B EEMES IUSDt 0F MLEASES I IM TOT 4. TIM FGt E ELEA6ES 44367.0 MlI8JTES Wlp Tilt FGt A lELEAM i 487.0 MiltITES
- 4ftfl0E TIE FOR A M11A6E I 267.3 Mll81TES MINIM TIM FGt A MLIAGE I 32.0 MIMJTES AS40E STEM FLOW 8 717522.3 OPM 049E0VS M11 AGES Id.fWt 0F MLIAGES 12 TOT 4. TIM FOR ALL MLEASES 2733.0 MiltJTES WIM TIE FOR A ELEA6E I 600.0 MiltJTES AelE TIM FGt A ELIA6E 227.8 MIMJTES MINIM llE FOR A MLEA6E 59.0 MIMJTES i
W310526HP 24
TAB 12 1 A (1 of 1)
REPORT CATEGORY SEMIN00L SLff%TICN OF N.I. ELEASES BY MTS TYPE OF ETIVITY 6 ALL A!RBO M EFFLUCITS MPORTING PG100 t MTER I 1 M QUMTER I 2 i (MIT IQUMTER 1 IQUARTG 2 IEST. TUT 4.1
- IHOLRS IHOLRS EMOR% :
TYPE (# EFFLLENT 1 1-216012161-4344 A. FIS$10N NO ACT!YATION PRODUCTS
- 1. TOTAL ELEASE :(1 RIES 2.49E 03 1.91E 03 : 1.50E 01
- 2. AVDAGE MLEASE MTE FOR PG100 IUCl/SEC 3.21E 02 2.43E 02 8
- 3. PDCENT OF M9t!CARLE LIMIT I 1 N/A I WA
- 3. M010100lMS
- 1. TOTAL IODI M-131 IC1 RIES I 8.63E-05 t 7.73E44 1.50E 01:
- 2. MMOE MLEASE MTE FOR PD100 IUC1/SEC 81.11E-05 t 9.83E451
- 3. PDCENT OF #9LICABLE LIMIT 1 I N/A N/A C. PMTICLLATES
- 1. PMilC1LATES(HN.F-LIWS)6 DAYS) ICLRIES 3.29E-07 2.62E-04 1.50E 01
- 2. AVERADE ELEASE MTE FOR PER100 IUC1/SEC i 4.23E-00 t 3.34E-05 I I WA :
3[ PGCENT OF M9LICABLE 1 LIMITWA
- 4. CROSS N.PM RN)!OACTIVITY ICLRIES 3.14E-05 6.65E 05 :
D. TRITitM
- l. TOTN. ELEASE ICLRIES 4.33E 01 8 1.73E 01 1.50E 011
- 2. AVDA0E MLEASE MTE FOR PG100 IUC!/SEC i 5.57E 00 8 2.20E 00 t
- 3. PGCENT CF #ftlCAILE LIMIT 1 : M7A I WA W310526HP 25
TABLE IB (1 of 1) l MPORTCATE00RY SEN!MOGA. AllWOftE CGITiltJOUS ELEMTED NO CROL90 t LENEL El.EAKS. TOT 4.S FOR EADI ptJC1.!E E1 EASED.
TYPE (F ACTIVITY FISSICBI OASES 100! PES. 40 PNtTICtA.ATES fEPORTIND PERl(B QJNtTER I 1 NW QuNtTER 0 2 s E11MTED ELEASES I OR0l80 fELIASES I L4tlT IGUNtTER 1 IGUNtTER 2 IGUARTER 1 IQUNtTER 21 1 3HOLRS IH0WtS IH00tS IH0WtS I ItXLIDE I l-2160 82161-4344 I 1-2160 t2161-4344 FISSION OASES IE-131M i CutlES 0.0lC4110.00E418 4.52E M 0.0(E-01 :
TJ 133 CutlES ^.00E41 10.00E41 2.40E 031 1.20E 03 I h.-135 i M IES 0.00E418 0.00E-01 I 9.21E 01 13.50E 01 TOT 4. FOR PER100 CARIES I 0.00E41 0.00E-0112.49E 03 1.23E 03 100lpES 1-131 8 (tRIES 1-133 0.00E41 0.00E-018 8.63E-05 I 7.TJE-04
- CutlES I 0.00E-01 1 0.00E-01 1.4X-06 1.26E-06 :
TOTM. FOR PER!00 : cut!ES 0.00E4110.00E4118.78E4517.74E44 :
PNtTICLLATES H-3 CR-51 CtRIES I 0.00E-01 4.00E 018 4.33E 011 1.67E 01 :
i CtRIES 0.00C41 1 0.N 01 1 0.00E-01 8 6.47E-05 W 54 C0-58 : Cut!ES I 0.00E41 10.fA 61 t 0.00E-01 14.90E-06 :
CD-60 t (1 RIES I 0.0(E41 10.00E-01 0 00E-01 I 9.46E-05 8 IR-95 : CutlES I 0.00E-01 10.00E41 0.00E41 12.41-05 :
W 95 : C1 RIES I 0.00E41 1 0.00E-01 8 0.0E41 2.33E-05 D 103 i CtRIES I 0.0(E41 10.0(E-01 0.00E-01 4.26E 05 :
&l06 CtRIES I 0.00E-01 10 0(E-0110.00E-01 145E 07 :
(1 RIES 0.00E41 0.00E-01 10.0(E-01 1.99E46 8 CS-134 CtRIES I 0.00E41 1 0.00E-01 8 0.
C$-137 I (1 RIES t 0.00E-01 1 0.00E-01 ZYE-0711 3.Q0E41 2.94E-06 8 2.95E-07 0 AliM I CtR1ES 0.0(E-01 1 0.00E-01 1 3.14E-05 I 6.65E 05 i
+ 20) : CutlES I 0.00E41 10.00E 01 10.00E-Cl I 4.03E47 TOTN. FOR PERIOD CLRIES 0.00E-0110.00E-01 I 4.33E 01 1.67E 01 I 6310526HP 26 l
TABLE 10 (1 of 1) lEPORT CATEGORY SEMIM601. AIRB015E RAT 04 ELEWTO MS On0L30 LIVEl. IELEASES. TOTALS FCft EA04 ItlCI.!E IEl.EAED.
TYPE (F ACTIVITY I FISSION 046ES,100!IES. 40 PNtT101ATES lEPORill0 PUt!(2 i GJNtTER 01 NG GUNITER 0 2 I E!IWTD IELIAES GROLSe IE1IAGES I I WIT IEINtTER 1 IGUARTUt 2 IQUARTFR 180 UNITER 2 tHolfts IHOLftS IMDLftS IHOLftS I MKl.!DE 1-2160 12161-4344 1-216012161-434']
FISSIONGASES IGt-65M i MIES I 0.00E-01 10.00E-01 10.00241 2.34-01 :
IGt-65 MIES 0.00E41 0.00E41 10.00E41 1 1.21E 01 I tJt-6B t QAtlES I 0.00E41 1 0.00E-01 1 0.00E41 1 1.57E-01 E-131M i MIES I 0.00E-018 0.00E41 10.00E41 1.18E 01 XE- 132 i MIES I 0.00E41 10.00E41 10.00E-018 4.50E 00 t E-133 I QAtIES I 0.00E41 10.00E-01 10.00E-01 I 6.43E 02 IE-135 I QAtlES I 0.00E-0110.00E4110.00E 018 4.01E 00 :
Nt-41 QAt1ES 0.0E-01 1 0.00E-01 1 0.00E-01 1 3.8R-01 :
TOTN. FOR PERIOC CURIE 3 10.00E-018 0.00E-0110.00E41 I 6.76E 02 1001ES IDE P#tT101ATES
, H-3 QPIES ! 0.00E4110.00E41 0.00E-0115.59E-01 :
l l
6310526HP 27 l -
TABLE 2A (1 of 1)
M.' ORT CATE00RY TYPE OF ACTIVITY SEMIA100ll Suf% TION 7 ALA. M1BdES BY MTER t ALL LIWID EFRENTS MPORTING PERIOD MTER 01 M MTER I 2
$li t INTER 1 INTER 2 IEST.TOTALI IHOURS HOURS TYPE OF EFFLENT ERROR 1 I l-216012161-4344 1 A. FISSION M ACTIVATION PRODUCTS
- 1. TOT 4. MLEASE(NOT INCLt2IMO I t :
TRlillM. OMES. ALPHA) : :
I M IES : 1.76E-01 8 6.45E-01 1.50E 01:
- 2. AMCE DILUTED CONCDmMTION : I 0$th0 PERIOD a t IUCtht. I 2.14E49 81.68E-081
- 3. PERCENT & 49t! CABLE LIMIT I I : WA N/A :
B. TRIT M
- 1. TOT 4. MLEASE I M IES : 1.23E 02 2.24E 01 1.50E 01:
- 2. MRAGE O! LUTED CONCDmMTION I I I t 041MO PERIOD IUCl/M. I 1.49E-06 5.84E-07 8
- 3. PDCENT F #PLICAR.E LIMIT 1 : WA WA t C. DISSOLVED M DmelED CASES b TOT 4. RELDSE I M IES : 4.22E 01 5.22E 001 1.5(E 01
- 2. AGIN DILUTED CONCDmMTION : I M i m PERIOD tVCint. t 5.12E47 81.36E47
- 3. PEMDIT OF MPt.lCABLE LIMIT 1 : WA WA i D. CROSS N.PM RADIDACTIVITY
- 1. TOTN. REl. EASE I M IES i -
- 5.29E-06 8 3.34E-06 1.50E Oli E. MSTE VOL MLEASED(PRE-DILUTI(M) 804Li 1.45E 06 7.10E 05 1.50E 01:
F. YCLtPE OF DILUTICM MTER USED 104.
t 2.18E 10 4 1.01E 10 t 1.50E 01:
W310526HP 28 l
t
i
, , k TABLE 2B (1 of 2)
'1 l
l l
IEPORT CATE00RY SEMI 40tJAL LIQUID CONT!IGJOUS Ale BAT 01IELEASES !
TWE (F ACTIVITY TOTALS FOR EA04 IGJQ.lDE lELEASED.
t 41 RA0lG8JQ. IDES lETRTIND PER100 M RTER I 1 # e QUNITER 8 2 t CONTIMJOUS IELEASES 1 94T04 RELEASES IMIT I lluftTER 1 IGUARTER 2 IGUARTER I IGUNtTER 2 IMuts IHOURS tH0UtS HOUt$ I IEl(LIDE I i 1-216082161-4344 1-2160 82161-4344 4 1 RKl.! DES l+ 3 NA-24 : MIES I 0.0(E-018 0.00E-011 1.2X 02 : ?.24E 01 Ot-51 : M1ES MIESI 0.00E41 10.00E41 12.21E-04 t o.77E44 196-54 0.00E-01 1 0.00E41 1 6.4X-05 3.5GE42 FE-55 MIES I 0.00E41 1 0.00E-01 1 1.61E44 1 2.7E-03 FE-59 MIES I 0.00E4110.00E 0112.3tE42 8 5.55E-02 :
CD-56 MIES I 0.0(E4110.0(E41 1.03E-04 8 2.77E43 (D-60 i MIES MIES10.0(E41 10.00E41 1 1.0E42 2.70E41 :
0.00E41 10.00E-01 19-08 SR-69
- M IES 0.00E41 1 0.00E41 1 1.61E-03 1.3E42 8 8.7542 1 0.00E-01 :
MIES 0.00E4110.00E4112.89E-04 81.51E N I R3 ZR-97 :l M BlRIES i8:H41!8:N11!!:M3l1:W3i 0.00E-01 1 0.0(E41 1 0.00E41 I 5.'62E-05 10-95 : MIES 10-99 0.00E41 10.00E41 2.51E441 1.0E-021 TC-99M M IES I 0.0(E-01 1 0.00E 41 1 0.0(E 41 1 1.0E-03 I RS103 ii MIES MIESI 0.00E-01 10.00E-01 I 8.39E-06 1.05E43 i 0.00E-01 1 0.00E41 1 0.00E-01 1.13E43 i M-110M i MIES 0.00E41 10.0(E418 3.90E-0512.21E-03 TE-132
- I-131 :i MIES I 0.00E41 1 0.00E41 1 0.00E41 1 8.1E-04 1 1-132 M!ES I 0.00E41 0.00E4119.12E43 8 8.69E421 1-133 M IES I 0.00E-01 1 0.00E-01 1 0.00E-01 4 3.92E-04 1 MIES 0.00E4110.0(E41 I 6.95E-04 8 3.3E 03 :
CS-134 CS-136 : M1ES t 0.00E41 10.0(E 01 1 1.3X421 1.90E 02 CS-137 I M MIESIES t 0.0(E-01 1 0.00E41 1 4.72E-05 1 2.91-04 :
CS-138 : 0.0(E-01 10.00E-01 t 1.7E42 t 2.0M-02 94-139 MIES MIES I 0.00E-01 0.00E41 l 3.1E-03 I 0.00E 01 i 94140 0.00E4110.00E4114.07E 03 8 4.59E-04 8 MIES I 0.00E41 1 0.00E 01 1 2.30E-051 0.0(E-01 U310526HP 29
's .
i TABLE 2B '
(2 of 2) l l
FEPWIT CATEO Rf a SEMIN84#t. LIR)l0 CGITilt.XUS N4 BATQi IELEASES TYPE OF ACTIVITV t TOTALS FOR EADI MIM IE38ED.
I E RADIGIIUQ.! DES IEPORTING PQ100 t QuNtTER 81 NG QUARTER 8 2 8 CD(TIMJOUS RELEMES I BATQi RELEMES i I L811T IGUARTER 1 IQuNtTER 2 IGjnRTER i IQUARTER 2 :
I IHOWts IHOWtS IHOWtS IN0WtS IWCLIDE : I l-2160 12161-4344 8 1-216082161-4344:
E MJCLIDES CONil;tED LA-140 CtRIES
&-141 1 0.00E41 10.00E-018 5.10E-0413.50E I E-144 cut!ES I 0.00E-01 I 0.006-01 10.00E-018 3.0F C1 RIES W-187 0.00E41 I 0.00E-01 0.00E-01 14.Or.-04 :
- CutlES I 0.00E41 1 0.00E41 1 0.00E-41 1 2.00E44 :
CE KR 85 l NM i 8:23 ! 8:N* ! tN3 !!.M3!
CutlES
@ 47 i CutlES 0.00E41 1 0.00E41 1 1.37E-01 1 3.85E-03 G48 Cut!ES 0.0M-01 10.00E41 18.88E44 I 0.00E41 4E-1311t t 0.00E41 1 0.00E41 1 5.57E43 1 0.0M41 8 XE-1334 CLRIES I 0.00E41 0.00E418 3.30E-0114.39E42 XE-133 i CLRIES I 0.00E4110.00E41 2.49E4113.01E42 XE-135 iI CtRIES I 0.00E41 10.00E-018 4.13E 01 15.13E 001 Nt-41 : CUI1ES I 0.0/141 0.00E411 1.00E4116.73E43 i SR-90 *I Cut 1ES 0.00E41 0.00E41 1.29E-05 I 0.0M41 i 0 ALPHA QJRIES~ I 0.00E-01 10.0E418 7.86E413.45E4 I CO-57 : CtRIES I 0.0M41 10.00E-018 5.29E4 I 3.34E4 98-124 : CLRIES I 0.00E418 0.00E-018 0.00E41 13.67E-04 s CtRIES I 0.00E-01 0.00E41 14.49E-04 8 3.27E42 8
$16-113 Cut 10-97 cut :ES I 0.00E41 1 0.00E41 1 3.llE-05 1 4.59E43 SB-122 cut ES ES I 0.00E41 10.0(E41 I 8.16E45 t 1.05E43 i t 0.00E418 0.00E4115.60E S8-125 Cut ES I 166E42 :
E-139 0.00E4110.00E41 I 3.12E I 3.'36E-02 i 50-127 : CLRIES I 0.0M41 1 0.0M41 8 1.30E-051 0.0M41 1 CtRIES I 0.00E-0110.00E41 0.00E4111.iM431 10TN. FOR PD100 CLRIES 0.00E41 0.00E411 1.65E 02 2.83E 01 B310526HP 30 i __ ._
TABLE 3 (1 of 4)
Solid Waste Shipped Offsite for Disposal .
During Period 1-1-88 thru 6-30-88 Container Volume Waste Volume Total Activity Waste Type Volume (ft 3) (m3 ) (Ci) % Error (C1)
Compacted Dry 95 142.59 2.57 i 25%
Active Waste Non Compacted Dry 95 15.91 2.96 i 25%
Active Waste 182 Liquid Waste 182 20.6 23 1 25%
Management Spent Resin Solidified With Cement Resin Waste 182 5.15 176 25%
Management &
Liquid Waste Management Spent Resin Solidified With l Cement i
L W310526HP 31
c i, ., , .* *-
L '.
TABLE 3 (2 of 4)
Estimates of Major Nuclides By Waste Type t
NUCLIDE' PERCENT CURIES NAME- ABUNDANCE Compacted Dry Cs-137 41.932% 1.08E+00 s
Active Waste Co-58 19.423% 4.99E-01 Cs-134 18.885% 4.85E-01 Fe-55 9.860% 2.53E-01 Co-60 3.745% 9.62E-02.
Ni-63 2.235% 5.74E-02 Mn-54 2.099% 5.39E-02 1-131 1.645% 4.23E-02 C-14 .178% 4.57E-03 Ni-59 .000% 0.00E+00 Nb-94 .000% 0.00E+00 H-3 .000% 0.00E+00 Sr-90 .000% 0.00E+00 Tc-99 .000% 0.00E+00 I-129 .000% 0.00E+00 Pu-241 .000% 0.00E+00 Cm-242 .000% 0.00E+00 NUCLIDE PERCENT CURIES NAME ABUNDANCE Non-Compacted Dry Cs-137 38.820% 1.15E+00 Active Waste Co-58 21.364% 6.34E-01 Cs-134 17.781% 5.28E-01 Fe-55 9.235% 2.74E-01 1-131 5.054% 1.50E-01 Co-60 3.490% 1.04E-01 Ni-63 2.067% 6.14E-02 Mn-54 2.024% 6.01E-02 C-14 .165% 4.89E-03 Ni-59 .000% 0.00E+00 Nb-94 .000% 0.00E+00 H-3 .000% 0.00E+00 Sr-90 .000% 0.00E+00 Tc-99 .000% 0.00E+00 I-129 .000% 0.00E+00 Pu-241 .000% 0.00E+00 Cm-242 .000% 0.00E+00 W310526HP 32
TABLE 3 l (3 of 4)
Estimates of Major Nuclides By Waste Type NUCLIDE PERCENT CURIES NAME ABUNDANCE Liquid Waste Co-58 44.772% 1.04E+01 Management Cs-137 24.087% 5.59E+00 i System Spent Cs-134 13.003% 3.02E+00 Resin Solidified Fe-55~ 8.306% 1.93E+00 ,
With Cement Co-60 3.135% 7.28E-01 '
I-131 2.550% 5.92E-01 Ni-63 1.860% 4.32E-01 Mn-54 1.666% 3.87E-01 H-3 .474% 1.10E-01 C-14 .148% 3.43E-02 Ni-59 .000% 0.00E+00 Nb-94 .000% 0.00E+00 Sr-90 .000% 0.00E+00 Tc-99 .000% 0.00E+00 I-129 .000% 0.00E+00 Pu-241 .000% 0.00E+00 Cm-242 .000% 0.00E+00 l NUCLIDE PERCENT CURIES l NAME ABUNDANCE l
l Resin Waste Management Cs-137 47.281% 8.29E+01
& Liquid Waste Cs-134 24.753% 4.34E+01 Management Spent Resin Co-58 11.578% 2.03E+01 Solidified with Cement Ni-63 6.160% 1.08E+01 Co-60 3.867% 6.78E+00 Mn-54 3.222% 5.65E+00 Fe-55 3.091% 5.42E+00 C-14 .029% 5.08E+02 H-3 .020% 3.47E-02 Ni-59 .000% 0.00E+00 Nb-94 .000% 0.00E+00 Sr-90 .000% 0.00E+00 Tc-99 .000% 0.00E+00 1-129 .000% 0.00E+00 Pu-241 .000% 0.00E+00 Cm-242 .000% 0.00E+00 W310526HP 33
TABLE 3 (4 of 4)
Solid Waste Disposition Summary Number of Mode of Shipments Transportation Destination 11 Truck Beatty Waste # of Type of Type of Mode Destination Class Shipments Shipments Container A 10 LSA Strongtight Truck Beatty B 1 LSA Type A Truck Beatty W310526HP 34
r l- , e. , e, l
l ATTACl&fENT 1 CORRECTIONS TO THE SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT FOR THE PERIOD OF JULY 1 TO DECEMBER 31, 1986 W310526)lP 35
-~
m y '*o ~
, \
TABLE 3 (3 of 4)
C-14 .04 9.1E-04 Sr-90 .04 .0E+00 ,
. Tc-99 .0% .0E+00 '
I-129 .0% .0E+00 Cs-134 11.8% 7.0E-01 Cs-137 22.8% 1.4E+00 Co-141 1.8% 1.1E-01 Pu-241 .' 2 % 3.5E-03 Cm-242 .0% 5.6E-07 LIQUID WASTE Mn-54 4.5% 5.5E-02 MANANGEMENT SYSTEM Co-58 79.3% 9.6E-01 SPENT RESIN Co-60 .9% 1.1E-02 DEWATERED Ni-59 .0% .0E+00 Ni-63 .4% 5.0E-03 ,
Nb-94 .0% .0E+00 Sb-124 3.4% 4.1E-02 H-3 .2% 2.1E-03 C-14 .0% .0E+00 Sr-90 .04 .0E+00 Tc-99 .04 .0E+00 Cs-134 2.1% 2.6E-02 Cs-137 5.7% 6.9E-02 I-131 2.1% 2.6E-02 Pu-241 .0% 9.5E-07 Cm-242 .0% .0E+00 LIQUID WASTE Mn-54 4.6% 1.1E+00 MANAGEMENT SYSTEM Co-58 59.0% 1.5E+01 AND RESIN WASTE Co-60 4.0% 9.7E-01 MANAGEMENT SYSTEM Ni-59 .0% 8.5E-03 SPENT RESIN Ni-63 2.2% 5.4E-01 SOLIDIFIED WITH Nb-94 .0% .0E+00 CEMENT H-3 .1% 1.5E-02 C-14 .0% 2.2E-03 Sr-90 .0% 7.6E-03 Tc-99 .0% .0E+00 I-129 .0% .0E+00 Cs-134 9.5% 2.3E+00 Cs-137 16.6% 4.1E+00 Sr-89 1.8% 4.4E-01 Pu-241 .0% 3.8E-03 Cm-242 .0% 4.9E-05
- SOLID WASTE DISPOSITION
SUMMARY
NUMBP1 0F SHIPMENTS MODE OF TRANSPORTATION DESTINATION
~
2 TRUCK BARNWELL 5 TRUCK RICHLAND 0 TRUCK BEATTY 0 TRUCK OTHER NUMBER OF TYPE OF 28 TYPE MODE OF
jh~~1 1
R2 fort 10CFR50.36s ;
l
- P. O. BOX 60340 LOUISI& AN POWER LIGHTA / 317NEWBARONNESTREET ORLEANS, LOUISIANA 70160 * (504)595 3100 UrT00bsY$
August 29, 1988 l W3P88-1268 A4.05 QA U.S. Nuclear Regulatory Cor: mission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Semiannual Radioactive Effluent Release Report Enclosed is the subject report on effluent releases which covers the period of January 1 through June 30, 1988. This report is submitted per Section 6.9.1.8 in the Waterford 3 Technical Specifications (NUREG-1117) of Appendix A to Facility Operating License No. NPF-38 and 10CFR50.36a(a)(2),
pursuant to 10CFR50.4.
Very truly yours, R.F. Burski Manager Nuclear Safety & Regulatory Affairs RFB BGM ssf Enclosure cc (w/ enclosure): R.D. Martin, NRC Region IV NRC Resident Inspectors Office cc (w/o enclosure): J.A. Calvo, NRC-NRR D.L. Wigginton, NRC-NRR E.L. Blake W.M. Stevenson
/
"AN EQUAL OPPORTUNITY EMPLOYER"
/
/l