ML20212F430
| ML20212F430 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 12/31/1986 |
| From: | Hintz D WISCONSIN PUBLIC SERVICE CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-87-25 NUDOCS 8703050084 | |
| Download: ML20212F430 (111) | |
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Y 5 .. [ KE.WAUNEE NOCLEA'R POWER IPLiANT
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'A NN UA[ OPERATING REPORT;
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' WISCONSIN PUBLIC SERVICE CORPORATION l 4
--WISCONSIN POWER 8 LIGHT COMPANY 2 M A DISON GAS a ELECTRIC COMPANY e &
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TABLE OF CONTENTS 1.0 Introduction -
2.0 Summary of Operating Experience ,
3.0 Plant Modifications, Tests, and Experiments 4.0 Licensee Event Reports
-5.0 Fuel Inspection Report.
6.0 Challenges to and Failures of Pressurizer Safety and Relief Valves 7.0 Steam Generator Tube Inspection 8.0 Personnel Exposure and Monitoring Report 9.0 Radiological Monitoring Program O
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1.1
1.0 INTRODUCTION
Th'e Kewaunee Nuclear Power Plant is a pressurized water reactor licensed at 1650 Wt. It is located in Kewaunee County, Wisconsin,-along Lake_
Michigan's northwest shoreline and is jointly owned by Wisconsin'Public Service Corporation, Wisconsin Power and Light Company and Madison Gas and Electric Company. The nuclear steam supply system was purchased from Westinghouse Electric Corporation'and is rated for a 1721.4 MWt output.
The turbine-generator was also purchased from Westinghouse and is rated at- l 535 MWe net. The architect / engineer was Pioneer Service and Engineering (PSE) from Chicago.
The Kewaunee Nuclear Power Plant achieved initial criticality on March 7, 1974. Initial power generation was reached April 8, 1974,'and the plant was declared commercial on June 16, 1974. ' Since being declared commer-cial, Kewaunee has generated 45,495,886 MW hours ~of electricity as of December 31, 1986, with a net plant capacity factor of 77.3 (using net
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DER).
o 1.1 Highlights l During the year, the Kewaunee Nuclear Power Plant was primarily base loaded. The unit was operated at 87.5% capacity factor (using net MDC)
- l. with a gross efficiency of 33.4%. The unit and reactor availability were 85.8% and 86.6% respectively. Table 2.1 is a compilation of the l monthly summaries of the operating data, Table 2.2 contains the yearly and total summaries of the operating data, and Figure 1.1 provides a histogram of the average daily electrical output of the Kewaunee Plant for 1986.
o
2.1 2.0 SUW ERY OF OPERATING EXPERIENCE Q' January Normal power operation continued through the month of January.
! The turbine Stop and Governor Valve Test was not performed this month due to the RCS boron concentration being less than 150 PPM (as allowed by the Technical Specifications).
PLANT SHUTDOWNS: There were no shutdowns during the month of January.
February Normal power operation continued through the month of February.
The Turbine Stop and Governor Valve Test was not performed this month due to-the RCS Boron concentration being less than 150 PPM _(as allowed by the Technical Specifications).
PLANT SHUTDOWNS: Commenced the Cycle 11-12 refuling outage on February 28.
February 28: Unit load was reduced to 44.5% for end of life physics testing, o)
( March In March; the Cycle 11-12 refueling outage continued.
PLANT SHUTDOWNS: Scheduled shutdown of 744.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
April The Cycle 11-12 The refueling outage concluded on April 20, and there were two plant trips during the month of April.
PLANT SHUTDOWNS:
April 1: Scheduled shutdown of 465.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Continued Cycle 11-12 refueling outage. On April 20, The refueling outage was concluded, and the unit was placed in service for the required turbine warmup period prior to testing.
April 20: Forced shutdown of 13.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. A reactor / turbine trip occurred when the binding in one Feedwater Regulating Valve caused it to open suddenly while transferring to the Feedwater Regulating Valves from the Feedwater Bypass Valves.
April 21: Scheduled shutdown of 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform turbine overspeed testing and turbine generator torsion testing.
April 24: Forced shutdown of 10.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. A reactor / turbine trip occurred when one Feedwater Regulating Valve failed to respond properly while transferring Feedwater Regulating Valve control between manual and automatic modes for surveillance testing.
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( ) Normal power operation continued, except for one plant trip, during the month of May.
PLANT SHUTDOWNS: l
. . s May 13: Forced shutdown of 4.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. A reactor /tdrbine trip occurred' during routine monthTy NIS channel testing when J & C technicians..
adjusted currents for a channel that was not the channel aligned 4
for testing. In conjunction with this out ye, the monthly -
Turbine Stop and Control Valve Test was performed. ,
,i June ,'.
Normal power operation continued through the month of June.
On June 15 the unit load was reduced to 387 MWE Gross to perform the monthly Turbine Stop and Control Valve Test. The unit was ,
returned to full load the samc day, , j .-
PLANT SHUTDOWNS: There were no shutdowns during the month ~of. 1une.
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July Normal power operations co'ntinued through the month of July.'
On July 14 the unit load was reduced to 390 MWE Gross to perfore the ,
monthly Turbine Stop and Control Valve Test. The uni.t was returned to full load the same day. i PLANT SHUTDOWNS: There were no shutde'ns.during tho bonth of July.
August Normal power operation continued through-the month of August. -
August 17: The unit load was reduced to 390 MWE Gross to perform the monthly Turbine Stop and Control Valve Test. The unit was returned to full load the same day.
August 21: The unit load was reduced to 225 MWE Gross for' inspection of the Condenser Waterboxes. s August 23: The unit load was reduced to 225 MWE Gross for r. leaning of the Condenser Waterboxes. The unit was returned tesfull lead tha same day.
PLANT SHUTDOWNS: There were no rhutdown: during the month of August.
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U Normal power operation continued through the month of September.
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- On September 21 the unit load was reduced to 388 MWE Gross to perform the monthly Turbine Stop and Control Valve Test. The unit was returned to e full load the same day.
PLANT SHUTDOWNS: There were no shutdowns during the month'of September.
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October .
Normalpobroperationcontinuedthrough'themonthofOctober.
October 15: The unit was reduced to 325 MWE Gross to repair a lube oil
,l 1eakonaFeepwaterPumpmotorinboardbearing.
The Turbine Stop and Control Valve Test was also performed.
The unit vac returned to full load the same day.
October 17: The unit load was reduced to 320 MWE Gross to repair a lube oil leak on s Feedwater Pump outboard motor bearing. The unit was returned to full load the same day.
, October 28: Theunitihadwasreducedto273MWEGrosstorepaira
. Condensate Pump motor lube oil cooler. A Turbine Stop and Control Valve Test was also performed. The unit was returned'to full load the same day.
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PLANT SHUTDOWNS: There were no shutdowns during the month of October.
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l Novenhg Normal power operation continues through the month of November.
. -November 23: The unit load was reduced to 390 MWE Gross to perform the monthly Turbine Stop and Control Valve Test. The unit was
,. returned to full load the same day.
PLANT SHUTDOWNS: There were no shutdowns during the month of November.
4 December Normal power operation continued through the month of December.
December 21: The unit load was reduced to 390 MWE Gross to perform the monthly Turbine Stop and Control Valve Test. The unit was returned to full load the same day.
PLANT SHUTDOWNS: There were no shutdowns during the month of December.
As of December 31, Kewaunee is in its second longest continuous run with 232 consecutive days of operation.
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TABLE 2.1 (Page 1 of 2)' ELECTRICAL POWER GENERATION DATA (1986) MONTHLY- 1 January February March April M- June-Hours RX was critical 744.0- 671.8 0.0 289.8 741.7 720.0-RX Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 0.0 Hours Generator On-Line 744.0 671.5 0.0 223.0 739.7 720.0 Unit Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 '0.0-Gross Thermal Energy Generated (MWH) 1,217,881.0 1,067,018.0 0.0 267,903.0 1,191,848.0 .1,176,926.0 l Gross Elec. Energy Generated (MWH) 407,400.0 357,800.0 0.0 88,100.0 401,000.0 396,200.0. ! Net Elec. Energy Generated (MWH) 389,259.0 341,664.0 0.0 83,122.0 382,373.0 377,561.0 RX Service Factor '100.0 99.9 .0.0 40.3 99.7 100.0 ! RX Availability Factor 100.0 99.9 0.0 40.3 99.7 100.0 Unit Service Factor 100.0 99.9 0.0 31.0 99.4 100.0 Unit Availability Factor . 100.0 99.9 0.0 31.0 99.4 , 100.0 Unit Capacity Factor (using MDC net) 104.0 101.1 0.0 23.0 102.2 104.3 Unit Capacity Factor (using DER net) 97.8 95.1 0.0 21.6 96.1 98.0s Unit Forced Outage Rate 0.0 0.0 0.0 9.8 0.6 -0.0 Hours in Month 744.0 672.0 744.0 719.0 744.0 -720.0 Nst MDC (Mwe) 503.0 503.0 503.0 503.0 503.0- 503.0 -[
4 <'
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i i I l' TABLE 2.1 (Page 2'of 2) [ 4 ELECTRICAL POWER GENERATION DATA (1986) MONTHLY l j July August September October November December i l Hours RX was critical 744.0 744.0 720.0 745.0- 720.0 744.0 l l RX Reserve Shutdown Hours 0.0 0.0- 0.0 0.0 0.0 . 0.0 { Hours Generator On-Line 744.0 744.0 720.0 745.0 720.0 744.0 Unit Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 0.0 j Gross Thermal Energy Generated (MWH) 1,218,187.0 1,179,835.0 1,180,207.0 1,202,091.0 1,179,133.0' 1,215,853.0 q Gross Elec. Energy Generated (MWH). 405,600.0 387,000.0 -396,600.0 403,600.0 394,400.0 406,460.0 f Net Elec. Energy Generated (MWH) 386,122.0 367,650.0 377,898.0 384,305.0 376,443.0- 388,277.0 ! RX Service Factor 100.0 100.0 100.0 100.0 100.0 100.0 I 1 { RX Availability Factor 100.0 100.0 100.0 100.0 100.0 100.0 Unit Service Factor 100.0 100.0 100.0 100.0 100.0 100.0
~
Unit Availability Factor 100.0 100.0 100.0 100.0 100.0 100.0 i i { Unit Capacity Factor (using MOC net) 103.2 98.2 ~ 104.3 102.6 .103.9 103.8 !~ Unit Capacity Factor (using DER net) .97.0 92.4 98.1 96.4 97.7, 97.5 Unit Forced Outage Rate 0.0 0.0 0.0 0.0 0.0 0.0-j Hours in Month 744.0 744.0 720.0 745.0 720.0 744.0-Nat MDC (Mwe) 503.0 503.0 503.0. .503.0- 503.0 503.0 ;,, L I . 4 .
2.6 TABLE 2.2 ELECTRICAL-POWER GENERATION DATA m. d- 1986 , Year Cumulative Hours RX was critical 7,584.3 93,601.3 RX Reserve Shutdown Hours 0.0 2,330.5 Hours Generator On-Line 7,515.2 92,070.7 Unit Reserve Shutdown Hours 0.0 10.0 Gross Thermal' Energy Generated (MWH) 12,096,882.0 144,813,714.0 Gross Elec. Energy Generated (MWH) 4,044,100.0 47,782,600.0 Net Elec. Energy Gensrated (MWH) 3,854,674.0 45,495,886.0 RX Service Factor 86.6 85.1 RX Availability Factor 86.6 87.2 Unit Service Factor 85.8 83.7 Unit Availability Factor 85.8 83.7 Unit Capacity Factor (using MDC net) 87.5 80.1 Unit Capacity Factor (using DER net) 82.2 77.3 Unit Forced Outage Rate' O.4 3.1 Hours in Reporting Period 8,760.0 109,969.0 0
3.1 3.0 PLANT MODIFICATIONS, TESTS AND EXPERIENTS O u This section is provided in accordance with the requirements of Part 50.59(b) to Title 10,' Code of Federal Regulations (10CFR50.59(b)). This regulation allows licensees to make changes in the facility as described in the Updated Safety Analysis Report, and conduct tests and experiments not described in the Updated Safety Analysis Report, without prior NRC appro-val, provided the change, test or experiment does not involve a change in the Technical Specifications or an unreviewed safety question. 10CFR50.59(b) requires that such changes be reported on an annual basis. Plant Modifications, 10CFR50.59 There were no modifications during 1986 which introduced an unreviewed
- safety question and, therefore, prior NRC approval was not required.
The following summary of modifications includes those significant modifi-cations completed during 1986 and not previously reported. Many of these modifications are not specifically required to be reported by 10CFR50.59(b) since they do not constitute a change in the facility "as described in the Updated Safety Analysis Report." -However, they are con-sidered to be of significance, warranting mention in this summary report. Inadequate Core Cooling Monitoring (ICCM) The existing Core Exit Thermocouple system was replaced from sensor to
-p v
display (DCR 1163). The previous Combustion Engineering Subcooling Margin Monitor was replaced with two train-oriented monitors. A Reactor Inventory Tracking system was added which provides the operator with reac-tor coolant void fraction with reactor coolant pump (s) running, and reac-tor vessel level with both pumps off (DCR 1077). Displays for the three systems are on a single ICCM panel. Summary of Safety Evaluation The replacement of the old equipment increases the margin of safety by providing redundant trains of environmentally qualified equip-ment. The addition of a reactor coolant inventory system with upgraded core exit thermocouples and subcooling margin monitors pro-vides an ICCM package which offers additional information displays from which the operator can choose to aid him in events such as steam generator tube ruptures, control system upsets, pump seal failures, overcooling events due to secondary side disturbances or small break LOCAs. Reactor Control and Protection The reactor trip circuitry was modified to provide actuation of the shunt trip attachment, as well as the undervoltage attachment, on the main reac-tor trip breakers on receipt of an automatic reactor trip signal. The installation included test switches and lights to verify proper operation of the shunt trip and undervoltage trip independently. (DCR 1460)
3.2
' Summary of Safety Evaluation .vf]. This modification increases the reliability of the reactor trip areakers by adding a second, diverse method of automatically opening the reactor trip breakers. -l l
Environmental Qualification Electrical equipment in various systems was upgraded / replaced to improve /. extend _its environmental qualifications. These upgrades included: Replacement of motors on valve actuators in the Steam Generator Blowdown Treatment System (DCR 1730)~, Replacement of limit switch compartment covers and motors on valve actuators in the Steam Generator Blowdown Treatment System, Residual Heat Removal System and Chemical and Volume Control System (DCR 1573, Rev. 1),
- Replacement of reactor coolant wide range RTDs which had reached the .
end of their qualified life. Also, splicing of all reactor coolant RTD wiring that had previously used terminal blocks and is subject to LOCA or HELB environmental conditions. (DCR 1521, Rev. 1) Splicing of wiring for environmentally qualified instruments that had previously used terminal blocks and is subject to LOCA or HELB envir . onmental conditions. (OCR 1802) , Summary of Safety Evaluation In each case the modification resulted in a higher degree of com-nonent reliability during post-accident operation; plant safety was improved. Fire Protection Significant work was completed on many of the modifications required by 10CFR50, Appendix R, Fire Protection Program, which includes the following:
- The dedicated shutdown panel has been partially put into service with approximately 80 components operational from the panel.
- The installation of required instrumentation on the dedicated shut-down panel is 95% complete.
- The three hour fire wall installation and penetration sealing is 98%
complete throughout the plant.
- Cable pulling required for separation of dedicated and alternate j shutdown equipment is complete.
- Fire detection system modifications are complete.
HVAC system modifications are 98% complete. (DCRs 1189, 1191, 1192, 1193, 1194, 1195, 1197, 1361)
- 3. 3 -.
s Summary of Safety Evaluation l
) These modifications enhance both automatic and manual control of the j plant in the'unlikely event of a fire. The Appendix R modifications-will. preclude a fire from affecting the capability to bring the plant l to a safe shutdown, i g The Kewaunee Nuclear Power Plant was granted an extension for completion
~ of a11' Appendix R modifications until the 1987' refueling outage.
-Pipe Thinning Portions of the carbon steel piping in the Heater and Moisture Separator j , Drain System, Bleed Steam System, and Turbine Room Traps and Drains System i were replaced with stainless steel piping. This change was required due ~!
to wall thinning, a result of water impingement erosion. (DCRs 1781, 1782,'and 1783). E Summary of Safety Evaluation Stainless steel is'less susceptible to moisture-related erosion than
- carbon steel. Installation of stainless steel piping, where moist.
- steam is transported, will increase plant reliability by reducing the 2
possibility of an outage due to secondary side unavailability. ,. Reactor Coolant Permanent vibration monitors were installed on both Reactor-Coolant Pumps. These monitors are equipped with remote readout and analytical capabili-ties in the control room. They provide continuous surveillance of Reactor Coolant Pump condition and immediate indication of trouble to control room operators. It also provides a remote means of checking the balance of the pumps. This is part of the Reactor Coolant System. (DCR 1218) Summary of Safety-Evaluation This modification will enhance plant safety as the vibration monitors will provide plant operators with early detection of Reactor Coolant Pump problems. They will provide information that can be used to mitigate the consequences of a pump failure. They will also help with ALARA because pump balance can be checked remotely. Miscellaneous Instrumentation I' A loose parts monitoring system (LPMS) was installed to monitor the reac-tor vessel and both steam generators. The sensors are located such that they monitor the natural collection points of the Reactor Coolant System. The LPMS panel is located in the relay room and tied into the plant annun-ciator system. The system's only function is to alert the operators of a
- loose parts condition and to record data for future evaluation. (DCR 1525) i L
. -. -. . -. ,~. - -
,'t.. .
- 3.4 s
Summary of Safety Evaluation . f-), -This system enhances th'e reliability of the RCS by providing a means
.of detecting: internal loose parts. Awareness of loose parts will
-afford operators the' opportunity to take appropriate action to avoid '
- damage that could result. Miscellaneous Instrumentation
- The Seismic monitoring system at KNPP was replaced because it was obsolete,
- .which caused problems in maintaining the systems. (DCR1085)
Summary of Safety Evaluation The new seismic monitoring system is more reliable and maintainable. It also provides faster.results for determining plant response. 4 4 4160 V Supply and Distribution The 4160 V switchgear X-Y relay circuits were modified to prevent inadver-4 tent closing of the breaker due to a ground in the permissive. circuits. An interposing relay has been installed in the permissive circuit of each breaker in place of the triac gate'of the X-Y unit. A.normally.open con-- If a. ground in the per-tact.off'of this relay is used to fire the triac. missive circuit should occur, the capacitive discharge, which was enough
- f. - .
to fire the'triac and close the breaker, would not be enough to' pick up the interposing relay. (DCR 1688) Summary of Safety Evaluation L This change does not affect the operation of the 4160 V Switchgear other than to prevent the malfunction described above. This change increases personnel safety by. preventing inadvertent starts of 4160 V' equipment. It does not create a change in the USAR or Technical , Specifications nor does it increase the probability of an accident.
~
Chemical and Volume Control The letdown isolation orifice stop valves (LD-4A, B, & C) were relocated to i a lower radiation area that is more physically accessible for valve main-tenance. (DCR 1561) . t t , Summary of Safety Evaluation The new location will facilitate valve maintenance, thereby
- - increasing system integrity and reliability. The function and opera-tion of the valves remains the same.
1 o m--- -,_-,__.,__.....___,_,....__ _,-_______
3.5 Reactor Coolant-N V Eight control rod guide tube support pins (split pins) were replaced under the Plutonium recycle ports in the reactor vessel internals. The flow restrictors on the four Plutonium recycle port guide tubes were replaced with flexureless flow restrictors. (DCRs 1843 and 1844) Summary of Safety Evaluation The replacement parts are less susceptible to mechanical failure and, therefore, reduce the possibility of introducing loose parts into the Reactor Coolant System. Steam Generator Blowdown Treatment The SGBT heat exchanger outlet temperature alarm setpoint was raised from .. 110 F to 135 F, and the temperature trip setpoint was raised from 115 F to - 145*F. The changes were necessitated by summer lake temperatures. Lake temperature increases to > 65 F cause condensate temperature increases to
.> 100*F. With these conditions, normal operation of a 10*F difference between condensate inlet temperature and blowdown outlet temperature place the system in an alarmed condition and close to the trip setpoint.
(DCR 1758) Summary of Safety Evaluation ~
~
This change is being made to accommodate plant operation during s summer months, when condensate temperature may rise above 100 F, making it impossible to cool steam generator blowdown below 100*F with condensate water. There are no safety consequences associated with this revision. Steam generator blowdown presently is discharged to Lake Michigan with the circulating water discharge. Blowdown flow, at approximately 80 gpm, is negligible compared to circulating water discharge flow, approximately 400,000 gpm. Radiation Monitoring The high alarm setpoint of Radiation Monitor R-20 (Service Water System Monitor) was lowered from 3.2 x 108 cpm to 3.0 x 108 cpm. (DCR 1839) Summary of Safety Evaluation This DCR did not require a change to Technical Specifications but was the result of Amendment 64 to the Technical Specifications. Amendment 64 required the setpoint to be determined in accordance with the Offsite Dose Calculation Manual (00CM). The calculated set-point, per the 00CM, Table A is 3.0 x 10' cpm. This change is in the conservative direction. ! Internal Containment Spray (ICS) The set pressure for the ICS pump suction relief valves was raised from 150 psi to 225 psi to facilitate hydrostatic testing. (DCR 1723)
3.6 Summary of Safety Evaluation O
=d The ICS pi;ing and pump design conditions were verified to be ade-quate for the pressures that could be experienced due to the increased setpoint for the ICS pump suction relief valves.
Reactor Building Ventilation The Containment. Clean-up System was removed to gain additional floor space in Containment. Its function was to clean up Containment prior to con-tainment entries and prior to a plant shutdown. Should this need arise in the future, the Reactor Building Ventilation System still has this clean-up cabability through filtration and dilution. (DCR 1702) Summary of Safety Evaluation Removing the Containment Clean-up System will eliminate the charcoal filter fire hazard from containment, thus increasing safety. The clean-up system is not connected mechanically or electrically to the Reactor Building Ventilation System and is only operated manually from the control room. Removing the system will not affect any other plant safeguard systems. The function of the Containment Clean-up System can be performed by other equipment in the Reactor Building Ventilation System. Miscellaneous Drains and Sumps The floor drains between the two diesel rooms were unplugged to allow drainage to the turbine building sump. Check valves were installed in the floor drains of each of the diesel rooms to prevent backflow into the rooms in the event of a fluid break. (DCR-1388) Summary of Safety Evaluation Installation of check valves will provide fire separation of the 4" diameter pipe between the two diesel rooms while allowing the floor drains to be drained to the turbine building sump. O
4.1 4.0 LICENSEE EVENT REPORTS ( V This section is a summary of the 16 Licensee Event Reports (LER) submitted to the NRC in 1986 in accordance with the requirements of 10CFR50.73. None of the events described in the 1986 LERs posed a threat to the health and safety of the public. LER 84-006-01 On April 16, 1984, with the plant in the refueling mode, testing was in progress to determine containment isolation valve leakage rates in accor-dance with 10 CFR 50, Appendix J. Both the inboard and the outboard iso-lation valves for the Chemical and Volume Control System Letdown line were found to have leakage rates greater'than the upper measuring limit of the Local Leak Rate Test (LLRT) measuring device. Corrective actions taken were: 1) replacement of the Seal Ring gaskets in two of the three parallel inboard valves and 2) adjustment of the stroke on the outboard valve. The post-repair LLRT values were within tne limits of the Kewaunee Nuclear Power Plant Technical Specification (TS) 4.4.b. During the 1985 and 1986 refueling outages, leak rate reduction repairs were performed on the inboard valves with the assistance of the valve ven-dor representative. No leakage in excess of TS 4.4.b were encountered for the Letdown line. A design change was implemented during the 1986 refueling outage to relocate the inboard isolation valves to provide improved workspace and a reduction of personnel exposure. Following the valve repair and relocation, no degradation of valve performance was Os noted. No further action was considered necessary for either the inboard or the outboard isolation valves in the Letdown line. This event was reported pursuant to 10 CFR 50.73(a)(2)(ii). LER 86-001 On February 11, 1986, with the plant at full power operation, a review to determine retest requirements following a solenoid valve replacement on the four 36-inch containment purge and vent valves revealed that a valve timing test had not been performed during the 1985 refueling outage. Failure to perform this test is contrary to Technical Specification Table 4.1-3, Item No. 16, which requires that a timing test be performed each refueling cycle. These valves are used to vent containment at plant shutdown; ouring power operations the valves' control switches are sealed shut and a two-inch mini purge line is used. During 1979 through 1984 the timing test was included in the Pump and Valve Inservice Testing Plan. Following the 1984 refueling outage, when the two-inch lir.e was installed, the timing test requirements were erroneously deleted from testing procedures. Failure to perform this valve timing test is considered reportable per 10 CFR 50.73(a)(2)(1)(B) as operation prohibited by Technical Specifications. Corrective actions were taken to add the valve timing requirements back into a procedure and to review and document the func-g tion of valves covered by the IST Plan to prevent a recurrence of this type of event.
y 4.2 L On March'1, 1986, a valve. timing test was performed. All four valves-
. operated within'.the required time limit; hence,-there was no concern that' f; they would be able to satisfy their-safety-function.
n U 'LER 86-002. On March l18,~1986, with~the plant in refueling shutdown, a review of'1986
. Containment Local. Leak Rate Test (LLRT) results determined that the cumula--
.tive leakage allocated to= areas beyond the Auxiliary Building Special
' Ventilation Zone (7.772 Std. Litres per Minute (SLM)) had exceeded the limit-specified in Technical Specification (TS) 4.4.b.6 (0.01 La or approximately 5.380 SLM).. The three-penetrations accounting for the majority.of the leakage were for Station ' Air, Service Water, and a Containment pressure transmitter. Rust particles were cleaned from the internals of the four' station air isolation valves. The valves were-lapped, repacked, and satisfactorily retested. A Service Water check valve, SW6011, was cleaned of internal deposits,. lapped, and~satisfac-. ,
torily retested. Loose fittings were found, retaped, and tightened on the l Containment pressure transmitter with-a satisfactory retest. .There were l no additional actions planned regarding these three penetrations aside from the normal performance of periodic LLRTs. LER 86-003
- , On March 26, 1986, with the annual refueling and maintenance outage in v progress, management concluded that there was a potential problem with the L long-term reliability of both emergency AC Diesel Generators (DG). On March 13, 1986, with the 1A DG disassembled for design change' work, the No. 1 idler gear floating bushing was found with small crack indications O and frozen to the stub shaft. Subsequently, on March 25, during the disassembly of the IB DG for implementation of the same design change, the L. No. 1 idler gear floating bushing was found with a through wall crack, i Cause of this deterioration was not conclusively identified; however, it i4 is believed to have been caused by numerous fast starts of the DGs without full main lube oil pressure. A new design stub shaft bracket, which does not require a floating bushing, was installed on both DGs. The new design l will supply filtered lubrication to the gear from the soak back pump which is in continuous operation.
This event was conservatively reported per 50.73(a)(2)(v) as an event or condition that alone could have prevented the fulfillment of a safety function. Both DGs had been demonstrated operable during previous routine survei_11ance testing, but it was concluded that their extended operation without corrective action may have resulted in gear failure. LER 86-004 On April 1, 1986, with the annual refueling / maintenance outage in progress, plant management was concerned with the operability of non-spring assisted fire dampers at the Kewaunee Plant. On the morning of April 1, functional tests were performed on two ventilation fire dampers to ensure that they would automatically close with maximum expected air flow. This testing was performed in response to information provided by INPO and the NRC on fire damper problems. In both tests, the dampers became mechanically bound in an intermediate position; securing the operating ventilation system did not result in further damper closure. The subject dampers provide isolation of fire zones with safety related equipment.
4.3 The root cause of the failure was use of the dampers in an application for
;(s ' which they were not tested; industry standards do not require that fire -L- dampers be functionally tested to demonstrate performance under air flow l conditions. This event did not meet the reporting criteria contained in l 10 CFR 50.73(a); however, the test results were believed to be of generic interest.
I As interim action, an hourly fire watch was established. Testing of the fire dampers continued, and a design change was initiated for the replace-ment of those fire dampers found to be inoperable. LER 86-005 At 0855 on April 17, 1986, with the plant in Hot Shutdown preparing for startup following the annual refueling outage, a reactor trip occurred during the performance of a work request on intermediate range nuclear instrumentation channel N35. The trip occurred as a result of a break-down in communication between plant personnel. An I & C man misin-terpreted instructions given by the Shift Supervisor concerning the . troubleshooting he was to perform on N35 to eliminate chattering of the P-6 bistable. As a result, the man removed the control power fuses for the channel prematurely. The loss of control power completed the one-out-of-two actuation logic for the Intermediate Range Hi Flux Reactor Trip. The Operators performed the immediate actions prescribed in the Reactor Trip Response Procedure. To mitigate the possibility of a similar occurrence, the Shift Supervisor notified the I & C Supervisor of the y_ situation, stressing the importance of maintaining proper communication between the I & C group and control room personnel while safety-related work is being performed. The reactor was in the shutdown condition prior to the event, and the reactor protection system performed as designed; hence, there was no impact on public health and safety. LER 86-006 On April 20, 1986, during startup after refueling, operators were increasing power to 25% of full power for an eight-hour hold prior to per-forming the turbine overspeed and torsional tests. At 1038, hi-hi water level in' steam generator 18 cause a turbine trip and subsequent reactor trip. The cause of the trip was the 18 Main Feedwater Control Valve sticking, then suddenly opening when the control room operator increased valve demand to transfer feedwater control from the IB Bypass Control Valve. Actions to control water level in steam generator 1B were unsuccessful, and the turbine tripped due to hi-hi water level in steam generator 18. Prior to restarting the reactor at 1250, the 1B Main Feedwater Control Valve was inspected, adjusted and cycled to verify operability. O
- 4.4 LER 86-007 I
-(] On April 24, 1986, with the plant at 63% power, I & C personnel in conjuc-U tion with the plant operators were performing a surveillance procedure,
" Steam Generator Flow Mismatch Instrument Channel Test". At 1630, a reactor / turbine trip occurred due to a Lo (25.5%) steam generator level coincident with steam flow greater than feedwater flow on 18 Steam ,
Generator. l l The cause of the trip was attributed to the 18 Main Feedwater Control l Valve moving from 40% open to 20% open when the controlling channel for i the steam generator level control was switched from the yellow to the blue channel. The operator immediately switched the steam generator level control system to the manual mode and attempted to further open the valve. The feedwater control valve did not immediately respond, which resulted in a decreasing steam generator level and subsequent reactor trip. Prior to restarting the reactor at 1942, the Auto / Manual Station for IB S/G Level Control was replaced and the IB Main Feedwater Control Valve was cycled to verify operability. LER 86-008 1 At 1030 on May 13, 1986, with the plant at full power, a reactor / turbine trip occurred during the performance of a surveillance procedure on the power range nuclear instrumentation. The initiating trip signal was a simulated two out of four coincidence of overpower delta temperature. rr The trip occurred as a result of a personnel error. The protection signal bistables were tripped on Channel N44 and the test signal simulating an increased power. level was input to Channel N43. All plant systems responded normally, and the plant was stabilized in a hot shutdown mode. There was no effect on public health and safety. Corrective actions included discussions with I & C personnel, procedure evaluation, and better labeling of the instrumentation in the Control Room. LER 86-009 On May 13, 1986, an engineering review concluded that the Local Leak Rate Test (LLRT) procedure for the Station Air to containment penetration 19 did not separately quantify the inboard and outboard isolation valve Type-C leakages. An historical review of Type-C tests for penetration 19, including the pre-op tests, indicated the test alignments were based upon a subtle error in the station air flow diagram. Apparently the original drafting of the station air flow diagram repre-sented the containment isolation valves and the Type-C test alignments for penetration 19 incorrectly as either gate or globe valves, which allow flow in the reverse direction, r&ther than as globe stop check valves. This event was reported under 10CFR50.73(a)(2)(i) as a Tech. Spec. viola-tion. O O
, . , - - - , - , _ , - . _ _ _ . , _ - . ,,.-_c-- , - - . , , - - - - _ , , _- - - - . , - - , , - ,
4.5 Immediate corrective actions included revising plant drawings, reviewing the remaining Appendix J reportable penetrations for similar con-
~(] figurations, drafting a revision to the LLRT procedure for use during the 'v 1987 refueling outage, and drafting a special test procedure to allow testing the isolation valves of penetration 19 while at power. The spe-cial test procedure was approved by PORC.on June 9, 1986, and performed June 13, 1986. The results indicated that the leakage through penetration 19's isolation valves is within Appendix J and KNPP Technical Specific'tions a acceptance criteria.
The integrity of penetration 19 has been repeatedly demonstrated through its ability to hold pressure during pre-op tests and periodic Type-C and Type-A tests; therefore containment integrity was not compromised and there was no impact on the health and safety of the public. LER 86-010 At 0847 on August 1, 1986, with the plant at 100% power, I & C personnel had just completed installing a jumper to bypass the Hi Radiation steam Generator Blowdown isolation trip function per plant surveillance proce-dure when the jumper disconnected and shorted to ground. This caused the circuit breaker for Circuit 11 to open. The engineered safety feature equipment associated with Circuit 11 then assumed their safeguard con-figuration. The I & C man immediately informed the Control Room Operators of the cause of the incident. The breaker was reset, and the equipment was returned to its normal state. The type of jumper used in this case was an alligator clip connection. Due to the low profile design of the connecting screws on the terminal strip, the alligator clip slipped off. To prevent a recurrence of this event, adding jacks where jumpers are used is being evaluated during monthly surveillance tests. All Engineered Safety Feature Equipment affected by this event assumed their safeguard configuration as required, and there was no impact on public health and safety. LER 86-011 On August 11, 1986, at 0230, with the plant at 100% power, both fire pumps were simultaneously out of service for approximately ten seconds. This event occurred during the' monthly fire pump flow test, when plant personnel performing the procedure opened the 18 pump breaker before the 1A pump breaker was closed. The Control Room Operator, immediately recognizing the situation after receiving annunciators, initiated a call to the plant personnel involved. However, before the call was completed, the 1A pump breaker was closed, returning the pump to service. Corrective actions included: 1) counseling the plant personnel;
- 2) revising the Administrative Control Directive detailing the operations group organization to designate the Shift Supervisor as responsible for the assignment of independent verification; 3) revising the Administrative Control Directive detailing the Independent Verification Program to more
- - clearly identify requirements and responsibilities and presenting it to each operating crew for discussion at a monthly safety meeting; 4) pre-l
4.6 senting, as part of-the Equipment Operator/ Auxilary Operator training accreditation program, a specific lesson on Technical Specifications Q V-during initial training and emphasizing system-related Technical Specifications during requalification training; and 5) including this event in a quality circle group's study of personnel errors and developing additional corrective actions as necessary. This event was reported pursuant to Technical Specification 3.15.b.4 on the Fire Water System. LER 86-012 At 1810,'on September 26, 1986, with the plant at full power, an Unusual Event (UE) was declared based on elevated Auxiliary Building Vent Stack Radiation Monitor indications. The radiation was due to radioactive Argon-41 gas. The Argon-41 was attributed to argon, instead of hydrogen, being mistakenly supplied by the vendor and used as a cover gas on the Volume Control Tank. The higher than normal radiation levels resulted in actuation of some of the plant's ESF equipment. From 1810 on September 26, until 1200 on September 27, access to the Auxiliary Building was limited, which prevented the performance of fire watches. Immediate corrective actions were taken to terminate the release and ven-tilate the Auxiliary Building. Based on decreasing radiation levels, the UE was terminated at 0333 on September 27. Long-term corrective actions included follow-up with the gas bottle supplier, identifying leaks in the waste gas system and revision to plant procedures. s pd At 1037 on October 6, Train A of the Zone Special Ventilation System auto-matica11y started on a high radiation signal. The start was due to 1) the actuation setpoint of the Auxiliary Building Vent Stack Radiation Monitor being set a decade low, and 2) Argon-41, which was still present in the Chemical Volume and Control System, being released when venting the 1A
- Reactor Coolant Pump seal water injection filter after filter replacement.
Corrective actions were taken to revise the radiation monitor test proce-l dure. The Technical Specification limits for RCS activity and the dose rate at ! the site boundary were not exceeded during either event, j LER 86-013 At 2128 on October 30, 1986, a security officer discovered the gate to the Waste Metering Tank Room, which was posted as a high contamination and radiation hazard area, unlocked and in the open position. Immediate action was taken to notify the on duty Radiological Protection Technician, Shift Supervisor, and Security Shift Captain. A subsequent survey of the room indicated less than 100mR/hr general area and 2R/hr on contact at the bottom of the tank. With only one small spot on the tank exceeding 1R/hr and the fact that its location is relatively inaccessible, it is doubtful that a major portion of the body could have been exposed to more than 100mR in one hour. Following the survey, the gate was secured at 2140. O
, 4.7 ;
l The root'cause of the event was personnel error. At approximately 0915 on October 30, two Radiological Protection Technicians entered the room to 1s'= perform the weekly radiation survey. Since further investigation indi-cated that no other authorized entries were made that day, and the gate
- was locked prior to the survey, it was concluded that the gate was-left unsecured and unattended from approximately 0915 to 2140 on October 30.
To prevent recurrence of this event, the Health Physics procedure for the
!; weekly surveys of the Auxiliary Building was revised to require verifica-tion of door or gate position upon completion of each area survey. In addition, this LER was included in the required reading list for the U Radiological Protection Technicians and presented to all personnel at a
- plant safety meeting.
This event was reported as required by 10 CFR 50.73(a)(2)(1)(B) as a con-dition prohibited by plant Technical Specifications. , LER 86-014 On December 2, 1986, with the plant ct 100% power, both Boric Acid Transfer Pumps were out of service for boric acid injection for approxima- , tely'four minutes. Technical Specification 3.2.b.2 requires both Boric Acid Transfer Pumps to be operable for boric acid injection during power - operation. The event occurred during performance of a normal operating procedure to h align the Boric Acid Storage Tanks for recirculation to obtain a represen-tative sample of each tank. As required by procedure, the 1A Boric Acid Transfer pump was removed from service to recirculate the contents of the Boric Acid Tank. In this recirculation mode, the pump was not available as a source of boric acid injection as required by technical specifica-tions. Both pumps were then started per the procedure with the 1B pump failing to start. Upon pump failure, immediate action was taken by plant operators to realign the 1A Boric Acid Transfer Pump for boric acid injec-tion, and at 0806, the boric acid injection path via the charging pumps
- was restored. Following co.tective maintenance, the IB Boric Acid- ,
! Transfer Pump was returned to service at 2140. ! The root cause of the pump failure was a burned off section of heat tape f around +%e pump casing allowing boric acid to crystallize. The insulation 4 on the pump was replaced and extended to cover the mechanical seal in addition to the pump casing. The heat tracing on the pump casing was ! replaced, the pump was rebuilt, and additional temperature sensors have been installed on the pump casing and seal housing to improve monitoring '~ capabilitles. This event was reported as required by 10 CFR 50.73(a)(2)(1) as a viola-tion of plant Technical Specifications. . C lO
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4.8 A LER 86-015'~ At-1115 on December 29,'1486, an I&C man discovered a service water inlet valve for Radiation" k nitor R-16 closed. With the1 valve closed, R-16, the
.-liquid effluent monitor on the' Service Water return line'for the Contain-
-ment Fan Coil Units, was unavailable to monitor the service water discharge from Containment Fan Coil' Units 1A and 18. Immediate actions were taken by the I&C man to notify control room personnel of the event and return the valve to its'normally open position.
The root cause of the event was personnel failure to follow procedure. On
-December 28, the supply lines for R-16 were flushed. During the flushing process, the service water inlet valve is closed, administratively remov-ing R-16 from service. At approximately 1300, following the flushing of the supply lines for R-16,'the plant personnel performing the procedure-failed to. reopen the service water inlet valve per procedure. Due to a faulty flow indicator, which erroneously indicated flow to R-16, the closed position.of the valve went undetected.
To prevent recurrence of the event, the flow meter was cleaned and returned to service, the operating procedure used to perform the flushing process was reviewed to determine if clarification was needed, and the current. system design will be evaluated to determine if modifications could improve the conditions for procedure performance. With R-16 in a degraded condition, Radiological Effluent Technical Specifications (RETS) allows effluent releases.to continue, provided that
-grab samples are taken at least once per 12 hours. As a' result, this event is.being' reported as a special report as a failure to satisfy a Radiological Effluent Technical Specification action statement, s
O
5.1 5.0 FUEL INSPECTION REPORT .n. ' &. Thirty-two (32) fresh Region N assemblies were loaded for Cycle XII. Startup physics testing was performed and reported in the Cycle XII Startup Report. The irradiated fuel inspection was performed with an underwater TV camera. All peripheral fuel rods were examined using one-half face scans. Seven assemblies were inspected, including one each in regions I, J and K and two each in regions L and M. All assemblies exhibited rod slippage to various degrees, with the majority having rods in contact with the bottom nozzle. Numerous scrapes to the rodlets, grids, and top and bottom noz-zles were also noted. However, no damage to the cladding or supporting structures was observed. All assemblies exhibited axially varying crud deposit . Overall condition of the fuel was very good, with no evidence of fuel cladding degradation on the fuel rods examined. Video. tapes were made of all examinations. O
.. . - . . . . _ - - . ._ _.m . . _ . _ _ _ - . - _ _ _ _ . _ _.. ....._._. _. ._. . ._ _ __ . _ --- _ - . . _ _ _
6.1 l: - 6.0 CHALLENGES TO AND FAILURES OF PRESSURIZER SAFETY AND RELIEF VALVES
.O F There were no challenges to or failures of pressurizer safety or relief -
. valves during 1986. e s W i b i h e f l 4 !O . 1 } l ( i i i O i i
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7.1 7.0 SUW4ARY OF 1986 STEAM GENERATOR ED0Y CURRENT EXAMINATION During the Kewaunee Nuclear Power Plant refueling outage of 1986
" (March - April), the following steam generator services were performed:
l l Plug Removal Prior to steam generator tube eddy current inspection, plugs were removed from both ends of the following tubes in steam generator A: Row 43, Column 32 Row 43, Column 33 Row 44, Column 33 These plugs were removed to permit the inspection of the tubes, installation of stabilizers, and to replace sentinel plugs with mecha-nical plugs. Eddy Current Inspection The inspection program consisted of 100 percent of all accessible tubes in both steam generators. All tubes in rows 4 through 46 were i inspected full length. All tubes in rows 1 through 3 were inspected from the inlet through the U-bend. All testing was performed using a 0.720-inch diameter probe (0.680-inch O low row U-bends) using multi-frequency techniques in both the differen-tial and absolute modes. Mechanical Plugging Steam Generator A Mechanical plugs were installed in both ends of 26 tubes of Steam Gene'rator A on the basis of eddy current indications. One tube, R24C36, was plugged on the inlet side; it was previously plugged in only the outlet side. Steam Generator B Mechanical plugs were installed in both ends of 46 tubes of Steam Generator B on the basis of eddy current indications. One tube, R24C30, was plugged on the inlet side; it was previously plugged in only the outlet side. Three plugs were installed in the outlet side to replace sentinel plugs previously removed earlier in the same outage. O
J 7.2 Stabilizer Installation
-(o.
+
Three stabilizer assemblies were installed in the inlet side of Steam Generator A, in the following tubes: Row 43, Column 32 Row 43, Column 33 Row 44, Column 33 APPLICABLE DEFINITIONS: Degraded Tube: A tube with greater than a 20% thru-wall indication. Defective Tube: A tube with greater than a 50% thru-wall indication. Defective tubes require plugging. HISTORICAL SLM4ARY OF TUBES PLUGGED IN THE KEWAUNEE STEAM GENERATORS Number of Tubes Plugged in: Steam Generator IA Steam Generator 1B 1983 23 50 1984 8 17 O 1985 1986 27 26 22 46 TOTE T W TOTAL AS PERCENT (3388 tubes / generator) 2.5% 4.0% I O
7.3 Table 7.1-
. Summary of 1986 Steam Generator Eddy Current Examination Inspection Extent STEAM GENERATOR 1A Extent of Inspection Hot leg Cold Leg Full Length 3,068 265 U-Bend
#5 TSP (1) 1 1
#2 TTS TSP (2) i TOTALS 3,334 2 STEAM GENERATOR 1B Extent of Inspection Hot leg Cold Leg Full Length 3,020
,O i U-Bend
#1 TSP
' 280 1
5 TTS TOTALS 3,300 6 (1) Tube support plates, counted up from the tube sheet inspected. (2) Top of tube sheet. l0
7.4 TABLE 7.2 (Page 1 of 3) y_ 1A Steam Generator
,. ) 1986 Eddy Current Examination i
% THRU-WALL INDICATIONU)
R0W COLUMN PENETRATION PLUGGED LOCATION 1 1 50' 'X H: TH 5 2 23 H: TSP #1 13 4 36 H: TSP #1 8 9 87 X H: TE + 8.4 18 10 55 X H: TSP #1 24 10 24 H: TSP #4 + 1.4 s 10 11 86 X H: TE + 2.4 1 14 50 X H: TH 18 14 85 X H: TE + 10.1 14 16 35 H: TSP #1 U 33 16 21 H: TSP #1 8 19 29 H: TS + 0.9 25 20 71 X H: TE + 6.3 11 23 24 H: TSP #1 2 24 30 H: TSP #3 9 24 93 X H: TE + 2.9 24 26 31 H: TS + 0.5 34 26 23 H: V3 25 27 29 H: TS + 0.6 23 30 22 H: TS + 3.6 3 32 28 H: TS + 0.7
, 28 33 22 H: TS + 1.8 32 33 UDS(2) X H:
35 33 28 H: TS + 0.3 2 34 82 X H: TE + 6.2
- - _ . _ _ _ _ .., . .._.,._,m.__-,, ...,,_..._-.-.--__._y_ . ~ . , , ,,y __ -. .__-,_.---_,..m,.. ~ , _ , . , . , _ _ _ - , , _ .-y .. c_m._
a 7.5
~
TABLE 7.2 (Page 2 of 3) p 1A Steam Generator
-Q 1986 Eddy Current Examination
% THRU-WALL INDICATION (U R0W COLUMN PENETRATION PLUGGED LOCATION 35 34 84 X H: TE + 4.3 17 35 24 H: TS + 2.4 24 36 33 X(3) H: TS + 0.5 34 36 81 X H: TE + 7.7 40 36 26 H: V3 4 39 80 X 'H: TE + 4.1 29 39 30 H: TS + 0.7 30 43 25 H: TS + 0.4 18 44 89 X H: TE + 7.7 1 45 65 X H: TC 11 45 DI(4)+ X H: TE + 2.9 28 46 57 X H: TS + 0.5 30 46 34 H: TS + 0.4 32 46 39 H: TS + 0.4 22 47 28 H: TS + 0.3 7 48 DI(4) X H: TS + 0.4 24 48 27 H: TS + 0.8 26 48 34 H: TS + 0.3 9 49 22 H: TS + 0.7 24 49 29 H: TS + 2.4 I 22 50 37 H: TS + 1.8 23 51 25 H: TS + 1.7 18 52 85 X H: TE + 5.3 29 52 31 H: TS + 0.6 11 53 22 H: TS + 0.9 2 54 89 X H: TE + 2.4
a.-- , 7.6 l .
.-)
TABLE 7.2(Page3of3) $ n- 1A StvawGenerator - () . 1986 Eddy Current Examination --
~
~-
% ThRU 4Ai.L INDICATIONU)
ROW COLUMN PENETRATION PLUGGE0 LOCATION 7 55 88 X H: TE+4J 4 10 56 25 , > H: T) + 0:7 j i 2 58 78 X H: TE +-2.4 - 11 59 32 H: TS +.0.n ' M 13 63 92 X H: TS + 1.5 13 64 39 H: TS + 1.0
.i.
23 65 35 ,, H: TS + 0.9 23 66 23 , H: TS + 0.8 1 8 68 29 - H: TS + 0.6 31 68 76 X H: TE + 4.5
) 16 70 .
51 X H: TE +11.! 19 70 24 H: TS + 0.5, 10 77 26 H: TS + 0.8 19 78 81 X H: TE + 5.6 6 80 24 H: TS + 0.6 , 31 16 27 - C: TSP #1 C: TSP #7 2 77 29 (1) H - Inspection frem Hot Leg C - Inspection from Cold Leg TSP - Tube Support Plate , TS - Tube Sheet TE - Tube End TC - Tangent Point Cold Leg Side TH - Tangent Point Hot Leg Side V3 - #3 Antivibration Bar Note that numbers added to TSP, TE, etc., are distances in inches above the indicated landmark, in the indicated leg. (2) UDS - Undefineable Signal, WPSC decision to plug. (3) Tube R24 C36 was plugged in the cold lag only in 1986. 4 (4) DI - Distorted Indication.
~ . .
F7 g. }[t ), 7.7
/ ,
V
,, 4- TABLE 7.3 (Page 1 of 10) 18 Steam Generator 1986 Eddy Current Examination t
~~~
*4 % THRU-WALL INDICATION W R0W COLUMN PENETRATION PLUGGED LOCATION
[. g 20 6 21 H: V3
'4 30 12 ) 25 H: TSP #5 1
j l / l 5 17 26 H: TS + 0.9
~ /' . 13 17 26 H: TS + 0.6 11 18 24 H: TE + 4.2 '
I f6 19 30 H: TS + 0.2 9 19 33 H: TS + 20.7 10 s 19 54 X H: TS + 10.2 4, 20 42 H: TS + 0.4 5 20 - 37 H: TS + 0.3 11 20 61 X H: TE + 3.8 6 21 24 H: TS + 14.8 il 22 46 H: TS + 0.7 5 22 41 H: TE + 7.2 t f-9 22 30 H: TS + 0.3 3 23 24 H: TS + 0.4 cf, 9 23 33 H: TS + 1.0 r j, 6 24 43 H: TS + 0.4
- 11 24 85 X H: TE + 3.2
'i 11 25 33 H: TS + 0.6 16 25 28 H: TS + 2.1
<. 24 25 26 H: TS + 0.7 25 25 48 X H: TS + 0.6 4 26 32 H: TE + 3.0 16 26 SQR(2) X H: TE + 2.8
/
7.8 3 TABLE 7.3 (Page 2 of 10)
- t. 18' Steam Generator O ,
W 1986 Eddy Current Examination
% THRU-WALL INDICATIONU'l R0W COLUMN PENETRATION PLUGGED LOCATION 5
27 .( 38 g H: TE + 4.2 7 27 31- H: TE + 3.1 12 28 , 29 H: TS + 0.4 15 28 30 H: TE + 10.7 16 28 40' X H: TE + 3.6 23 28 44 H: TE + 3.9 24 28 , 33 H: TE + 3.6 25 28 'y 23 H: TE + 4.1
. 27 28 i 29 H: TS + 1.2 12 29 33 H: TS + 0.8 0.-
r 16 29 38 H: TS + 1.2 26 29 35 H: TS + 0.9 L, t ; 24 30 27(3) X H: -TE + 3.8 27 30 21 H: TS + 1.1 28 31 45 H: TS + 0.9 8 32 48 X H: TE + 4.3 24 32 27 H: TE + 3.5 27 32 30 H: TS + 1.2 l l 28 32 28 H: TS + 1.2 5 33 35 H: TE + 3.9 6 33 55 X H: TE + 3.7
.8 o 33 49 X H: TE + 3.5 ~
11 34 35 H: TE + 3.6 8 36 SQR(2) X H: TE + 8.0 16 36 26 H: TS + 2.3
7.9 TABLE 7.3 (Page 3 of 10) IB Steam Generator 1986 Eddy Current Examination
% THRU-WALL INDICATIONU)
R0W COLUMN PENETRATION PLUGGED LOCATION 30 36 32 H: TS + 1.9 31 36 34- H: TS + 0.3 38 36 24 H: TSP #4 3 37 66 X H: TE + 3.6 4 9 37- 23 H: TE + 3.5 24 37 SQR(2) X H: TE + 3.0 29 37 37 H: TE + 3.5 30 37 34 H: TS + 2.3 32 37 22 H: TS + 1.6 11 38 32 H: TS + 0.3 30 38 41 X H: TE + 8.0 31 38 26 H: TS + 2.1 29 39 SQR(2) X H: TE + 4.6 30 39 31 H: TS + 1.7 31 39 39 H: TS + 0.6 32 39 21 H: TE + 3.6 11 40 ?? H: TS + 0.8 15 40 47 X H: TS + 3.7 26 40 39 H: TE + 6.5 28 40 78 X H: TE + 4.4 1 41 93 X H: TE + 4.4 7 41 25 H: TE + 3.2 16 41 43 H: TS + 4.6 , 17 41 36 H: TS + 1.1 23 41 93 X H: TE + 2.9
7.10 TABLE 7.3 (Page 4 of 10)
& 18 Steam Generator Aj 1986 Eddy Current Examination
% THRU-WALL INDICATIONW1 RDW ' COLUMN PENETRATION PLUGGED LOCATION 25 41 49 X H: TE + 3.0 30- 41 42 H: TSP #1-31 41 40 H: TE + 10.5 32 41 45 H: TS + 1.2 1 42 90 X .H: TE + 3.4 6 42 25 H: TE + 3.5 8 42 27 H: TS + 40.5 15 42 89 X H: TE + 3.1 16 42 93 X .H: TE + 3.0 25 42 48 X H: TE + 6.7 ,
30 42 78 X H: TE + 3.9 2 43 44 X H: TE + 5.2 43 72 X H: TE + 2.9 10 43 25 H: TS + 0.5 16 43 34 H: TE + 4.5 32 43 24 H: TS + 0.8 43 43 38 H: TSP #5 4 44 78 X H: TE + 4.3 10 44 29 H: TS + 0.1 13 44 31 H: TS + 3.2 16 44 53 X H: TE + 4.0
- 25 44 25 H
- TE + 3.8 j 31 44 35 H: TE + 5.3 14 45 27 H: TS + 2.8 l 23 45 88 X H: TS + 6.4 l
7.11 TABLE 7.3 (Page 5 of 10) ^- IB Steam Generator (,) 1986 Eddy Current Examination
% THRU-WALL INDICATION (U RDW COLUMN PENETRATION PLUGGED LOCATION 24 45 87 X H: TS + 4.6 29 45 29 H: TS + 1.8 32 45 37 H: TS + 1.1 10 46 51 H: TS + 0.3 17 46 27 H: TS + 5.1 25 46 39 H: TE + 3.7 28 46 28 H: TS + 4.9 38 46 37 H: TSP #5 40 46 26 H: TSP #3 6 47 41 H: TS + 1.0 10 47 40 H: TS + 1.6 14 47 34 H: TS + 5.0 17 47 25 H: TS + 5.5 23 47 27 H: TS + 5.5 39 47 30 H: TSP #1 4 48 82 X H: TE + 3.7 18 48 44 H: TS + 3.3 19 48 24 H: TS + 4.8 20 48 41' H: TS + 4.6 24 48 45 H: TS + 3.7 30 48 66 H: TS + 1.3 !
31 48 41 H: TE + 2.8 33 48 38 H: TS + 0.8 j 11 49 25 H: TE + 4.7 13 49 33 H: TE + 4.4
7.12 TABLE 7.3 (Page 6 of 10) j ("'T 1B Steam Generator () 1986 Eddy Current Examination
% THRU-WALL INDICATION (i)
ROW COLUMN PENETRATION PLUGGED LOCATION 14 49 21 H: TE + 3.5 15 49 43 X H: TE + 3.3 17 49 43 H: TS + 4.6 29 49 37 H: TS + 3.4 30 49 56 X H: TS + 3.6 31 49 46 H: TS + 2.7 33 49 32 H: TS + 0.8 29 50 55 X H: TE + 3.8 31 50 35 H: TE + 3.1 3 51 87 X H: TE + 3.4 30 51 30 H: TSP #2 3 52 90 X H: TE + 3.1 32 52 33 H: TS + 1.4 4 53 46 X H: TE + 3.1 6 53 23 H: TE + 4.6 9 53 52 X H: TE + 3.4 10 53 66 X H: TE + 4.5 21 53 24 H: TE + 4.9 23 53 33 H: TE + 5.0 33 53 21 H: TS + 0.3 5 54 44 X H: TE + 3.8 29 54 22 H: TS + 7.3 33 54 39 H: TS + 0.2 6 55 22 H: TS + 0.5 32 55 44 H: TS + 6.1
7.13 TABLE 7.3 (Page 7 of 10)
~x 18 Steam Generator 1986 Eddy Current Examination
% THRU-WALL INDICATION (i)
R0W COLUMN PENETRATION PLUGGED LOCATION 26 56 26 H: TS + 4.7 29- 56. 24 H: TS + 6.9 33 56 30 H: TS 8 57 24 H: TE + 4.0 21 57 SQR(2) H: TE + 9.7 26 57 43 H: TE + 3.5 29 57 26 H: TS + 5.8 14 58 31 H: TS + 0.7 24 58 39 H: TE + 4.1 27 58 31 H: TE + 3.2 0 33 58 30 H: TS + 1.0 25 59 35 H: TE + 3.2 29 59 26 H: TSP #4 31 59 26 H: TS + 3.9 33 59 29 H: TS + 2.6 24 60 28 H: TE + 2.7 28 60 22 H: TS + 0.7 26 61 24 H: TE + 4.4 30 61 41 H: TS + 1.2 14 62 47 H: TS + 0.7 25 62 40 H: TE + 5.3 27 62 39 H: TS + 2.0 30 62 41 H: TSP #1 33 62 83 X H: TE + 8.0 39 62 41 H: TSP #6
7.14 TABLE 7.3 (Page 8 of 10) 18 Steam Generator () m. 1986 Eddy Current Examination
% THRU-WALL INDICATION (1)
RDW COLUMN PENETRATION PLUGGED LOCATION 22 63 32 H: TE + 3.6 23 63 26 H: -TS + 3.1 24 63 38 H: TE + 6.2 25 63 l 38 H: TS + 2.1 27 63 39 H: TS + 3.4 31 63 26 H: TSP #1 25 64 45 H: TS + 3.2 37 64 38 H: V4 11 65 47 H: TE + 2.9 17 65 27 H: TE + 3.2 0 26 65 27 H: TS 27 65 47 H: TS 9 66 35 H: TS + 0.6 10 68 40 H: TS + 0.8 6 69 40 H: TS + 0.4 17 69 24 H: TS + 2.5 25 69 29 H: TS + 0.9 i 11 70 49 H: TE + 2.8 21 70 22 H: TS + 1.8 24 70 30 H: TS + 1.0 33 71 25 H: TSP #3 35 71 26 H: TSP #3 l 12 72 33 H: TS + 2.5 26 72 78 X H: TE + 9.2 10 73 35 H: TS
7.15 i TABLE 7.3 (Page 9 of 10) IB Steam Generator 1986 Eddy Current Examination
% THRU-WALL- INDICATION (i)
ROW COLUMN PENETRATION PLUGGED LOCATION 13 73 25 H: TS + 1.4 33 73 '33 H: TSP #3 38 73 27 H: TSP #1 31 76 30 H: TSP #3 33 76 38 H: TSP #3 4 83 28 H: TSP + 35.1 19 90 30 H: TSP #4 9 2 32 C: TSP + 2.1 25 12 25 C: TSP #6 11 25 24 C: TSP #1
'O 41 41 28 C: TSP #6 43 41 23 C: TSP #6 43 42 41 C: TSP #7 21 43 21 C: TSP #5 43 43 45 C: TSP #7 25 44 25 C: TSP + 3.6 44 44 23 C: TSP #6 43 45 38 C: TSP #7 34 46 35 C: TSP + 0.5 35 46 29 C: TSP + 0.7 43 48 28 C: TSP #7 4 49 28 C: TSP #3 21 49 28 C: TSP #1
)
38 49 23 C: TSP #6 38 53 30 C: TSP #4
7.16 TABLE 7.3 (Page 10 of 10) 1B Steam Generator O' 1986 Eddy Current Examination
% THRU-WALL INDICATION (1)
ROW COLUMN PENETRATION PLUGGED LOCATION 43 53 33 C: TSP #7 41 56 26 C: TSP #6 33 59 29 C: TSP + 2.6 43 59 28 C: TSP #1 35 72 33 C: TSP #6 33 73 36 C: TSP #7 35 73 38 C: TSP #7 24 75 22 C: TSP #6 33 76 29 C: TSP #7 16 88 38 C: TSP #7 O 19 90 29 Ci TSP #7 (1) H - Inspection from Hot Leg C - Inspection from Cold Leg TSP - Tube Support Plate TS - Tube Sheet TE - Tube End V3 - #3 Antivibration Bar V4 - #4 Antivibration Bar Note that numbers added to TSP, TE, etc., are distances in inches above the indicated landmark, in the indicated leg. (2) SQR - Squirrel l O
8.1 8.0 PERSONNEL EXPOSURE AND MONITORING REPORT /- 8 ( Pursuant to 10CFR20.407(a)(2) and 20.407(b), a tabulation of the number of individuals for whom monitoring was provided is shown in Table 8.1. Tables 8.2, 8.3 and 8.4 provide a breakdown of the total number of individuals for whom personnel monitoring was provided. ' Table 8.1 TOTAL NUMBER OF INDIVIDUALS FOR WHOM PERSONNEL MONITORING WAS PROVIDED IN 1986 Range (mR) No. of Individuals No Measurable 278
< 100 183 100 - 249 96 250 - 499 92 500 - 749 59 750 - 999 37 1000 - 1999 30 2000 - 2999 5 3000 - 3999 0 4000 - 4999 0 5000 - 5999 0 6000 - 6999 0 7000 - 7999 0 8000 - 8999 0 9000 - 9999 0 10000 - 10999 0 11000 - 11999 0
> 12000 0 O Total Badged 780
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- 8.2 Table 8.2.
TOTAL NUMER OF CONTRACTORS PROVIDED WITH PERSONAL DOSE MONITORING DEVICES Range (mR) No. of Individuals No Measurable 171
< 100 101 100 - 249 58 250 - 499 49 500 - 749 30 750 - 999 30 1000 - 1999 20 2000 - 2999 5 3000 - 3999 0 Total Badged 464 O Table 8.3 TOTAL NUNBER OF WPSC PLANT STAFF PROVIDED WITH PERSONAL DOSE MONITORING DEVICES Range (mR) No. of Individuals No Measurable 61
< 100 50 100 - 249 29 j 250 - 499 35 500 - 749 25 l
750 - 999 6 1000 - 1999 9 2000 - 2999 0 3000 - 3999 0 l Total Badged 215 l
8.3 Table 8.4
- ' TOTAL NUISER OF PERSONNEL (WPSC NON-PLANT STAFF) PROVIDED WITH PERSONAL DOSE MONITORING DEVICES
]
Range (mR) No. of Individuals No Measurable 46
< 100 32 100 - 249- 9 250 - 499 8 500 - 749 4
,750 - 999 1 1000 - 1999 1 2000 - 2999 0 3000 - 3999 0 Total Badged 101 D A tabulation of numbers of personnel exposure and man-rem received by work and job function is shown in Table 8.5 in accordance with Section 6.9.1.b of the Kewaunee Nuclear Power Plant Technical Specifications. The table shows the total man-rem exposure for the year was 168.637.
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.Ta bl e i HP022/06PRNT ; U.S.N.R.C. REGULATORY GUIDE 1.16 - REPORTING OF OPERATING INFORMATION h
i STANDARD FORMAT FOR REPORTING NUMBFR OF PFRSONNFL AND MAN-REM DY WORK AND .300 FUNCTION ON VEAR OF.1986-KEWAUNEE NUMBER OF PERSONNEL (GT 100 MkEM) IOlAL MAN-REM = WOkK AND JOB FUNCTION STATION UTILITY CONTRACT STATION UTILITY. CONTRACT EMPLO(EES EMPLOYEES. WORK & OTHER 'EMPLOYLES EMPLOYEES WORK & OTHER a
; AEACTOR OPERATIONS j SURVEILLANCE i MAINTENANCE PERSONN. 3 O' 4 0.550' O.000 l '. t G9 i OPERATING PERSONNEL 17. .2 0 3.024 0.013 0.000 I HEALTH PHYSICS PERSONNEL 0 0 0 0.000 0.000 0.000 SUPERVISORY PERSONNEL to 0 0 1.194 0.000 0.000 ENGINEERING PERSONNEL 7 8' O. 0.502 1.786 0.000 1
ROUTINE MAINTENANCE 13.051 HAINTENANCE PERSONNEL 51 12 102 3.304' 40.263. l 0.169 2 OPERATING PERSONNEL 13 3 3 1.036 0.334 HEALTH PHrSICS PERSONNEL 19 0 13 9.791 0.000 7.392 0.271 i SUPERVISORY PER$JNNEL 3 1 2 0.039 0.772 ENGINEERING PERSONNEL 7 1 1 0.959 0.274 0.113 i j INSERVICE INSPECTION MAINTENANCE PERSONNEL ? O 17 0.009 0.000 1.375 OPERATING PERSONNEL 0 0 5 0.000 0.000 0.741 ] HEALTH PH(SICS PERSONNEL 0 0 0 0.000 0.000 0.000 SUPERVISOR ( PERSONNEL 1 0 0 0.000 .0.000 0.000 i j ENGINEERING PERSONNEL t 6 1 0.075 0.000 0.335 SPECIAL MAINTENANCE 4 MAINTENANCE PERSONNEL 50 10 74 11.772 1.218 36.385 UPERATING PERSONNEL 11 1 1 0.071 0.005 0.000-HEALTH PHYSICS PERSONNEL 9 0 0 0.901 0.000 0.000 IUPERVISORY PERSONNEL 4 't 17 0.379 0.278 14.635 ENGINEERING PERSONNEL 5 3 3 0.374 0.147 1.079 i i WASTE PROCESSING ] MAINTENANCE PERSONNEL 19 8 4 0.540 0.192 9.426
- OPERATING PERSONNEL 4 0 0 1.462 0.000 0.000 j HEALTH PHYSICS PERSONNEL 5 0 0 1.389 0.000 0.000
- SUPERVISORY PERSONNEL 1 1 0 0.000 0.002 0.000 I ENGINEERING PERSONNEL 0 1 0 0.000 0.000 0.000 I REFUELING ' MAINTENANCE PERSONNEL to 3 11 0.904 0.068 5.220 l OPERATING PERSONNEL 0 0 ,3 0.000 0.000 1.566 i ' HEALTH PHYSICS PERSONNEL 0 0 0 0.000 0.000 0.000 ! SUPERVISORY PERSONNEL 3 1 13 0.113 0.000 0.000 ENGINEERING PERSONNEL 4 0 0 0.230 0.000 0.000 i TOTAL MAINTENANCE PERSONNEL 135 33 214 26.826 . 4.782 '84.778 45 6 12 6.393 0.352 2.476, OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL 33 0 13 12.081 0.000 7.392 l TOPFRVISORY PFRSONNEL 22 4 tv s.7?S 1.052 14.906 tNoINEERINb PERSONNEL 24 13 5 2.140 - 2.207 1.527 . po l GRAND TOTAL 259 56 263 49.165- 8.393 111.079 il
9.1 9.0 RADIOLOGICAL MONITORING PROGRAM-Per your request, IEN 86-32 Supplement 1, the licensee radioactivity measurements attributed to the Chernobyl nuclear plant accident are included in the following report from Teledyne Isotopes on the Radiological Monitoring Program for the Kewaunee Nuclear Power Plant for 1986. o
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T TELEDYNE ISOTOPES / ~N, MIDWEST LABORATORY \d 1509 FRONTAGE RD. NORTHBROOK, IL 60062-4197 (312)5644700 REPORT TO WISCONSIN PUBLIC SERVICE CORPORATION WISCONSIN POWER AND LIGHT COMPANY nADIS0N GAS AND ELECTRIC COMPANY RADIOLOGICAL MONITORING PROGRAM FOR THE KEWAUNEE NUCLEAR POWER PLANT KEWAUNEE, WISCONSIN ANNUAL REPORT - PART I
SUMMARY
AND INTERPRETATION O January - December 1986 PREPARED AND SUBMITTED BY TELEDYNE ISOTOPES MIUWEST LABORATORY PROJECT N0. 8002 Approved by: [ff /6L V~ , (;, (luebner eneral Manager O 13 February 1987
= . .. . _ - . ,
N r' PREFACE The staff members of the Teledyne Isotopes Midwest Laboratory were responsible for the acquisition of data presented in this report. Assistance in sample collection was provided by Wisconsin Public Service Corporation personnel. The report was prepared by L. G. Huebner, General Manager. He was assisted in report preparation by other staff members of the laboratory. o f a f J s T O ii
y ~ 1( .. 4 - TABLE OF CONTENTS - Page Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 List of Figures . . . . . . . . . . . . . . . .-... . . ... . . iv List of Tables ......................... v
1.0 INTRODUCTION
' .-. . . . . . . . . . . . . . . . . ... . . . . . 1- } 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.0 RADIOLOGICAL-SURVEILLANCE PROGRAM ............... 4 3.1 Methodology . . . . . . . . . . . . . . . . . . . ... . . 4 3.1.1 The Ai r Program . . . . . . . . . . . . . . . . . . ~4 3.1.2 The Terrestrial Program . . . . . . . . . . . . . . 5 3.1.3 The Aquatic Program . . . . . . . . . . . . . .-. . 6
-3.1.4 Program Execution . . . . . . . . . . . . . . . . . 7 3.1.5 Program Modifications . . . . . . . . . . . . . . . 7 3.2 Results and Discussion . . . . . . . . . . . . . . . . . . 8 3.2.1 Atmospheric Nuclear Detonations and
- Nuclear Accidents . . . . . . . . . . . ... . . . 9 3.2.2 Tne Air Environment . . . . . . .......... 13
. 3.2.3 The Terrestrial Environment . . ......-.... 15 3.2.4 The Aquatic Environment . ............. 18 4.'O FIGURES AND TABLES ...................... 21 1
5.0 REFERENCES
. ......................... 38 APPENDICES A Interlaboratory Comparison Program Results .......... A-1 B Statistical Notations . . . . . . . . . . . . . . . . . . . . . B-1 C Maximum Permissible Concentrations of Radioactivity
.(') in Air and Water above Natural Background in Unrestricted Areas . . . . . . . . . . . . . . . . . . . . C-1 iii
LIST OF FIGURES No. Caption Page 4-1 Sampling locations, Kewaunee Nuclear Power Plant . . . . . . . 22 O O iv _ . _ _ . _ _ _ _ _ . _ . _ . . _ _ _ _ _ . . _ _ - . . . . _ , _ _ . _ . _ _ _ _ . . , . . . , _ _ _ _ . . _ . - - . , ._-. . _ ._ ___ _ _ _ . _ .~ -
- - .= . . . . . -- .
A q Q. l LIST OF TABLES i No. Title Page
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4.1 Sampling' locations, Kewaunee Nuclear Power Plant . . . . ... . 23-
- 4. 2 - Type and frequency of collection . . . . . . . . . . . . . . . 24 4.3 -Sample codes used in Table 4.2 . . . . . . . . . . . . . . . . 25 4.4 Sampling summary, January - December, 1986 . . . . . . . . . .
26 4.5 Environmental Radiological Monitoring Program Summary. . . . . 27 In addition, the following tables are in the Appendix: Appendix A Interlaboratory Comparison Program Results, A-1
. milk, water, air filters, and' food samples, 1983-1986 . . .. A-3 A-2 Interlaboratory Comparison Program Results, tnermoluminescent dosimeters (TLDs) . . . . . . . . . . . . A-12 Appendix C C-1 Maximum permissible concentrations of radioactivity in air and water above natural background in unrestricted areas . . . . . . . . . . . . . . . . . . . . . C-2 O
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-1. 0 INTRODUCTION The Kewaunee Nuclear Power -Plant is a 535 megawatt pressurized water reactor located on the Wisconsin shore of Lake Michigan in Kewaunee County. The Kewaunee Nuclear Power Plant became critical on March 7,1974. Initial power generation was achieved on April 8,1974, and the Plant was declared commer-cial on June 16, 1974. This report summarizes the environmental operation data collected during the period January - December 1986.
Wisconsin Public Service Corporation, an operating company for the Kewaunee Nuclear Power Plant, assumes the responsibility for the environmental program - at the Plant and any questions relating to this subject should be directed to them. f O, 1
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SUMMARY
Results of sample analyses during the period January - December 1986 are ' summarized in Table 4.5. Radionuclide concentrations measured at indicator locations are compared with levels measured at control locations and in preoperational studies. The comparisons indicate background-level radio-activities in all samples collected with the following exceptions: A. Effect of the nuclear' reactor accident on April 26, 1986 at-Chernobyl (USSR) on the environment in the vicinity of the Kewaunee Nuclear Plant.
- 1. Iodine-131 was detected in 14 charcoal samples collected between May 13 and June 10,.1986. The concentrations ranged from 0.08 to 0.81 pCi/m3 and averaged 0.39 pCi/m3.,
- 2. Tnere was an increase in gross beta activity in airborne particulates in the samples collected between May 12 and June 10, 1986.
- 3. The isotopes of Ru-103, Cs-134 ano Cs-137 were detected in the second quarter air particulate composites.
- 4. Iodine-131 was detected in 25 milk samples collected between May 20 and July 15, 1986. The concentrations ranged from 0.6 to 61.5 pCi/l and averaged 7.9 pCi/1.
- 5. Cs-137 was detected in one milk sample collected May 21, l 1986 (11.1 pCi/1).
L
- 6. The isotopes of Ru-103, Cs-134 and Cs-137 were detected J.
in several grass samples collected in early June 1986. O 2
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V ' B. Presence of radioisotopes listed oelow in bottom sediments, periphyton, and water are probably plant related.-
- 1. Trace amounts of Co-58 (0.115 pCi/g dry weight), Co-60 (0.076 pCi/g dry weight), Cs-134 (0.077 pCi/g dry weight) and Cs-137 (0.04 pCi/g dry weight) were detected in several bottom sediment samples collected in the discharge area. The levels are insignificant.
- 2. Trace amounts of Co-58 (0.116 pCi/g wet weight), Co-60 (0.076 pCi/g wet weight) and Cs-137 (0.035 pCi/g wet weight) were detected in several periphyton samples l collected in the discharge area. The levels are insig- l nificant.
NOTE: The presence of Co-58 and Co-60 in bottom sedi-ments and periphyton is probably plant related. The origin of Cs-137 is not clear since similar levels are found in control samples. The source could be the nuclear plant or previous nuclear tests or both.
- 3. Low levels of tritium were detected in several samples collected in the vicinity of the discharge and in one
(~N sample from a creek on site. The highest level, 2590 pCi/l above the background level, constituted less than 0.09% of the maximum permissible concentration of 3,000,000 pCi/l established in the 10 CFR 20 document. Tritium levels of this magnitude are not significant. l l l t 3
~ 3.0 RADIOLOGICAL SURVEILLANCE PROGRAM Following is a description of the Radiological Surveillance Program and its execution. I 3.1 ' Methodology The sampling locations are shown in Figure 4-1. Table: 4.1 describes the locations, lists for each its direction and distance from the reactor, and indicates which are indicator and which are control loca-tions. The sampling ' program monitors the air, terrestrial, and aquatic envi-ronments. The types of samples collected at each location'and the frequency of collections are presented in Table 4.2 using sample codes defined in Table 4.3. The collections and analyses that' comprise the program are described below. Finally, the execution of the program in the current reporting year is discussed. 3.1.1 The Air Program Airborne Particulates The airborne particulate samples are collected on 47 mm diameter glass fiber filters at a volumetric rate of approximately one cubic foot per minute. The filters are collected weekly from six locations (K-1f, K-2, K-7, K-8, K-15, and K-16), and dispatched by mail to TIML for radiometric analysis. ~ The material on the filter is counted for gross beta activity approximately five days after receipt to allow for decay of naturally-occurring short-lived radionuclides. Quarterly composites from each sampling location are analyzed for gamma-emitting isotopes by a germanium detector. O 4
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<a Airborne Iodine-Charcoal filters are located at locations K-1f,' K-2, K-7, K-8, K-15, and K-16.
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The filters are changed bi-weekly : and analyzed for iodine'-131 immediately af ter arriv'al at .the laboratory. Ambient Gamma Radiation - TLDs The: integrated. gamma-ray background is measured ~ at six air sam-pling locations (K-1f, K-2, K-7, K-8, K-15,~ and _ K-16), at four milk sampling locations (K-3, K-4,;K-5, and K-6), and at two-additional locations (K-17, located .4.25 miles west of the plant; and .K-27, located 1.5 miles northwest of- the plant) -with thermo-luminiscent dosimeters (TLDs). CaF2 :Mn bulb TLDs are. exchanged - quarterly and annually.
. Precipitation
$ Monthly composites of precipitation samples collected at K-11 are analyzed for tritium _ activity by liquid scintillation technique.~ i 3.1.2 The Terrestrial Program t Milk Milk samples are collected semimonthly-(one gallon from each L , location)' from May through October and monthly (two gallons from each ' location) during the ' rest of the year from four herds that graze within four miles of the reactor site .(K-4, X-5,
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K-12, and K-19) from -two herds that . graze between four, and ten i miles from the reactor site (K-3 and K-6), and from a dairy in Green Bay ,(K-28). : The milk samples are analyzed for iodine-131, ' strontium-89, and -90, cesium-137, barium-140, potassium-40, calcium, and stable potassium. l- Well Water .
- l. One gallon water samples are collected quarterly from four off-site wells located at K-10, K-11. K-12, and K-13 and from two on-site wells located at K-1g and K-lh.
I The gross beta and gamma spectroscopic analyses are performed on i the total residue of each water sample. Tne concentration of potassium-40 is calculated from total potassium, wnich is deter- .t mined by flame photometry on all samples. + y Additionally, samples of water from two on-site wells (K-1g and K-lh) are also analyzed for gross alpha. Water from one on-site well (K-19) is also analyzed for tritium, strontium-89, and strontium-90. 5 l
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L) Domestic Meat Domestic meat samples (chickens) are obtained annually (in the
, _ third quarter) at locations K-20, K-24, K-25, and K-27. The flesh is -separated from. the bones, gamma scanned, and analyzed for gross alpha, and gross beta activities.
E_g.gs Eggs are collected quarterly at Location K-27. The samples are gamma scanned and analyzed for gross beta, strontium-89, and strontium-90 activities. Vegetables Vegetable samples (5 varieties) are collected at locations K-18
- and K-26, and two varieties of grain, if available, at location K-23. The samples are gamma scanned and analyzed for gross beta, strontium-89, and strontium-90 activities.
Grass and Cattle Feed Grass samples are collected during the second, third and fourth quarters from two on-site locations (K-lb and K-1f) and from six . dairy farms (K-3, K-4, K-5, K-6, K-12, and K-19). The samples b]
'/ are gamma scanned and analyzed for gross beta, strontium-89, and strontium-90 activities. During the first quarter cattle feed is collected from the same six dairy farms, and the same analyses are performed.
Soil Soil samples are collected twice a year on-site at K-1f and from the six dairy f arms (K-3, K-4, K-5, K-6, K-12, and K-19) . The samples are gamma scanned and analyzed for gross alpha, gross beta, strontium-89, and strontium-90 activities. 3.1.3 The Aquatic Program Surface Water One-gallon water samples are taken monthly from three locations on Lake Michigan: 1) at the point where the condenser water is discharged into Lake Michigan (K-1d); 2) at Two Creeks Park (K-14) located 2.5 miles south of the reactor site; and 3) at the Rostok water intake (K-9) located 11.5 miles north of the reactor site. Additionally, one-gallon water samples are taken monthly from three creeks that pass through the site (K-la, K-lb, and K-le). Samples from North and Middle Creeks (K-la, K-lb) are collected near the mouth of each creek. Samples from
/' the South Creek (K-le) are collected about ten feet downstream from the point where the outflows from the two drain pipes meet.
Water samples at K-14 are collected and analyzed in duplicate. 6
3 6 3 4 The : water samples are gamma scanned .and analyzed for gross beta
- activity- in theltotal residue, dissolved solids, 'and. suspended solids, and potassium-40._ The concentration of potassium-40 is calculated from total potassium, which is determined by flame photometry. ~The tritium activity is determined by liquid scin-tillation technique. Quarterly composites of. monthly grab samples are _ also . analyzed for tritium, ' strontium-89 and - strontium-90.-
x Fish i Fish zsamples. are collected _ in the second, third, and fourth -i quarters at-Location K-1d. The flesh is separated from the bones, gama scanned and analyzed for gross beta activity. Ashed bone samples .are analyzed for gross beta, strontium-89 and strontium-90 ~' activities. Slime-Slime samples are collected during the second and third quarters-from three Lake Michigan locations (K-1d, K-9, and K-14), and from three creek locations (K-la, K-lb, and K-le), if available. The. samples are analyzed for gross beta activity. If the quantity
;is sufficient, -they are also gamma scanned and analyzed for
-strontium-89 and strontium-90 activities.
i-Bottom Sediments Bottom sediments are collected in May and November from five loca - tions (K-1c, K-1d, K-lj, K-9, and K-14). . The samples are analyzed for gross beta activity and for strontium-89 and strontium-90. . Each sample is also gamma ' scanned. Since it is known that the measured radioactivity per unit mass of sediment increases with decreasing particle size, the sampling procedure is designed to assure collection of very fine particles. 3.1.4 Program Execution Program execution is summarized in Table 4.4. The program was executed as described in the preceding sections. 3.1.5 Program Modifications The following program modifications were made in 1986.
- a. Analysis for gross alpha was discontinued on all samples, except on two on-site wells (K-19 and K-lh)), domestic meat, and soil.
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d'g b. The frequency- of collection of milk during the summer months. was changed from Weekly to semimonthly. One location (Green- Bay, K-28) was added to the program.
- c. Two TLD locations, K-17 and K-27, were added to the ,
program. Annual TLDs were, retained but are used as an
' emergency set, f
- d. The frequency of collection frpm off-site wells was changed from monthly to quarterly. Gamma spectroscopic analysis of all samples 'was added to the program.
- e. Gamma spectroscopic analysi , of all sur'ade water samples was added to the program.
- f. The frequency of collection for 'Dottom '. sediments was changed from four times a year to semiannutlly.
- g. Analysis for Sr-89 and Sr-90 in . surf >ce water w'as changed from quarterly at three locations (K-1d, K-9, and K-14) to quarterly at all locat"ons (K-la, K-lb ,
K-1d, K-le, K-19, and K-14) . . 3.2 Results and Discussion The results for the reporting period January lo December 1986 are presented in summary form in Table 4.5c. For each type of analysis of . each sampled medium, this table shows = the annual mean and range for all indicator locations and for all control locations. The location with the highest annual mean and the results for this location are also given. , The discussion of the results has been divided into three broad cate-gories: the air, terrestrial, and aquatic environments. Within each category, samples will be discussed in the order listed in Table 4.4. Any discussion of previous environmental data for the Kewaunee Nuclear Power Plant refers to data collected by Tel,edyne Isotopes Midwe'it Laboratory or its predecessor, Hazleton Environmental Sciences. < The tabulated results of all measurementi ede in 1936 are not included in this section, although referencies :ta tn'ese results will De made in the discussion. The complete tabulation of the 1986 results is contained in Part II of the 1986 annual report on the Radiological Monitoring Program for the Kewaunec fluclear Power Plant. The nuclear accident at Chernobyl (see Section 3.2.1) had an effect on the level and type of radionuclides detected in some environmental samples, specifically in air particulates, air iodine, milk, and grass. This effect is discussed in the following sections. 8
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x '3.2.1 Atmospheric Nuclear Detonations and Nuclear Accidents There were no reported atmospheric nuclear tests in 198C. The ^ - .last teported test-was conducted by the People's Repc iic of China 'on October 16, 1980. The reported yield was in the L 200 kiloton to 1 megaton range.
~
Tnare was an accident at Reactor .No. 4 of the Chernobyl Nuclear i Dian't complex in the U.S.S.R. The accident occurred on April 26, 1986. 'The following is a brief description of the reactor, its locaticn, and the chronology of events. 3 Location 4 L The Chernobyl nuclear complex is located in the semirural area known as the Byelorussian-Ukrainian Woodlands on the Pripyat J j I f Riyer in the Soviet Ukraine. The nearest town is Pripyat, six - .6 miles away, with a population 'of' about 25,000. An additional r *- ! 25,000 people live within twenty miles, many of them in the old townfofChernobyl. Eighty miles to the south lies Kiev, the capica't city of the Ukraine, with a population of 2.5 million. 2 Reactor Reactor No. 4 is one of six similar reactors at the Chernobyl site. It is a graphite-moderated, light-water-cooled pressure s tube reactor, RBMK-1000, with 1000 MW(e) and 3200 MW(th) capac-
- p. ity. Reactors No. 1 and 2 went into operation in 1977, and Reactors No. 3 and 4 went into operation in 1983; construction of Reactors No. 5 and 6 began in 1981 but has not been completed.
E All four completed reactors were in operation at the time of the accident. There are fourteen R8MK-1000 reactors and one BBMK-1500 [1500 MW(e)] in operation within the Soviet Union. Out of approximately fif ty operating reactors in the U.S.S.R. ,
!W a'oout half are graphite-moderated and half are water-moderated, similar to -Western design. In the United States, the only
' reactor similar to the RBMK-1000 is the graphite-moderated
,r.eactor in Hanford, Washington, which is used exclusively for the ,
- production of weapons-grade plutonium. Another major difference between Western and Soviet reactors is that Western reactors have a steel and concrete containment surrounding the complete reactor 4 while Soviet reactors do not. There are certain advantages and disadvantages of the Chernobyl-type reactors. The advantages
. are:
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' - .a) The' absence of cumbersome pressure vessels;-
'b) Fuel elements ?can be withdrawn - for exchange, repair,'oh extractiop of. plutonium without necessi :
'tating reactor- shutdown; and L; .s .
q ~ c)4 Only about 2% enrichment with U-235.is required, in. comparison with 4% for - water-moderated reactors.
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- The disadvantages are: .
a) .High sensitivity of the neutron field, requiring a '>-
, complicated control' system Jin' order to stabilize .
thel power density distribution in.the core; b) ' Complex branching .of the. coolant delivery Land
. removal system for each channel, which ;is subject to frequent small-scale failures;'
c) A ~ large amount of heat energy. accumulating in the
- metal structures, fuel elements, and ig raphite
- structure, and' slightly radioactive -steam in the s ..
. turbine; and d) Inherent danger. of hydrogen production when _leaksi I or a breakdown of channels . carrying hot steam and water can bring hot steam in contact with graphite, which can ' lead to a chemical explosion (and did). /
,u At the time of the accident, 75% of the fuel elements were It is estimated that the reactor was' inM
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first-load bundles. operation for about 400 to 600 days. This means that the core had a large and nearly maximum inventory of radioactive fission- 9 5-products. and that the accident could not have happened under worse conditions. Chronology of Events The reactor was in the process of being shut down-for the maintenance of one of two turbogenerators and to perform a test.- The test was designed to determine how long the tdrbo-generatw could provide electrical power utilizing the mechanical energy of the rotor without steam supply.
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Apri1 25,.1986-0100 hr.: Shutidown started. 1305 hr' . :. ' Turbogenerator No. 7 switched off. Electrical needs;- y' for feed pumps and other equipment switched to turbo-
-generator No.18 (to be' tested).
1400 hr. : Emergency ' core cooling ' system disconnected, but '. unit -
. continued-to operate..
2300 hr.: . Power reduction resumed'. . At.tnis time, tne power.
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- level dropped -to about'i30 MW(th), which is an Lunstable '
condition.-- .The operators brought it tup to 200 MW(th) by withdrawing control' rods and- thenTby reducing the-cooling -water - flow. The . reactor started to heat up. t
-Since all automatic controls were very, quickly.
turned off, the reactor could not' scram itself. April 26, 1986> 0123 nr. : Some of the fuel rods and water tubes . burst and steam reacted with zirconium ;and graphite, ' producing large : 4 p)
. ( :, quantities of- hydrogen .and other gases. There were two explosions, one after the'other.
4-There were three stages of releases. Stage 1 During this stage, occurri.ng immediately after the explosion, parts of irradiated fuel and gases escaped to the atmosphere Some of the D'urning debris fell on tne- adjoining ' buildings and started a fire. .' Also; , grapnite which was; enclosed 'in a stainless steel container was exposed to the atmosphere and started burning. - -Stage 2 This stage lasted from April 26, 1986 to May ' 2, 1986. The releases -of radioactivity ' were similar to those during Stage 1. The temperature inside .the reactor
- continued to rise.
I Stage 3 Tnis stage lasted from May 2, 1986 to May 6, 1986. During. this stage, there was a sharp increase in tne rate of fission products releases due' to the high E. temperature (2000* C) caused by burning graphite. 11
...a-.-.. . _ _ _ _ . _ . _ . . _ . _ _ . _ , _ _ _ . _ . _ , . _ _ _ _ , _ _ - . _ . _ _ _ _ _ , .
1
~
l' in X -
'~ -
L On. May l 6, .1986, the? reactor -was ' sealed and releases of- fission
. products' to . the' ' atmosphere were reduced to negligible ~ amounts.
Soviet scientists (IAEA Report, 1986) have estimated 'that about 50 l mci ofc radioactivity E(excluding noole gases) were1 released
* .during. these three stages and- constituted about ' 3.5% of the total
. inventory of , isotopes in . the. reactor. .It should be noted that -
-most of the noble gases-and volatile elements (Kr, Xe, I, Te, Cs)
- -escaped'from the reactor. The estimated. releases are listed-below. For comparison, releases from the Three Mile Island and -
Windscale accidents and from a 20 KT -nuclear test are f also listed. 1
~ Releases (in megacuries)
' ~ Incident Noble Gas I-131- Cs -137-- Sr-90 Chernobyl* 100-200 10-50 1-6 0.001-0.007 Three Mile' Island. 10 0.00017 <0.0001 . <0.0001 Windscale' - 0.3-0.4 0.2 0.001 -- l 20 KT ground atomic bomb test 5 2 0.004 0.004
-* Evaluation error 150%.
Because of the high temperature, gases .and aerosols went straight up :and reached an estimated height of 16,000- feet. The disper-sion pattern was similar to that of a tall stack of a coal' plant;
-it reduced local fallout while increasing deposition at distant
~ downwind locations.
In addition to the aforementioned isotopes, the following iso-topes and their total releases (from April 26 to May 6, 1986) were_ estimated in megacuries. 1 - IO ^ l 12
..._.. _ _ _._ _ u..-.,-_ _ _ _ _ _ _ . _ _ _ ._ - . _ _ . _ . _ .. _ _ _ _ _ _
a
'wJ Total . Tot al'
. Release
- Release *
-(Estimated (Estimated
-Isotope in mci) Isotope in MC1)
Xe-133 45 Ce-141 2.8 Kr-85m -- Ce-144 2.4 Kr-86 0. 9 Sr-89 2.2 I-131 7.3 Sr-90 0.22 Te-132- 1.3 Pu-238 0.0008 Cs-134- 0. 5 Pu-239 0.0007 Cs-137 1. 0 Pu-240 0.0011 Mo-99 3.0 Pu-241 0.14 Zr-95 3.8 Pu-242 0.000002 Ru-103 3.2 Cm-241 0.021 Ru-106 1. 6 Np-239 1. 2 Ba-140 4.3
- Evaluation error i 50%.
fs The prevailing wind direction at the time of the accident was in ('~) a northwesterly direction. It is estimated' that the radioactive cloud reached the' Polish border on April 26, 1986 and arrived over Sweden in the afternoon of April 27, approximately eighteen (18) hours before it was detected at the Forsmark Nuclear Plant, located 125 miles northwest of Stockholm. As a result of the accident (as of August,1986), thirty-one (31)' people died, two -(2) from the explosion and twenty-nine (29) from radiation sickness; two hundred ninety-nine (299) people were hospitalized; and one thousand people were injured. ' The long range effects of the accident are yet to be determined. 3.2.2 The Air Environment Airborne Particulates In air particulates, the annual gross beta concentration was nearly identical at indicator and control locations (0.031 and 0.028 pCi/m3, respectively) and was hi her than in 1985 by a f actor of two (0.016 and 0.015 pCi/m , respectively). The increase of activity was due to the Chernobyl accident. The radioactive debris was first detected in samples collected on May 12, 1986, and continued through the collection period ending June 10, 1986. The next collection, June 17, 1986, showed that O- gross Deta activity returned to the pre-Chernobyl level. 13
. , _ _ , _ , . _ . ~ -
- _ _. _. 4 - _ _ ._ _
m ~ - b } + ke '] or: l - r , 4 The spring . peak,; wnich has;been , observed almost -annually. for many ~
- years (Wilson Let al,e1969) could not' be .. identified because .it was overshadowed-by releases-of. radioactivity from Chernobyl', .
' Gamma - spectroscopic analyses .of quarterly composites - of; air -
fparticulate1 filters yielded .similar-:results fore indicator and - i
, control ' locations. . - Beryllium-7, which is produced continuously.
~
+ -
.in 1the upper atmosphere by cosmic radiation -( Arnold- andJAl-Salin, 1955), :was detected in all samples. All . other"gama-emitting
" - ' isotopes 'were' below' their respective LLDflimits during the-
- first', third and fourth quarters.
IL During the second quarter, .tnree gamma-emitting . isotopes, .Ru-103,
. Cs-134 -and - Cs-13j, . were identified and averaged 0.0087, 0.0058,.
g and ' 0.013 pCi/m , respectively. It should be noted that the ratio of Cs-134 to Cs-137 measured in; air particulates was about. - 1: - 1:2, the same _-as in the mix of gases released at the-time of the accident' at Chernobyl as reported by Soviet authorities. Airborne Iodine Of the '156 analyses for airborne iodine-131 in 1986, 142 were-below the LLD of 0.03 pCi/m3, p;~ Iodine-131 levels in 14 samples collected netween May 13 and June i 10, pCi/mg986_ ranged from 0.08 to 0.81 pCi/m3 :and. averaged 0. l The presence of airborne I-131 in the environment in May and June is attributable to the Chernobyl accident.
~
Ambient Gamma Radiation'- TLDs
. Ambient gama radiation was monitored by TLDs at twelve locations:
- six indicator and six control.
L The quarterly TLDs at the indicator locations measured a mean dose equivalent of (16.0tl.4)* mR/91. days, in agreement' with the E mean at the control locations of (14.8tl.5) mR/91 days, and were about the same as the means obtained in 1985 (16.1 and 15.5 mR/91 days, respectively). All- these values are slightly lower than the United States average value of 19.5 mR/91 day; due to natural e background radiation (National Council on Raatation Protection and Measurements, 1985). The highest annual mean was 18.6 mR/91 days at indicator location K-17.
*. Unless otherwise indicated, uncertainties of average values are standard deviations of the individual measurements over the period averaged. Uncer-tainties of- individual measurements represent probable counting errors at the 95% confidence level.
~
i' 14 U 4 w, . . - , ,m,. ~,,,....m-.,__ _...--,_,--.,,-,._m.m_---.,,. -
, - - , , . , . - _ _ _ _ , , = . _ . . - _ , _ _ .
~-
=_ ;
~ s T- a
- " Precipitation
' Precipi_tation: was- monitored only at an indicator location, K-11. The- tritium ' concentration was ' below the J-LLD
- level of : 330 - -
lpCi/1.in all samples. 3.2.3 The. Terrestrial Environment e Milk.' Of the 126. ' analyses for iodine-131.in milk,- 101 'were below, the LLO level'of 0.5 pCi/1. Twenty-five milk samples . collected between . May -20 and July 15,'
-1986 ' contained 1-131 in: various' concentrations. The'1evel. ranged from 0.6 to ,61.5 pC1/1 with ' the average for > all samples' of 7.9; Variability in ~ concentration of I-131 1n ' samples. from pCi/1.
different farms-depended _on the methods of feeding the-cows'. The presence of I-131rin ' milk collected during this ~ period is
-attributable to the:Chernobyl accident.
1
. Strontium-89 concentration was- below tne LLD level ' of 1.5 pC1/1 in' all samples.
Strontium-90 was found in all samples. The.mean v.alues were-nearly identical for -indicator and - control locations (2.4 pCi/l and 2.5-pCi/1,.respectively). Barium-140 concentration was below the .LLD of 10 pCi/l in all
-samples. Cesium-137. concentration was also below the LLD of 10 pCi/1 in.all' samples but one. The sample collected on May 21, 1986 from. Location K-12 had a Cs-137 concentr'ation of 11.1' pCi/1.-
and-is attributable to"the Chernobyl accident. Potassium-40 results were nearly. identical at both the indicator. and control locations and were essentially identical to.the levels observed in 1978 througn 1985. p Due to the cnemical similarities between strontium and . calcium, I and cesium and potassium, organisms tend to deposit ~ cesium-137 in " the soft' tissue and muscle and strontium-89 and -90 in the bones. ' Consequently, the ratios of strontium-90 activity to the' weight of calcium in milk and cesium-137 activity to the weight of potassium in milk were monitored in order to detect potential E environmental accumulation of these radionuclides. No statis- ! tically significant- variations in the raios were observed. The measured concentrations of stable potassium and calcium are in agreement with previously determined values of 1.5010.21 g/l 7 and 1.16i0.08 g/1, respectively (National Center for Radiological Health,1968). p
- 15
! i
. . . - . - . _ _ . . _ _ . , . . _ _ . _ - - _ _ . . _ _ . . . , _ . . . . _ . . _ . , _ . . _ . . _ _ _ ~ . . . _ - _ _ . . .
./ \ D Well Water
. Gross alpha concentration in well water was below the LLD level of 3.0 pC1/1 in all samples.
Gross beta concentration in well water was 1.2 pCi/l'in samples from the control location. The mean value for all indicator locations was 2.5 pCi/l and was ' nearly identical to the values observed in 1977 through 1985 (3.3, 3.4, 3.0, 3.0, 3.6, 3.2, 2.9, 2.3 and 2.6 pCi/1, respectively). Tritium concentration in the on-site well (K-1g) was below the LLD of 330 pCi/l in all samples. The concentrations of strontium-89 and strontium-90 in well water were below their respective detection limits. Potassium-40 levels were quite low (under 3.4 pCi/l), in agree-ment with the previously measured values. Domestic Meat In meat (chickens), gross alpha concentration was similar at both D indicator and control locations - (0.11 and 0.13 pCi/g . wet weight, ( respectively). Gross beta concentration averaged 3.11 pCi/g wet weight for indicator locations and 3.59 pCi/g wet weight for control locations. Gamma-spectroscopic analyses showed that most of the beta activity was due to naturally occurring potassium-40. All other gamma-emitting isotopes were below their respective LLD limits. ELgs, In egg samples, gross beta concentration averaged 0.99 pCi/g wet weight, about equal to the concentration of the naturally-occurring potassium-40 observed in the samples (1.23 pCi/g). All other gamma-emitting isotopes were below their respective LLD's. The level of strontium-89 was below the LLD of 0.004 pCi/g wet weight in all samples. Strontium-90 was below the LLD level of 0.002 pCi/g wet weight in all samples. Vegetables In vegetables, gross beta concentration was slightly lower at the indicator location (2.16 pCi/g wet weight) than at the control location (2.45 pCi/g wet weight) and was due primarily to the potassium-40 activity (2.06 pCi/g wet weight). Strontium-89 was detected in one sample (0.091 pCi/g wet weight). Strontium-90 activity .was below LLD at the indicator location and averaged Ov 0.0021 pCi/g wet weight at the control location. All other gamma-emitting isotopes were below their respective LLO levels. 16
(%
]
The samples of oats and wheat were of similar composition but the concentration of radionuclides was slightly higher due to the lower water content of tne grain in comparison with the vegatables. Grass and Cattle Feed In grass, gross beta concentration was actually identical at both indicator and control _ locations (6.75 and 6.73 pCi/g wet weight, .respectively) and in both cases was predominantly due to naturally occurring potassium-40 and beryllium-7.- All other gamma-emitting isotopes were below their respective LLD's. Strontium-89 was below the LLD of 0.13 pCi/g wet weight in all samples. Strontium-90 activity was detected in all samples and was slightly higher at control than at indicator locations (0.020 and 0.016 pCi/g wet weight, respectively). The presence of radiostrontium in grass samples is attributed to the fallout from the previous nuclear tests. For cattlefeed, the mean gross beta concentration was slightly . higher at control locations (10.3 pCi/g wet weight) than at indicator locations (7.5 pCi/g wet weight). The highest gross beta level was in the sample from the indicator location K-4. (12.1 pCi/g wet weight), and reflected the high potassium-40 level (11.5 pCi/g wet weight) observed in the sample. The pattern -Q V_ was similar to that observed in 1978 through 1985. Strontium-89 levels were below the LLO level of 0.007 pCi/g wet weight in all samples. Strontium-90 activity was lower at indicator locations than at control locations (0.064 and _0.096 pCi/g wet weight, respectively). The presence of the radiostrontium is attributable to the fallout from the previous nuclear tests. All other gamma-emitting isotopes were below their respective LLD levels. Soil No significant differences were found between indicator and control values in soil samples. The difference of 2.1 pC1/g dry weight in mean gross alpha concentration between indicator locations and control locations is not statistically significant because the counting uncertainties of the individual measurements are typically 3-5 pCi/g dry weight. Mean gross beta levels were similar at both indicator and control locations (18.7 and 20.8 pCi/g dry weight, respectively), and is primarily due to the potassium-40 activity. Strontium-89 was below the LLD level of 0.056 pCi/g dry weight in all samples. Strontium-90 was detected in all samples and was higher at control than at indicator locations (0.089 and 0.057 pCi/g dry weight, respectively). Cesium-137 was detected in all samples and was higher at control locations than at indicator locations (0.47 and 0.36 pCi/g dry weight, respectively). All other gamma-emitting isotopes were below their respective LLD's. Tne levels of detected activities were similar to those observed in 1979 through 1985. 17
y, w m
.. . ~ ,
+
t 3.2.4 The Aquatic-Environment Surface Water'
~
x 'In l surface:-water, mean gross' beta ; activity in suspended solids was detected in one . sample and was 2.l' pC1/1. Mean gross beta. concentration in dissolved solids was higher oy a factor"of -two
^
'at indicator locations (4.1 pCi/1) as compared _ to ' the control locations '(2.1 pCi/1) and was nearly identical to the activities observed - in 1978 (5.4 and 2.7 pCi/1, 1979 (5.7 - and 2.7 pCi/1),
1980. (5.1 and 2.7 pCi/1),1981 (4.3 and 2.7 pCi/l),1982 (4.9 and L 2.4 pCi/1),- 1983 -(5.1 and 2.6 pCi/1),1984 (5.0 and 2.7 pCi/1) ~ and 1985 (5.6 and_2.5 pCi/1). The control sample is Lake Michigan
+
water cwhich while . varies indicator. very little samples include in -concentration during (the
.. two creek locations- K-la' year, and.
L- 'K-le) which are much higher in ' concentrations and exhibit large r -month-to-month variations- in gross beta concentration. The K-la creek - draws its water from the surrounding fields which are heavily fertilized and the K-le creek draws its water mainly from the . Sewage -Treatment Pond No. 1. In general, gross beta concentration -levels were nign-when potassium-40 levels were nigh
' and Iow when potassium-40 levels were low, indicating that the fluctuations in beta concentration were due _to . variations 'in 1 -potassium-40 contentrations and not to plant operations. The fact L 1 that'similar fluctuations at these locations were observed in the
'~ pre-operational studies conducted prior to 1974 supports - this assessment. Annual mean . tritium concentration was 1160 pCi/l at indicator locations and was below the LLD of 330 pCi/l at control locations. The concentration in one quarterly composite (fourth quarter) at
- the discharge (K-1d) was 420 pCi/1 above the background level of 330 pCi/1. The elevated level of 420 pCi/l above background in the discharge water is attributable to the plant operation, but constitutes less than 0.02% of the maximum permissible concentra-tion of 3,000,000 pCi/1 establisned in the 10 CFR 20 document.
The highest level of 2590 pCi/l above background level was detected in the first quarter composite from South Creek on site (K-le). The highest level measured at Two Creeks Park was 180 pCi/l above background level and constitues about 0.006% of the permissible level. However, because of the associated counting error, a concentration of this low magnitude is indistinguishable from the background level. Strontium-89 activity was below the LLD of 1.8 pCi/l in all samples. Strontium-90 activity was detected in three of twenty-eight samples and averaged 1.1 pCi/1. All gamma-emitting isotopes were Delow their respective LLDs in all samples. 18 I . . . . . . . . .
y ' ( Fish In fish samples, the gross beta concentration averaged. 2.3 pCi/g : wet weight in muscles and 3.1 pCi/g wet weight;in-bone fractions. In muscle, the gross beta concentration was primarily'due to potassium-40 activity. The . average'_ beta concentration of 2.70 ' pCi/g wet weight: was near the average, of the11973 range of 2.26
~
to 3.62 pCi/g wet ~ weight. The cesium-137_ concentration .in muscle-averaged 0.11 pCi/g ~ wet weight and was' nearly ident.ical to -_the level observed in 1979 and 1980 -(0.~12 pCi/g wet weight' in both 4 years). 1981 (0.15 pCi/g wet weight), 1982 - (0.17_ pC1/g wet ~ weight),- 1983 (0.14; pCi/g wet weight), 1984 (0.'10 pCi/g wet-weight) and 1985-(0.11 pCi/g wet weight).- The strontium-89.
' concentrations were below the LLD of 0.96 pCi/g wet weight 'in all samples. _ Strontium-90 was detected in all samples' and averaged 1-0.25 pCi/g wet weight.
[ Periphyton (Slime) In periphyton-(slime) samples, mean gross beta concentration
~
i was- higher at indicator than at control locations' (3.86 and 2.30 pCi/g wet . weight, respectively). Strontium-89 concentration -was - below the LLD level of 0.061- pCi/g' wet weight 'in. all samples. _. ~' . Strontium-90. concentrations'were similar at indicator and control locations, . averaging 0.053 and 0.039 pCi/g wet weight, respec-tively. A trace quantity of Co-58 (mean 0.116 pCi/g wet weight) was detected in two samples' at location K-Id and trace quantities . on Co-60 (mean 0.074 pCi/g wet weight) were ' detected in two ' samples collected _ at indicator locations K-lb and K-1d. Cs-137 was detected in three samples, one at control location K-9 (0.059 pCi/g wet weight) and two at indicator ~ locations K-la and K-lb (mean 0.035_ pC1/g wet weight). All other gamma-emitting isotopes, except naturally-occurring beryllium-7 and potassium-40,
- were below their respective LLDs.
Bottom Sediments In bottom sediment samples, the mean gross beta concentration was slightly higher at indicator locations than at the control location (8.1 and 7.2 pCi/g dry weight, respectively) and was due mostly to potassium-40. The difference is not statistically significant. i' Cesium-137 was detected in nine of ten samples and averaged 0.04 pCi/g dry weight at indicator locations and 0.02 pCi/g dry weight at control locations. Cs-134 was detected in one sample at Location K-lj and was 0.077 pCi/g dry weight. The cesium-137 level was slightly lower than the levels observed in 1979 (0.12 L pCi/g dry weight), in 1980 (0.19 pCi/g dry weight), in 1981(0.18 pCi/g dry weight), in 1982 (0.13 pCi/g dry weight), in 1983 (0.16 1 19
, .,-,-i -m,,+< * -,m,,,r,-w.r----%-., - -m.w,,,,,um,w,,,,.--ew,,w-,-,ww,--w,e-. ,,v --..-.-w,~.,_-.,.---.-,+y,.--.w..-,---.r...r--,.--
. pCi/g dry weight), in 1984 (0.07 pCi/g . dry weight)- and in --1985
- (0.05 - pCi/g - dry weight). Strontium-89 and strontium-90 levels 4 were below their respective LLDs (0.040 and 0.015 pCi/g dry weight, respectively) in all samples. Trace- amounts of Co-58 (five samples, mean 0.115 pCi/g dry ' weight) and Co-60 (six samples, mean 0.076 pCi/g dry weight).were detected near the condenser discharge and at K-14. The presence of trace ~ amounts of these activation products in bottom sediments is.probably.
plant related. O I A 1 r 1 i O 20 i
- - - _ . , . . _ . _ _ _ _ _ _ _ _ _ _ _ - ~ . . _ _ . _ . _ _ _ . _ . . _ _ _ _ _ . . _ -
Aublump q Wum m b u 10 1 4.0 FIGURES AND TABLES O O 21
7 . :. r K-9 '] KEWAU$EE $9.:.F d. NUCLEAR POWER PLANT $.~2~'. ?'.! . 4
- ; u.K 2' f .. b..
Kewaunee ph
=
..Ecs: Krok
.. w .
Green Ecy(K-16) K '- .- K-15 N 26 Miles . (K-28) K-24 * '. 'r z. A NE
} MICHIGAN h ,'
l .l
-2 m" Stengetville x[
J K-5* K-21 K-4.',) W _ z b ~ ' o
$ 9.o S K-19
\
/
O'/ / K-Il . K-10
-- b '
K-17 E D* - y,'
*z g K-27 //
g ' , u K-23/! SITE m.' - 6 K-1 j , ,. i Kewaunee Co. K-22M W / 2
/1 Jisch Milk.\gw-25 AK K-12 Mcnitowoc Co. M V V . /-
K-6 K-8 . K-7 .:. p CT g,i y *, K-14 . i d K-13 N l 1 sh g ,1,. . \ MK-18 (K 26) k2 ' ' ' t0.7tniSSW . ;.
.i.
/ . , . .
O m O SCALE IN MILES I 2 3 4 Figure 4-1. Samoling locations, Kewaunee Nuclear Power FT wr 22
m.
~ ~
4 ,
- I Y e
.Q Tab l e 4.1. - Sampling locations, Kewaunee Nuclear Power Plant.
. Distance:
.(miles)b' and
- , Code Typea - Sector Location K-1 -Onsite 1
la. I 0.62 N North Creek 5 .lb I 0.12 N Middle Creek .
.lc IL 0.10 N 500' north of condenser discharge ld I 0.10 E Condenser discharge le I 0.12 S South Creek.
If I 0.12 S Meteorological tower Ig- :I 0.06 W South Well North Well lh I 0.12 NW lj I 0.10 S 500' south of condenser . discharge K-2 C 9.5 NNE WPS Operations building in Kewaunee i - K C 6.0 N Lyle and. John Siegmund farm Route 1, Kewaunee K-4. I 3.0 N Dan Stangel farm, Route 1, Kewaunee K-5 I 3.5 NNW Ed Papiham farm, Route 1, Kewaunee [- K-6 C 6.7 WSW Novitsky farm K-7 I 2.75 SSW' Earl Bruenner f arm, Route 3,. Two Rivers 2 > K-8 _C . 5.0 WSW Saint Mary's Church, Tisch Mills
- K-9 C- .11.5 NNE Rostok Water Intake for Green Bay, Wisconsin two miles north of Kewaunee K-10 -I 1.5 NNE Turner farm, Kewaunee' site
- K-11 I 1.0 NW Harlan Ihlenfeld farm K-12 I 1. 5 WSW Lecaptain farm, one mile west of site K-13 C 3.0 SSW Rand's general store
, K-14 I 2.5 S Two Creeks Park, 2.5 miles south of site K-15 C 9.25 NW Gas Substation, 1.5 miles north of Stangelville
- i. K-16 C' 26 NW WPS Division Office Building, Green Bay, Wisconsin K-17 I 4.25 W Jansky farm, Route 1, Kewaunee
, K-18 C 7. 0 SSW Schmidt's Food Stand, Route 163 (3.5 miles south of"BB") + K-19 I 1.75 NNE Wayne Paral farm, Route 1, Kewaunee K I 2.5 N Carl Struck farm, Route 1, Kewaunee K-23 I 0.5 W 0.5 miles west of plant, Kewaunee' Site j K-24 C 5.45 N Fectum farm, Route 1, Kewaunee i f K-25 I 2.75 WSW Wotachek f arm, Route 1, Denmark ? K-26c- C 10.7 SSW Bertler's Fruit Stand (8.0 miles south of "BB") K-27 I 1. 5 NW Schlies Farm 0.5 miles west of K-11 K-28 C 26 NW Hansen Dairy, Green Bay, Wisconsin j b I = indicator; C = control p Distances are measured from reactor stack. c Location K-18 was changed because Schmidt's Food Stand went out of business i d and was replaced by Bertler's Fruit Stand (K-26). 23 f i
... _.- - _ ~ , . - _ . _ . .__._._._m . - , , _ . _ _ , . . . _ _ _ _ . _-,.m _
p, (m) v a
%/
-(.3 wl; i
Table 4.2. Type and frequency of collection. Frequency Location Weekly Biweekly Monthly Quarterly Semiannually Annually K-la SW SL' K-lb SW GRa SL- - K-Ic BSb K-Id SW Fla BSb S L' - K-le SW SL K-lf AP AI GRa TLD SO TLD K-Ig WW K-lh WW K-lj BSb K-2 AP AI TLD TLD K-3 MIc sta TLD CFd SO TLD K-4 MIc g<a TLD CFd SO TLD K-5 Mic gta TLD CFd S0 TLD K-6 Mic gga TLD CFd S0 TLD K-7 AP Al TLD TLD m K-8 AP AI TLD TLD
- K-9 SW BSD SL K-10 WW-K-Il PR WW K-12 MIC GRa CFd wg so K-13 WW -
K-14 SW BSb SL K-15 AP Al TLD TLD K-16 AP Al TLD TLD K-17 TLD TLD VE K-18e VE K-19 Mic GRa CFd 50 K-20 DM K-23 GRN K-24 DM K-25 DM K-26 VE K-27 TLD EG TLD DM ' K-28 MIC
- Three times a year, second (April, May, June), third (July, August, September), and fourth _(October, November, December) quarters.
b To be collected in May and November. c Monthly from November through April; semimonthly from May through October. 4 First (January, February, March) quarter only.
" Replaced by K-26 in summer of 1982.
. =_. _ . .-_. . . .
- 3 L-Table 4.3.~ ' Sample codes used in-Table 4.2.
~ Code Description AP Airborne Particulate 4
AI Airborne Iodine TLD Thermoluminescent Dosimeter 1 PR Precipitation MI Milk WW Well Water DM Domestic Meat l EG Eggs VE Vegetables GRN Grain GR Grass CF Cattlefeed
- SO Soil r
! SW Surface Water FI Fish SL Slime BS Bottom Sediments l. i l lO 25 i e n, - - - - ~ - - , , - - - , - - - - , - - . _ ,en.---- em n, ,,_.. , nn,_n,..,,,,,,---nww-~,,, .m.--,,,-,w--,-r-
T
\ . .d - Table 4.4. Sampling summary, January - December 1986.
Collection Number of Number of Sample Type and Number of Samples Samples Type Frequencya Locations Collected Missed Air Environment Airborne particulates C/W 6 312 0 Airborne iodine C/BW 6 156 0 TLD's C/Q 12 48 0 Precipitation C/M 1 12 0 Terrestrial Environment Milk (May-Oct) G/W 7 84 0 (Nov-Apr) G/M 7 42 0 Well water G/Q 6 24 0 Domestic meat G/A 4 4 0 p Eggs' G/Q 1 4 0 Vegetables - 5 varieties G/A 2 7 0 Grain - oats G/A 1 1 0
- wheat G/A 1 1 0 Grass G/TA 8 24 0 Cattle Feed G/A 6 6 0 Soil G/SA 7 14 0 Aquatic Environment Surface water G/M 6 72 0 Fish G/TA 1 5 0 Slime G/SA 6 12 0 Bottom sediments G/SA 5 10 0 a
- Type of collection is coded as follows: C = continuous; G = grab.
Frequency is coded as follows: W = weekly; M = monthly; Q = quarterly; SA = semi-annually; TA = three times per year; FA = four times per year;
- A = annually; BW = bi-weekly.
O 26
(.
, , - -y
/ \ [ I '
, N._/
Taole 4.5 Environmental Rasiological Monitoring Program Sumary, same of Facility Kewaunee Nuclear Power Plant Docket No. 50-305 Lccation of Facility Kewaun ee Coucty. wisconsin Reporting Pertos January - Decemoer 1986 gaunty, state) Incicator Location witn Hignest Control Samole Type and Locations Annual Mean Locations Nencer of Type Numoer of Mean (F)C Hean (F) Mean (F) Non-routine (Units) Analysesa LLDb RangeC Locationd Hange Range , nesults e uroorne GS 312 0.005 0.031 (99/104) K-7 Eruemer Farm 0.032 (51/52) 0.028 (200/208) l 3 particulates (0.006-0.240) 2.75 mi 55. (0.008-0.240) (0.004-0.271) i (pct /ra3) G5 24 Be-7 0.022 0.063 (8/8) K-2, kP5 Op. Bu11- 0.076 (4/4) 0.068 (16/16) 0 (0.031-0.12) ding, 9.5 mi ANE (0.056-0.095) (0.052-0.095) ND-95 0.0051 <tLD - <LLD u Zr-95 0.0045 <LLD - - <LLO O N RJ-103 0.0037 0.0091 (2/8) K-2, =PS Op. 8utt-cing, 9.5 mt NNE 0.014 (1/4) 0.0085 (4/16) 0 (0.009u-0.0091) - (0.0019-0.014) Ra-106 0.013 (LLD - - <LLO O Cs-134 0.007 0.0057 (2/8) K-2, WPS Op. Bu11- 0.0071 (1/4) 0.0058 (4/16) 0 (0.0055-0.0057) ding, 9.5 mi hNE - (0.0045-0.0071) Cs-137 0.0006 0.012 (2/8) K-16, Green Bay 0.016 (1/4) 0.013 (4/16) 6 i (0.010-0.013) 26 mi NW - (0.010-0.016) l Ce-141 0.0058 <tLD - - <LLD 0 Ce-144 0.0070 <tLD - - (LLD 0 Airoorr.e 1-131 156 0.03 0 42 (4/52) K-15, Gas Substa- 0.50 (2/2e) 0.36 (10/104) 14 Iodine (0.08-0.73) tion, 9.25 mi NW (0.27-0.74) (0.12-0.81) (pC1/m3) TLD -Quarterly Gama 43 1 16.0 (24/24) K-17, Jansky Farm 18.6 (4/4) 14.8 (24/24) 0 . (mR/91 days) (11.2-23.1) 4.25 mi W (14.7-23.1) (10.4-21.9) 0
m--;73
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isole 4.5 Environmental Raoiclogical Fonttoring Program Smeary (contin.4ec) same of Faciltty tewac ee Actear Power plant- Docnet No. 50-305-Location of facility newaanee sounty, .isconsin deporting Perico January - uecemoer 19de , N unty staie; Indicator Location witn Hignest Control Sample Type and Locations Annual Mean Locations mmoer ofc Tspe Mumper of Mean (F)C Mean V ) Mean (F)~ non-routine. An alyses3 LLDD RangeC-' Locationd Range ' Range -Aesults' ' (Units) Precipitation H-3 12 330 <tLD - -- .None 0 (pci/1) mit 1-131 126 0.5 11.7 (15/72) K-12. Lecaptain Fara 28.8 (3/18) 2.3 (10/54) 25 (pCi/1) ( 0. 6-61.5) 1.5 at WSW (5.6-61.5) (0.8-5.2) Sr-89 84 1.5 <LL3 - - <LLD 0 Sr-90 S4 0.5 2.4 (43/48) K-12. Lecaptain Fars 3.3 (12/12) 2.5 (36/36) 0-(1.3-5.4) 1.5 at wSd (2.0-4.5) (0.8-4.3) GS 126 K-40 50 1360 (72/72) ~ K-3. Stangel Fara ~ 1400 (IS/la) 1340 (54/54) 0 - (1110-1840) 3.0 mi N (1180-1590) (1150-1600) l Cs-137 10 11.1 (1/72) K-12. Lecaptain Farm 11.1 (1/18) ALO - :0 S$ 1.5 at ush -
<LL3 - - <LLD 0 84-140 10 (g/1) K-st able 84 1.0 1.55 (4S/49) K-3. Stangel Fars 1.60 (12/12) 1.52 (36/36) ,0 (1.26-2.01) 3.0 mi N (1.40-1.77) (1.31-1.82) 0.5 K-6. Movitsky Fara 0.% (12/12) 0.92 (36/36) 0 (g/1) Ca 84 0.92(48/48)
(0.79-1.14) 6.7 at WSW (0.85-1.09)- (0.47-1.25) 3.0 - (LLD 0
= ell dater GA 8 <LLO -
(pct /1) 1.2(3/4) 0 GB 24 2.9 2.5 (12/20) K-19. Soutn Well 3.4 (3/4) (1.3-4.6) thsite, 0.06 at W (2.8-3.8) (1.0-1.4)
' K-In, North hell 3.4 (3/4).
Onsite. 0.12 mi NW (2.3-4.6) H-3 4 330 <LLD
- - None 'O K-40 24 0.10 1.96 (20/20) K-Ig. Soutn dell 2.68 (4/4) 1.29 (4/4)' O (flame) (1.13-3.34) .Onsite, 0.06 at W (2.04-3.34), (1.17-1.49) 0
- None Sr-89 4 1.0 <LLD -
~~
- None 0 Sr-90 4 0.5 <tLD -
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Table 4.5 Environmental dadiological Monitoring Program Sumary (continued) Name of Fact 11ty Ke aunee Nuclear Power Plant Docket No. 50-305 Location of Facility Kewawnee county, misconsin Reporting Perica January - Decemoer 1986 (County. State) Indicator Location witn Hignest Control Sample Type and Locations Annual Mean Locations Numoer of Type Number of Mean UJ Mean (F) Non-routine (Units) Analysesa LLDb Mean dange {F)c Locationd Range Range Resultse Well Water GS 24 (pC1/l) (continued) Mn-54 15 RLD - - <LLD 0 Fe-59 30 4LD - - <LLD 0 Co-Sa 15 <LLD - - (LLD 0 Co-60 15 <LLO - - <LLD 0 Zr-Nb-95 15 ELD - - <LLD 0 Cs-134 10 4LD - - <tLD 0 ro
- Cs-137 10 <LLD - - <LLD 0 Ba-La-140 15 <LLD - - <LLO O Domestic Meat GA 4 0.04 0.11 (3/3) K-27. Scnlies Farm 0.15(1/1) 0.13 (1/1) 0 (chickens) (0.07-0.15) 1.5 mi NW - -
9 "' GB 4 0.03 1.84 (3/3) K-27. Scnites Fara 2.29 (1/1) 2.22 (1/1) 0 (1.33-2.29) 1.5 mi NW - - GS 4 Be-7 1.29 4LD - - <LLD 0 K-40 0.5 3.11 (3/3) K-24. Fectum Farm 3.59 (1/1) 3.59 (1/1) 0 (3.20-3.38) 5.45 mi N - - Nb-95 0.13 GLD - - <LLD 0 Zr-95 0.23 4LD - - RLD 0 Ru-103 0.28 <LLO - - <tLD 0 Ru-106 0.38 4LD - - ALD 0-Cs-134 0.039 4LD - - <LLD 0 Cs-137 0.039 <LLD - - ALO O Ce-141 0.62 4LD - - 4LD 0 Ce-144 0.29 4LD - - (LLD 0
, f ,
Table 4.5 Environmental Raatological Mcnitoring Program Summary (continueo) . Name of Factitty Ke aunee Nuclear Power Plant Docket No. 50-305 Location of Factitty aesaunee County, sisconsin Reporting Perloo January - Decemoer 1986 (County, State) Indicator Location =tth Hignest Control l-Samole Type and Locations Annual Mean Locations Namoer of. - . Type Nuncer of Mean (F)c Mean (F) Mean(F)' en-routine , (Units) Analysesa LLDO Rangec Locationd Range Range ' Results* 2 Eggs G8 4 0.01 0.99 (4/4) K-27, Schlies Fare 0.99 (4/4) None 0- '!* (pct /g met) (0.69-1.17) 1.5 at W (0.69-1.17)
- Sr-89 4 0.004 4LD - - None =0 ,
t Sr-90 4 0.002 RLD - - None 0. j GS 4 , l Se-7 0.19- ALD - - None 0 i K-40 K-27, Scnites Fars 0.01 1.23 (4/4) 1.23(4/4) None 0 j (1.04-1.41) 1.5 at W (1.04-1.41) ; ! I w 2-95 0.019 RLD - - None 0 O i Zr-95 0.039 RLD - - None O r. Ru-103 0.038 <LLD - - None 0 Ru-106 0.098 RLD - - None .O Cs-134 0.011 ELD - - None 0 { Cs-137 0.011 4LD - - None 0 ~! I-Ce-141 0.057 4LD - - None 0 Ce-144 0.067 . CLD - - None 0 kegetaoles GB 7 0.1 2.16 (2/2) K-26, Bertler's Frutt 2.45 (5/5) 2.45 (5/5) 0 (pC1/g wet) (2.02-2.30) Stand, 10.7 mi SSW .(1.26-4.37) (1.26-4.37) Sr-89 7 0.0038 CLD K-26, Sertler's Frait 0.0091 (1/1) 0.0091(1/5) O Stand, 10.7 mi SS.d - - Sr-90 7 0.0013 RLD K-26, Bertier's Fruit 0.0021 (4/5) 0.0021 (4/5) 0 1
- Stand,10.7 at 55. .(0.0015-0.0026) (0.0015-0.0026) !
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Taole 4.5 invirorcental dadiological Monitoring Program Summary (continued) Nrne of Facility Kewaunee Nuclear Power Plant Occtet No. 50-305 Locatien of Facility sewaunee county, .isconsin Reporting Perioo January - vecemoer 19c6 (Co nty, 5 tate) Inaicator Location witn Hignest Control Sample Type and Locations Annual Mean Locations Numoer of Type Numoer of Ptean p ) Mean (F) Non-routine (units) Analysesa LLDb Mean{F)C Range Locationd Range Range Wesultse tvegetaoles GS 7 (pct /g wet) Be-7 0.097 <LLD - - <tLD 0 (continued) K-40 0.75 1.98 (2/2) K-26, Bertler's fruit 2.14 (5/5) 2.14 (5/5) 0 (1.59-2.37) Stand,10.7 mi SSW (1.11-3.03) (1.11-3.03]
!~
ND-95 0.012 <tLD - - <LLD 0 Zr-95 0.020 <tLD - - (LLD 0 0.013 <LL3 - <LLD 0 Ru-103 - Ru-106 0.11 <tLD - - (LLD 0 f - <LLO O Cs-137 0.013 <LLD - 0.021 <LLD - - <LLD 0 w Ce-141 w <LLD 0
. Ce-144 0.093 4LD - -
K-23, Kewaunee Site 6.48 (2/2) None U Grain - Dats ' GB 2 0.1 6.43 (2/2) (pC1/g wet) (5.25-7.70) 0.5 mi W (5.25-7.70) 0.04 CLD - None O Sr-89 2 - Sr-90 2 0.002 0.008 (2/2) K-23, Kewaunee Site 0.008 (2/2) None 0 (0.008-0.009] 0.5 at W (0.008-0.009l GS 2 K-23. Kewaunee Site 1.60 (2/2) None U i Be-7 0.2 1.60 (2/2) (1.13-2.08) 0.5 mi W (1.13-2.08) K-23, Kewaunee Site 4.77 (2/2) None 0 K-40 0.1 4.77 (2/2) (4.40-5.14) 0.5 mi W (4.40-5.14)
<tLD - None 0 Nb-95 0.017 -
0.029 - None .O Zr-95 <LLD -
- None O du-103 0.016 <LLD -
- None O Ru-106 0.15 <tLD -
<tLD - None 0 Cs-137 0.016 -
None O I Ce-141 0.022 4LD - - RLD - None 0 Ce-144 0.095 -
7 ~ p
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Tamle 4.5 Environmental Radiological Moottoring Progra:n Summary (continued) name of Facility Kewaunee Nuclear Power Plant Docket 4o. 50-305 Location of Factitty (e.aunee county, .isconsin Reporting Perica January - uecemoer 1986 (County, 5 tate) Indicator Location witn tilgnest Control Sample Type and Locations Annual Mean Locations Numoer of Type Numoer of Mean (F)C Mean (F) Mean (F) Non-routine (Units) Analysesa LLDo RangeC Locationd Range Range Results* Tattlefeed GB 6 0.1 7.5 (5/5) K-4, Stangel Fara 12.1 (1/1) 10.3 (1/1) 0 (pct /g wet) (2.9-11.9) 3.0 mi N - - Sr-89 6 0.007 , <LLD - - <LLD 0
! K-4, Stangel Farm Sr-90 6 0.01 0.064 (5/5) , 0.11 (1/1) 0.096 (1/1) 0 (0.030-0.11) ' 3.0 mi N - -
GS 6 Be-? 0.15 0.39 (5/5) K-4, Stangel Farm 0.71 (1/1) 0.13(1/1) 0 (0.18-0.71) 3.0 mi N - K-40 1.0 6.55 (5/5) K-4, Stangel Farm 11.54 (1/1) 10.10 (1/1) 0 (2.ti5-11.54) 3.0 mi N - - Nb-95 0.012 CLD - - <LLD 0 ta N Zr-95 0.021 <LLD - - <LLD 0 Ru-103 0.010 <LLD - - <LLD 0 RJ-106 0.09] <LLD - - <LLD 0 Cs-134 0.012 <LLD - - <LLD 0 Cs-137 0.011 <LLD - - <LLD 0 Ce-141 0.020 <tLD - - <LLD 0 Ce-144 0.082 <LLD - - <LLD 0 Grass GS 24 0.1 6.75 (18/18) K-If, Meteorological 7.80 (3/3) 6.73 (6/6) 0 j (DCi/g wet) (4.46-9.62) Tower On Site, (7.17-8.83) (5.79-7.74) j 0.12 mi 5 Sr-89 24 0.13 <LLD - - <LLD 0 . Sr-90 24 0.010 0.016 (7/18) K-3, Stangel Farm 0.026 (2/3) 0.020 (4/6) 0 l (0.010-0.021) 3.0 mi N (0.022-0.029) (0.012-0.029) l G5 24 Be-7 0.53 2.23 (12/18) K-lb, Middle Creek 3.12 (3/3) 1.28 (4/6) 0 (0.26-5.74) 01 site, 0.12 mi N (0.26-5.74) (0.79-2.18) K-40 0.1 6.02 (18/18) K-5, Pap 1 nam Fara 6.60 (3/3) 6.33 (6/6) 0 (3.65-8.37) 3.5 mi NhW (4.47-8.37) (4.70-7.92) Nb-95 0.094 <LLD - - <LLD 0
~ ,., ,~~. -
I f w/ J Tacle 4.5 Environmental Radiological Monitoring Program Summary (continuec) name of Facility Kewaunee Nuclear Po.ee Plant Docket ho. 50-305 Location of Facility nei,4anee County, wisconsin Reporting Perloc January - Decemoer 1980 (County, 5 tate) Indicator Location witn Hignest Control Sample Type ed Locations Annual mean Locations Numoer of Type Ommer of Mean (F)C Mean (f) Mean (F) Non-rout ine (Units) Analysesa LLDb RangeC Locationd Range , Nedge Results8' Grass Ir-95 0.093 <LLD - - <LLD 0 (pCi/g wet) (continued) Ru-103 0.072 0.070 (2/18) K-If, Meteorlogical 0.0077 (1/3) <LLO O (0.064-u.077) Tower on Site - 0.12 mi S Ru-106 0.31 <LLD - - <LLD 0 Cs-134 0.040 0.043 (2/18) K-1f, Meteorological 0.054 (1/3) <LLO O (0.043-0.054) Tower On Site - 0.12 mi 5 Cs-137 0.033 0.066 (4/18) K-1f, Meteorological 0.082 (1/3) 0.026 (2/6) 0 (0.030-0.082) Tower On Site - (0.018-0.034) 0.12 mi S w Ce-141 0.18 <tLD - - <LLD 0 w Ce-144 0.29 <LLD - - <LLD 0 Soil GA 14 3.0 7.1 (8/10) K-5, Papinam Farm 10.0 (2/2) 5.1 (4/4) 0 (pC1/g cry) (3.9-11.4) 3.5 mi WSd (8.5-11.4) (3.5-7.2) GS 14 2. 0 18.7 (10/10) K-12 Lecaptain 24.4 (2/2) 20.8 (4/4) 0 (6.8-27.7) Farm, 1.5 at WS4 (21.1-27.7) (15.6-27.2) Sr-89 14 0.056 RLD - - <tLD 0 Sr-90 14 C.01 0.057 (10/10) K-6, Novitsky Farm 0.132 (2/2) 0.089 (4/4) 0 (0.016-0.138) 6.7 mi wSd (0.093-0.171) (0.031-0.171) GS 14 Be-7 0.45 <tLD - - <LLu O K-40 t 1.4 15.3 (10/10) K-12, Lecaptain 19.3 (2/2) 14.9 (4/4) 0 (6.57-20.10) Farm, 1.5 mi WSW (18.4-20.1) (12.5-18.6) No-95 0.067 4LD - - <LLO 0 Zr-95 0.037 <LLD - - <LLu 0 Ra-103 0.073 <tLD - - <LLO O Au-106 0.30 <tLD - - <tLD 0
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q,7 l h ) , - ) Taole 4.5 Environmental dad:ological Monitoring Program Sumary (continuec) . Name of Facility (e.aunee Nuclear Po er Plant Docaet No. 50-305 Location of Factlity tenaunce County, nisconsin deporting Perica January - Decemcer Ivco p;ounty, state) Incicator Location wita Hignest Control Sample Type and Locations Annual '9ean Locations Numoer of Type Nunoer of Mean (F)C Mean (F) Mean (F)~ hon-raat ine (Units) Analysesa LLuD dangeC Location 4 Range Range ' desults* Soil Cs-137 0.01 0.36 (10/1G) K-5. Papinam Farm 0.90 (2/2) 0.47(4/4) 0 (pCi/g cry) (0.028-1.59) 3.5 mi h w (0.22-1.59) (0.23-0.71) (continuea) ' Ce-141 , 0.14 4LD - - <Ltu 0 Ce-144 0.20 <LLO - - <LLu o Surf ace inater G(SS) 84 i 0.8 2.1 (1/72) K-la, Conoenser 2.1 (1/12) (LLU 0 (pCi/1) , - Discnarge. Onsite - l 0.10 mi E Gts(D5) 84 3.4 4.1 (71/ 72) K-la. Nortn Creer, 7.9 (12/12) 2.1 (12/12) 0 (1.7-18.0) Onsite, 0.62 mi N (2. 7-18.0) (1.3-2.5) GS(TA) 84 4.0 4.1 (71/72) K-la, Nortn Creek, 7.9 (12/12) 2.1 (12/12) 0 (1.7-18.0) Onsite, 0.62 mi N (2.7-18.0) (1.3-2.5) w G5 84 4 Mn-54 10.5 <LLD - - <LLu u Fe-59 19.8 (LLu - - <LLu U Co-53 9.2 <tLD - - <LLu O Co-60 10.6 4LD - - <LLu O Zr-hb-95 11.2 RLD - - <LLu 0 Cs-134 10.1 ELD - - <LLu o Cs-137 9.6 <LLD - - <LLO O l da-La-140 15.0f <tLD - - <LLD 0 H-3 28 330 1160 (4/24) K-le, Soutn Creek 292u (1/4) <LLO O , (450-2920) Onstte, 0.12 at 5 - 1 Sr-89 23 1. 8 (LLD - - <LLu O Sr-90 23 0.9 1.1 (3/28) K-Id Concenser 1.4 (1/4) <tL0 o (u.9-1.4) utscnarge, Onsite -
.l 0.10 at E K-40 84 0.5 3.30 (72/72) K-la, Nortn Creek 8.17 (12/12) 1.51 (12/12) 0 (flame) (1.21-19.71) Onsite, 0.62 mi N (2.00-19.71) (1.2u-1.80)
I
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l [ Taole 4.5 Environmental Radiological Monttoring Program Summary (continued) " Name of Facility Ke aanee Nuclear Power Plant Docket No.- 50-305 Location of Factlity ex.a. nee t.ounty. .isconsin Reporting Persoo January - uecemoer 19u6 (Loanty, State) Incicator Location witn Hignest Control Sample Type and Locations ~ kaner of' Annual Mean Locations Type huser of Mean (F) Mean (F) Non-rout ine Analysesa ttge Mean(FjC (Units) Range Locationd Range Range .Resalts* Fisa - Muscle GB 5 1.0 2.3 (5/5) K-Id. Concenser 2.3 (5/5) None' 0' (pCi/g wet) '(1.4-7.1) Disenarge Onsite (1.4-7.1) 0.10 mi E G5 5 sei-54 0.033 RLD - - None 0
~
j Fe-59 0.20 4LD - - hone 0 Co-58 0.056 <LLD - - None O Co-60 0.030 <LLO - - None 0
$ Cs-134 0.032 ,
Cs-137 0.027 0.11 (4/5) K-id, Concenser Dis- 0.11 (4/5) hone 0 (0.095-0.13) enarge, Onsite -(0.095-0.13) 0.10 mi E Fisa - Bones (pCilg met) GB 5 1.1 3.1 (4/5) K-id, Condenser Dis- 3.1 (4/5) None 0 (1.1-6.9) cnarge. Disite (1,1-6.9) 0.10 mi E - Sr-89 5 0.% <LLD - - None 0 Sr-90 5 0.10 0.25 (5/5) K-1d, Concenser Dis- 0.25 (5/5) None 0 (0.04-0.40) cnarge, Onsite (0.04-0.40) 0.10 mi E Pertonyton GB 12 0.1 3.86 (10/10) K-14, Two Creeks 4.6 (2/2) 2.30 (2/2) 0 ,i l (Slime) (1.84-5.08) Park, 2.5 mi 5 (4.48 -4. 72) (1.49-3.11) (pci/g wet) l Sr-89 12 0.061 <LLD - -
. <LLD 0 t
Sr-90 12 0.0C9 0.053(10/10) K-le, Soutn Creek 0.112 (2/2) 0.039 (1/2) 0-(0.000-0.185) Onstte, 0.12 mi 5 (0.038-0.185)
+ ,
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(v) (s _-) ~ Taole 4.5 Environmental Radiological Mcnitoring Progra:n Summary (continueo) same of Facility Kewaanee Nuclear Power Plant Doctet No. 50-305 Location of Facility Kews nee County, nisconsin Reporting Persoo Jar:uar y - Decemoer 19eo - (County, 5 tate) Indicator Location uitn Hignest. Control-Sam le Type and Locations Annual Mean Locations Numoer of - Type Necer of Mean (F)C Mean (F) Mean (F) hon-rout ine (Units) Analysesa Ltob RangeC Locatiend Range Range Resultse Perionyton GS 12
.(Sline) f(pct /gmet) Ee-7 0.78 0.83 (4/10) K-le, South Creek 1.21 (1/2) 1.05(1/2) 0 3(continueo) (0.52-1.21) 5 - -
! l Onsite, 0.12 lii i K-40 0.50 2.92 (10/10) K-lo, Mid:le Creek 3.66 (2/2) 1.60 (2/2) 0 (1.18-4.66) Disite, 0.12 mi N (J.40-3.91) ~(1.07-2.14)
Mn-54 0.C36 CLD - - ELG ..; O Co-53 0.CS9 . 0.116 G/10) K-lo, Concenser Dis- 0.116 (2/2) . <Ltu e 1 (0.071-0.15) ' cnarge Onsite, (0.04-0.16) . ca O.10 E ,.
- K-le, Concenser Dis- 0.045 (i;J) 4LD 0 Co-60 0.049 0.074 (2/10) ~
(0.063-0.085) charge Onsite, - 0.10 E m-95 0.065 CLD - - ~ <LLD 0
..Zr-95 0.14 RLD - t <LLD 0 l ! '
+ ' Ru-103 OJO , _ CLD- - - CLD 0-l .
~
l :
.j Ru-106 0.42 '4LD - - (LLD 0 l% -
6 . Cs-1M 0.083 CO - - (LLO O Cs-137 0.081 0.035(7/10) K-9, 8tostot Water 0.059'(1/2) ' 0.059 (1/2) 0 (C.025-0.045) Inta e. 11.5 hMf. - - e. Ce-141 0.13 <t 8. 0 - -
'LLD a' O . -
I
- C+-141 0.40 Ew ,,
- - ALD 0 I
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Taole 4.5 Environmental 44&fo ical Manitortng Program Sunswf (continuec) _ - l ' !~
~
r,4:e of Facility me unce Octear Po=er Plant Doccet No. 50-305 c: - Location of fact 11ty _ ae.aucee county. =isconsia deporting Persoo January - acemoer M r, _ tLounty, 5tates , -: u. : ~ _ s j l Indicator Location eitn Hignest b An cl < . Numpef of l Samp't .- Type ana -~ Locations Annual stean i Locittons .. . .a Type - ' k w of hean (F)C .*wn e j Mean (F)~ Non-routine ;,f'f g (units) Aaanyses a uten gange c Location 4, Rare mange' desultse ,
~
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! 'Sotton GB 10 1.0 8.1 (8/8) E-14. Two Creets 9.2 (2/2) 7.2 i2/2) - ~0 M WG h
- 6 Sentrats ! (5.1-10.3) P m , 2.5 mi 5 '
'ef?.1-9.2) . (6.5-7.9)' "i ~c' (p~tig ory) _
g j - 1 l fr-89 10 40.040 <LLD - - <LLD . 0- ' Sr-90 10 0.015 o LD - - <tLD 0 GS 10 K-40 5.06 (8/8) K-14. Two Creets. 5. M (2/2) 5.23 (2/2) -0 l1.0 (3.96-5.97) Part 2.5 mi S- (5.90-5.9?) (4.83-5.68) l Co-58 0.027 0.115 (5/8) K-lj, 500* 5 of Con- 0.26 (1/2) (LLO O-w l (0.021-0.26) censer Disenarge. - N ; Onsite, 0.10 at S Co-60 0.019 0.076 (6/8) K-1J. 50u' S of Cdn ' O.11 (2/2) <LLD 0' (0.039-0.14] denser Discnarge. (0.068-0.092) Onstte, 0.10 et S Cs-134 0.023 0.077 (1/8) K-lj, 500' S of Con- 0.077 (1/8) (LLD. 0 . denser Disenarge. - Onstte, 0.10 mi 5 Cs-137 0.005 0.042 (7/8) K-lj, 500' 5 of Con- 0.056 i/2) 0.020 (2/2) 0 denser Discharge . (0.041 J72)~ (0.014-0.027) l(0.016-0.072)Onstte, 0.10 mi 5
- GA = gross 41pna, 68 - gross neta, GS = gamma spectroscopy, SS = suspenoeo solids DS = dissolved solids TR =. total residue.
LLD = nominal lo er Itait of detection Daseo on 3 sigma counting error for Dactgrouno sample. _ g $ lean base.J upon detectan1e measurements only. Fraction of cetectable measurements at spectf ted locations ts indicateo in d E##'"*"**** IO*
, Locations are specified by station code (Taole 4.1), distance (atles) and direction relative to reactor site.
Non-routine results are those unicn exceed ten times tne control station value. If no control station value is available, the
, result is considered non-routine if it exceeds ten times tne pre-operational value for.tne. location.
Eight results (<15.1, <15.1, <15.3, (16.2 <17.8, (18.5, (25.2 ed <25.6 pct /1) have neen excluded from the determination of LLD. - Tney resulted from a delay in counting. l l
i
,m b
5.0 REFERENCES
4 l Arnold, J. R. and H. A. Al-Salih. 1955. Beryllium-7 produced by cosmic - rays. Science 121: 451-453. Eisenbud, M. 1963. Environmental Radioactivity, McGraw-Hill, New York, New York, pp. 213, 275, and 276. Gold, S., H. W. Barkhau, B. Shlein, and B. Kann, 1964. Measurement of Naturally Occurring Radionuclides in Air, in the Natural Radiation Environment, University of Chicago Press, Chicago, Illinois, 369-382. ! Hazleton Environmental Sciences,1979. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report - Part II, Data Tabulations and Analysis, January - December e 1978. (
\
r
. 1980. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report -
! Part II, Data Tabulations and Analysis, January - December 1979. f- ! . 1981. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report - ' Part II, Data Tabulations and Analysis, January - December 1980. ,.
. 1982. Annual Report. Radiological Monitoring Program for f the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report -
;; Part II, Data Tabulations and Analysis, January - December 1981.
. . 1983. Annual Report. Radiological Monitoring Program for
- d. the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report -
f Part II, Data Tabulations and Analysis, January - December 1982. Industrial BIO-TEST Laboratories, Inc.1974. Annual Report. Pre-operational Radiological Monitoring Program for tne Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin. January - December 1973.
. 1975. Semi-annual Report. Radiological Monitoring Program
. for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin. January -
' June 1975.
NALC0 Environmental Sciences. 1977. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, January - December 1976. 38
p.
. ,m O . 1978. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power -Plant, Kewaunee, Wisconsin, Final Report - Part II, Data Tabulations and Analysis, January - December 1977.
National Center for Radiological Health. 1968. Section 1. Milk surveillance. Radiological Health Data Rep., December 9:730-746. National . Council on Radiation Protection and Measurements. 1975. Natural Radiation Background in the United States. NCRP Report No. 45. J l ! Solon, L. R., W. M. Lorder, A. Shambron, and H. Blatz. 1960. Investigations of Natural Environmental Radiation. Science. 131: 903-906. Teledyne Isotopes Midwest Laboratory. 1984. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II, Data Tabulations and Analysis, g January - December 1983.
. 1985. Annual Report. . Radiological Monitoring Program for the l Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II, Data Tabulations and Analysis, January - December 1984.
. 1986. Annual Report. Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II,
(~} V Data Tabulations and Analysis, January - December 1985.
. 1987. Annual Report. Raoiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II, Data Tabulations and Analysis, January - December 1986.
l U.S.S.R. State Committee on the Utilization of Atomic Energy. The Accident at the Chernobyl Nuclear Power Plant and its Consequences. Information compiled for the I.A.E.A. Experts' Meeting, 25 - 29 August 1986, Vienna, Austria. Wilson, D. W., G. M. Ward, and J. E. Johnson,1969 In Environmental Contamina-tion by Radioactive Materials, International Atomic Energy Agency, p. 125 O 39
E
'O a
l Appendix A Interlaboratory Comparison Program Results O O l A-1 _ . . _ . . . . . . _ _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . . , _ _ _ . . _ _ _ . . . _ . . _ . ~ _ _ , _ _ . . _ _ _ _ _ . . . _ _ . ._
n .
- q. .
Appendix A h ,- Interlaboratory-Comparison Program Results Teledyne Isotopes Midwest Laboratory (formerly Hazleton Environmental Sciences) has participated in interlaboratory comparison (crosscheck)-programs .since the-
- iforruinion of its - quality control- program in December 1971. These programs.-
. ' are operated by agencies which supply environmental-type' samples (e.g.', milk or.. -r water) containing concentrations of radionuclides known to the -issuing agency but not to participant laboratories. The purpose' of such a program is to '
- provide an independent check on the laboratory's analytical procedures and to alert it to any'possible problems.
Participant laboratories measure the concentrations of. specified radionuclides and L report them to ' the issuing agency. Several months later, the agency reports the known' values to the participant laboratories and specifies control limits. . Results consistently higher or lower than the known values.or outside
- the control limits indicate a need to check the instruments or - procedures.
used. The results in Table A-1 were obtained through, participation in the environ- ! imental sample crosseneck program for milk, water, air filters, and food - .
- samples -during the period May 1984 through _ November 1986. This program has a been conducted by the U. S. Environmental Protection Agency Intercomparison and'
' , . Calibration Section, . Quality Assurance Branch, Environmental Monitoring 'and Support Laboratory, Las Vegas, Nevada.
\
The results in Table A-2 were obtained for thermoluminescent dosimeters. (TLD's) during the period 1976, 1977, 1979, 1980, 1981,' 1984 and 1985-86 : through participation in the Second, Third, Fourth, Fifth,- Seventh and Eightn International Intercomparison of Environmental Dosimeters under the sponsor-ships listed in Table A-2. O w --v- -,- ---c,--- .~.,,->-.,,,,,,mevw., w,,.,v.+trew..,,,--,.mw-,w.w ,w, . - - , . - , . . , - , - - - .-,-v.,r,.-,.rv. -
.i,-r-.-,,.w.e
v y Table A-1. ' U.S. Environmental Protection Agency's- crosscheck program, com-- parison' of EPA and Teledyne Isotopes Midwest. Laboratory results
.for milk, water, air filters, and food samples, 1984 through 1986.a Concentration in )Ci/lb Lab Sample 'Date- TIML Result EPA lesult Code' Type Collected . Analysis i&c ilo , n=3d STW-358 Water May 1984 Gross alpha 3.0*0.6 3i8.7 Gross beta 6.7*1.2 618.7 STM-366 Milk June 1984 Sr-89 -2113.1 25t8.7 Sr-90 1312.0 17*2.6 I-131 4615.3 43*10.4 Cs-137 3814.0 35t8.7 K-40 1577*172 1496t130 STW-368 Water July 1984 Gross alpha 5.lil.1 6t8.7 Gross' beta 11.912.4 13i8.7 STW-369 Water August 1984 I-131 34.3*5.0- 34.0i10.4 STW-370 Water August 1984 H-3 30031253 28171617 STF-371 Food July 1984 Sr-89 22.0*5.3 25.018.7 Sr-90 14.713.1 20.012.6 I-131 <172 39.0*10.4 Cs-137 24.0i5.3 25.0*8.7 #
K-40 2503t132 26051226.0 STAF-372 Air August 1984 Gross alpha 15.3tl.2 17i8.7 Filter Gross beta 56.010.0 5118.7 Sr-90 14.311.2 1812.4 Cs-137 21.012.0 15*8.7 STW-375 Water Sept. 1984 Ra-226 5.li0.4 4.9*1.27 . Ra-228 2.210.1: 2.3i0.60 STW-377 Water Sept . 1984 Gross alpha 3.3tl.2 5.0*8.7 Gross beta 12.712.3 16.018.7 STW-379 Water Oct. 1984 H-3 28601312 28101356 STW-380 Water Oct. 1984 Cr-51 <36 40 8.7 Co-60 20.3 1.2 20i8.7 Zn-65 150t8.1 14718.7 Ru-106 <30 4718.7 Cs-134 31.317.0 3118.7 Cs-137 26.711.2 2418.7 i- A-3 i
.-n . . - , , , . .,,,,,-,-..,-n, ---,-,,,.,,.n, , ,,,,-,,,.,,,,,,,,,.,-,~..---_,,,,n. ,.,,,,--,.n. . , - , - . , - , , , ,,,..-.-r,-
*^
V Table A-1. (continued) Concentration in )Ci/lb Lab Sample Date TIML Result EPA tesult Code Type Collected Analysis ibc ilo, n=3d STM-382 Milk Oct. 1984 Sr-89 15.714.2 22i8.7 Sr-90 12.711.2 16i2.6 I-131 41.7*3.1 42110.4 l Cs-137 31.3i6.1 3218.7 K-40 1447166 1517i131 STW-384 Water Oct. 1984 Gross alpha 9.7*1.2 1418.7 (Blind) Sample A Ra-226 3.3 0.2 3.0f0.8 Ra-228 3.411.6 2.110.5 ! Uranium nae 5t10.4 Sample B Gross beta 48.3t5.0 6418.7 Sr-89 10.714.6 1118.7 Sr-90 7.311.2 12i2.6 Co-60 16.3tl.2 14i8.7 Cs-134 <2 2i8.7 Cs-137. 16.711.2 1418.7 STAF-387 Air Nov. 1984 Gross alpha 18.711.2 15t8.7 Filter Gross beta 59.0 5.3 52i8.7 Sr-90 18.311.2 21*2.6 Cs-137 10.3 1.2 10t8.7 STW-388 Water Dec. 1984 I-131 28.0*2.0 36i10.4 STW-389 Water Dec. 1984 H-3 3583t110 31821624 STW-391 Water- Dec. 1984 Ra-226 8.4tl.7 8.6i2.2 Ra-228 3.110.2 4.lil.1 STW-392 Water Jan. 1985 Sr-89 <3.0 3.0i8.7 Sr-90 27.315.2 30.0i2.6 STW-393 Water Jan. 1985 Gross alpha 3.3 1.2 58.7 Gross beta 17.313.0 1518.7 ( STS-395 Food Jan. 1985 Gross alpha 4.712.3 6.0 8.7 Gross beta 11.311.2 15.0 8.7 Sr-89 25.316.4 34.018.7 Sr-90 27.018.8 26.0t2.6 I-131 38.0 2.0 35.0 10.4 Cs-137 32.7i2.4 29.0 8.7 K-40 1410 212 13821208
, 0(
- Table A-1. -(continued) h Concentration in pCi/lb Lab. Sample Date .
.. TIML Result EPA Result Code Type Collected- Analysis- *2ac ilo, n=3d STW-397- Water Feb. 1985 'Cr-51 <29. 48*8.7 Co-60 21.313.0 20i8.7 Zn-65 53.715.0 55i8.7-Ru-106 <23 25i8.7 Cs-134 32.~3tl.2 35*8.7 Cs-137 25.3i3.0 25*8.7 STW-398 -Water Feb. 1985 H-3 38691319 3796i634 r
STM-400 Milk March 1985 I-131 7.3i2.4 9.0*1.6
- STW-402 -Water March 1985 Ra-226 4.6i0.6 5.0il.3 Ra-228 <0.8 9.0i2.3 -
Reanalysis Ra-228 9.010.4 STW-404 Water . March 1985 Gross alpha 4.7*2.3 618.7 Gross beta 11.3tl.2- 15i8.7 4 STAF-40'5 Air March 1985 Gross alpha 9.311.0 10.0i8.7 Filter Gross beta 42.0il.1 36.0i8.7 : Sr-90 13.311.0 15.0i2.6 i Cs-137 6.311.0- 6.0i8.7 STW-407 Water April 1985 I-131 8.0 0.0 7.5tl.3 , STW-408 Water April 1985 H-3 33991150 3559*630 STW-409 Water April 1985 (Blind) Gross alpha 29.711.8 32.018.7 . - Sample A Ra-226 4.4i0.2 4.lil.0
- - Ra-228 nae 6.2il.6
, Uranium nae 7.0110.4 Sample B Gross beta 74.3111.8 72.0*8.7
- Sr-89 12.3 7.6 10.018.7 Sr-90 14.712.4 15.0i2.6
- Co-60 14.712.4 15.018.7 Cs-134 12.0 2.0 15.018.7 Cs-137 14.012.0 12.018.7
.a - 5
Table A-1.- (continued) Concentration in 3Ci/lb Lab Sample Date TIML Result EPA lesult Code Type Collected Analysis 12ac ilo, n=3d STW-413 Water May 1985 Sr-89 - 36.0t12.4 39.0i8.7 Sr-90 14.314.2 15.0t2.6 STW-414 Water May 1985 Gross alpha 8.314.1 12.018.7 Gross beta 8.711.2 11.0i8.7 STW-416 Water June 1985 Cr-51 44.7i6.0 44.0*8.7 Co-60 14.311.2 14.018.7 Zn-65 50.317.0 47.0i8.7 Ru-106 55.3i5.8 62.0i8.7 Cs-134 32.7tl.2 35.018.7 Cs'-137 22.7i2.4 20.018.7 STW-418 Water June 1985 H-3 24461132 24161609 STM-421 Milk June 1985 Sr-89 10.314.6 11.018.7. ' O sr-90 I-131 9.o*2 o 11.7tl.2 11 o*2 8 11.0110.4 Cs-137- 12.7tl.2 11.018.7 K-40 1512162 1525t132 STW-423 Water July 1985 G30ss alpha 5.010.0 11.0i8.7 Gross beta 5.012.0 8.018.7 STW-425 Water August 1985 I-131 25.713.0 33.0t10.4 STW-426 Water August 1985 H-3 4363183 44801776 STAF-427 Air August 1985 Gross alpha 11.310.6 13.018.7 Filter Gross beta 46.011.0 44.018.7 Sr-90 17.7i0.6 18.0 2.6 Cs-137 10.310.6 8.0 8.7 STW-429 Water Sept. 1985 Sr-89 15.7 0.6 20.0i8.7 Sr-90 7.010.0 7.0 2.6 ST4-430 Water Sept. 1985 Ra-226 8.210.3 8.912.3 Ra-228 4.110.3 4.611.2 ST4-431 Water Sept. 1985 Gross alpha 4.710.6 8.018.7 Gross beta 4.711.2 8.0t8.7 O - A-6
- , . . - . - .. . - - . - . ~ =
Tab le : A-1. .(continued)' L
' Concentration in'oCi/lb Lab . Sample Date TIML Result EPA Result Code. Type . Collected Analysis - 12ac ilo, n=3d STW-433 Water Oct. 1985 Cr-51 <13 21.0*8.7-Co-60 19.30.6 20.0i8.7 Zn-65 19.710.6 19.0*8.7 -
1 Ru-106 <19 20.0*8.7 Cs-134 17.0il.0 20.0*8.7 Cs-137f . 19.3tl.2 ~20.0f8.7 STW-435 - Water Oct . :1985 H-3 - 1957150 1974tS98-STW-436 Water Oct. 1985-r 437 (811nd) Sample.A Gross - alpha 53.011.0 52.0i22.6 Ra-226 5.9 0.1 6.311.6 Uran m NA 8l0ib4 Sample 8 Gross' beta 85.7*2.5 75.0i8.7 Sr-89 21.311.5 27.0*8.7 Sr 10.310.6 9.0i2.6 - Co-60 18.3 1.2 18.0i8.7 Cs-134 16.311.2 18.0i8.7 Cs-137- 19.0t1.0 18.018.7 1 STM-439 Milk Oct. 1985 Sr-89 50.3*0.6 48.0*8.7-Sr-90 23.310.6 26.0*2.6 I-131 45.7*3.2 42.0*10.4- r Cs-137 60.710.6 56.0*8.7 K-40 1547129 1540*134 STW-441 Water Nov. 1985 Gross alpha 5.3i0.6 10.018.7 Gross beta 11.711.2 13.018.7 1 STW-443 Water Dec. 1985 I-131 46.7i2.1 45.0*10.4 STW-444 Water Dec . 1985 Ra-226 6.5 0.1 7.1*1.9 .. Ra-228 6.1 0.1 7.311.9 i- STW-445 - Water Jan. 1986 Sr-89 29.712.5 31.018.7 Sr-90 13.710.6 15.012.6 (). - STW-446 Water Jan. 19?6 Gross alpha Gross beta 3.0*0.0 5.3*0.6 3.0i8.7 7.018.7 u l :
-7 v-- +-w.w v, .r- -,ww,e,.ww-w.- , ,,,w,. -,,,v---.v,,e- ~,,.,w -en w m , v ,,w mw e n,--,n--+--wn, ,-.-e--,,,-m-r---wm_ -,,--~n-<e-m+
-me we ~
l I l ()' Table A-1. (continued) Concentration in pCi/lb Lab Sample- Date' TIML Result EPA Result Code Type Collected Analysis 12ac ilo, n=3d STW-447 Food -Jan. 1986 Sr-89 24.312.5 25.0i8.7 Sr-90 17.310.6' 10.0*2.6 I-131 22.7i2.3 20.0110.4 Cs-137 16.310.6 15.0i8.7 K-40 927146 950*249 STW-448 Water- Feb. 1986 Cr-51 45.013.6 38.0i8.7 Co-60 19.711.5 18.0*8.7 Zn-65 44.013.5 40.018.7 Ru-106 <9.0 0.018.7
. Cs-134 28.3t2.3 30.018.7 Cs-137 23.7*0.6 22.0i8.7 STW-449 Water Feb. 1986 H-3 5176148 5227 910 STW-450 Water Feb. 1986 U total 8.010.0 9.0110.4 STW-451 Milk Feb. 1986 I-131 7.0i0.0 9.0i10.4 STW-452 Water March 1986 Ra-226 3.810.1 4.1*1.1 Ra-228 11.010.5 12.4i3.2 STW-453 Water March 1986 Gross alpha 6.710.6 15.018.7 Gross beta 7.3i0.6 8.018.7 STW-454 Water April 1986 I-131 7.0t0.0 9.0i10.4 STW-455 Water April 1986 456 (Blind)
Sample A Gross alpha 15.0 1.0 17.018.7 Ra-226 3.1 0.1 2.910.8 Ra-228 1.5 0.2 2.0i0.5 Uranium 4.7 0.6 5.0110.4 Sample B Gross beta 28.7 1.2 35.0 8.7 Sr-89 5.7 0.6 7.0i8.7 Sr-90 7.0:0.0 7.012.6 Co-60 10.711.5 10.018.7 Cs-134 4.0:1.7 5.0 8.7 Cs-137 5.3:0.6 5.0t8.7 .O ... - A-8
. . . . = . . , _ - . - - . - _ - . . - _ - - - . . . - . - - . _ - . . . - - . . . . . . -.
Table A-1. (continued) Concentration in pCi/lb Lab Sample. Date TIML Result. EPA Result Code Type Collected- Analysis *?ac ' ila , n=3d STAF-457. Air' Apri1 1986' Gross alpha 13.710.6 15.0i8.7 Filter Gross beta 46.3t0.6 47.018.7 Sr-90 14.7*0.6 18.0i2.6 Cs-137 10.710.6 10.0*8.7. i Urine. April 1986 Tritium 4313170 '44231327 STU-458 STW-459 Water May 1986 Sr-89 4.3i0.6 5.0i8.7 Sr-90 '5.010.0 5.0t2.6 STW-460 Water May 1986 Gross alpha - 5.310.6 8.018.7 ; Gross beta 11.311.2 15.018.7 l STW-461 Water June 1986 Cr-51 <9.0 0.0i8.7 Co-60 66.011.0 66.0*8.7. In-65 87.311.5 86.018.7' O Ru-106 39.712.5 50.0i8.7 d- Cs-134 Cs-137 49.312.5 10.3tl.5 49.0i8.7 10.018.7 STW-462 Water June 1986 Tritium 3427125 31251626 i I STM-464 Milk June 1986 Sr-89 <1.0 0.0i8.7 ' Sr-90 15.310.6 16.0t2.6 I-131 48.3 2.3 41.0i10.4 Cs-137 43.7tl.5 31.0i8.7 K-40 1567t114 1600i139 STW-465 Water July 1986 Gross alpha 4.7i0.6 6.0i8.7 Gross beta 18.7tl.2 18.0*8.7 STW-467 Water August 1986 I-131 30.310.6 45.0*10.4 l STW-468 Water August 1986 Pu-239 11.310.6 10.lil.8 STW-469 Water August 1986 Uranium 4.010.0 4.0110.4 STAF-470 Air September 1986 Gross alpha 19.311.5 22.018.7 471 Filter Gross beta 64.012.6 66.0t8.7 472 Sr-90 22.011.0 22.012.6 Cs-137 25.711.5 22.018.7 STW-473 Water September 1986 Ra-226 6.010.1 6.lil.6 Ra-228 8.711.1 9.112.4 A-9
9
'0 Table A-1.- (continued).
'ConcentEation in 9Ci/lb Lab Sample Date TIML Result EPA Result-Code : Type' Collected Analysis - *2ac ' ilo, n=3d STW-474- Water September 1986 Gross alpha 16.3t3.2 15.0i8.7 ,
Gross beta- 9.0il'.0 8.0i8.7 STW-475. Water October 1986 Cr-51 .63.3t5.5 59.0*8.7 Co-60 31.0*2.0 31.0*8.7-Zn'-65 87.3t5.9 85.0i8.7 Ru-106- 74.7*7.4 74.0*8.7 Cs-134 25.7i0.6 28.0i8.7 Cs-137 46.311.5 44.0*8.7 STW-476 Water October 1986 H-3. 5918 60 5973t1035
\
STM-479 Milk November 1986 Sr-89 7.7*1.2 9.0i8.7 Sr-90 1.010.0- . 0.0i2.6 I-131- 52.313.1 49.0t10.4 Cs-137 45.7i3.1 39.0i8.7 K-40 1489i104 1565t135 STU-480 Urine ' November 1986 H-3 5540i26 5257i912 STW-481 Water November 1986 Gross alpha 12.0i4.0 20.018.7 Gross beta 20.013.5 20.0i8.7 i a Results obtained by Teledyne Isotopes - Midwest Laboratory as a participant
-in the environmental sample crosscheck program operated by the Intercom-
-parison and Calibration Section, Quality Assurance 8 ranch, Environmental'
~ Monitoring and Support Laboratory, U.S. Environmental Protection Agency,
'(EPA), Las Vegas, Nevada.
- b All results are_ in pCi/1, except for elemental potassium (K) data, which are
'in mg/1; air filter samples, which are in pCi/ filter; and food, which is in pCi/kg.
c Unless otherwise indicated, the TIML results are given as the mean *2 standard deviations for three determinations. d USEPA results are presented as the known values i control limits of b for n = 3. e NA = Not analyzed. f Analyzed but not reported to the EPA. , 9 Results after calculations corrected (error in calculations when reported to EPA).
-O A-10
1 Table A-2. Crosscheck program results, thermoluminescent dosimeters (TLDs). mR Teledyne Average 12 d ' Lab TLD Result Known (all .) Code Type Measurement 12 a .Valuec participants) ' International Intercomparisonb 110-2 CaF2:Mn Field 17.011.9 17.1 16.417.7' Bulb Lab 20.814.1- 21.3 18.817.6 3rd International Intercomparisone 115-3 CaF2:Mn Field 30.713.2 34.914.8 31.513.0 BuTb Lab 89.616.4 91.7114.6 86.2124.0 4th International Intercomparisonf 115-4 CaF2:Mn Field 14.111.1 14.111.4 16.019.0 Bulb Lab (Low) 9.311.3 12.212.4 12.017.6 Lab (High) 40.411.4 45.819.2 L43.9113.2 Sth International Intercomparison9 115-5A CaFp:Mn Field 31.411.8 30.016.0 30.2i14.6-Bulb Lab at 77.415.8 75.217.6 75.8140.4 beginning Lab at 96.615.8 88.418.8 90.7131.2. the end
O O O 4 Table A-2. (Continued) i mR Average i 2 d Teledyne i Lab TLD ' Result Known .(all-Code Type Measurement 12 a valuec participants). 115-5B LiF-100 Field 30.314.8 30.016.0 30.2114.6 i Chips
- Lab at 81.117.4 75.217.6 75.8140.4 i beginning i Lab at 85.4111.7 88.418.8 90.7131.2 the end i
! 7th International Intercomparisonh 115-7A LiF-100 Field 75.412.6 75.816.0 75.1129.8- ! Chips i
Lab (Co-60) 80.013.5 79.914.0 77.9127.6 Lab (Cs-137) 66.612.5 75.0t3.8 73.0122.2 115-78 CaF2:Mn Field 71.5i2.6 75.816.0 75.1129.8 i Bulbs
! Lab (Co-60) 84.816.4 79.914.0 77.9127.6 1
l Lab (Cs-137) 78.8tl.6 75.013.8 73.0122.2 115-7C CaSO4 :Dy Field 76.812.7 75.816.0 75.li29.8
- Cards Lab (Co-60) 82.Si3.7- 79.914.0 77.9127.6
] i Lab (Cs-137 79.013.2 75.0t3.8_ L73.0122.2 4
o o o o - Table A-2. (Continued) mR Teledyne 4 Average i 2 d' - Lab TLD Result ~ Known . (all . Code Type Measurement 12.a. Valuec- ' participants) - 8th International Intercomparisoni 115-8A LiF-100 Field Site 1 29.511.4~ 29.711.5' 28.9112.4-Chips Field Site 2 11.310.8 10.410.5 10.119.06~ Lab (Cs-137) 13.710.9 17.210.9 16.216.8 115-8B CaF2:Mn Field Site 1 32.311.2 29.711.5 28.9112.4- ~ Bulbs Field Site 2 9.0t1.0 10.410.5 10.119.0-Lab (Cs-137) 15.810.9 17.210.9 16.216.8-115-8C CaSO4 :Dy Field Site 1 32.310.7 .29.711.5 28.911214 Cards F1 eld Site 2 10.610.6 .10.410.5 10.119.0 Lab (Cs-137 18.110.8 17.210.9 16.216.8 a Lab result given is the mean 12 standard deviations of three determinations. b Second International Intercomparison of Environmental Dosimeters conducted in April of 1976 by the Health and Safety Laboratory (GASL), New York, New York, and the School of Public Health of the University of Texas, Houston, Texas. -
.E c Value determined by sponsor of the intercomparison using continuously operated pressurized ion chamber.
d Mean 12 standard deviations of results obtained by all laboratories participating in the program. e Third International Intercomparison of Environmental Dosimeters conducted in sunmer of 1977 by Oak Ridge National Laboratory and the School of Public Health of the University of Texas, Houston, Texas. f Fourth International Intercomparison of Environmental Dosimeters conducted in summer of 1979_by the School of Public Health of the University of Texas,. Houston, . Texas.. 1 Firth International Intercomparison of Environmental Dosimeter conducted in -fall of 1980 at Idaho Falls,
' i tho and sponsnred by the School of Public Health of the University of Texas, ' Houston, Texas ? and , 0,v ironmental Measurements Laboratory, New York, New York, U.S. Department of- Energy. % wnth International Intercomparison of Environmental Dosimeters conducted in the spring' and summer. of -
1984 at Las Vegas, Nevada, and sponsored by the U.S. Department of Energy,:the U.S. Nuclear Regulatory
' Commission, and the U.S. Environmental Protection Agency.
Eighth International Intercomparison of Environmental Dosimeters conducted in the fall and winter of 1985-1986 at New York, New York, and sponsored by the U.S. Department ~of: Energy.;
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c h O i Appendix B Data Reporting Conventions 'l l O
' 8-1 ,-g.m- 1.w,, .,,
r_ l l Data Reporting Conventions 1.0. Al'1 activities are decay corrected to collection time. 2.0. Single Measurements Each single measurement is reported as follows: xis where x = value of the measurement; s = 2a counting uncertainty (corresponding to the 95% confidence level). In cases where the activity is found to be below the lower limit of
. detection L it is reported as
<L where.L = is the lower limit of detection based on 3a uncertainty for a background sample.
es U.
. 3.0. Duplicate Analyses 3.1. Individual results: x1 i s1 x2 i s2 Reported result: xis where x =.(1/2) (x1 + x2) s= s 2+s2 3.2. Individual results: <L1
<L2 Reported result: <L where L = lower of L1 and L2 3.3. Individual results: xis
<L Reported result: x i s if x > L; i <L otherwise O
B-2
r; s. A
~
4.0. . Computation o'f-Averages and Standard Deviations-4.1 Averages =and standard devia'tions listed in the tables are computed from all of the individual . measurements over the period averaged; for exangle, an annual standard deviation would_not be the average of quarterly standard deviations. The average x and standard deviation (s) of a set of n numbers x1, x2, .. . . . xn are defined as follows: Y = hE x s= - rI*~*I2 n-1 4.2 Values below the highest lower limit of detection are not included ' in the average. 4.3 If all of the values in the averaging group are less than the highest LLD, the highest LLD is reported.
.- ~
If all but one of the values are less than the highest LLD, the b)' .
.4.4 single value x and associated two sigma error is reported.
4.5. In rounding off, the following rules are followed: 4.5.1. If the figure following those to be retained is less than 5, the figure is dropped, and the retained figures are kept unchanged. As an example,11.443 is rounded off to 11.44. 4.5.2 If the figure following those to be retained is greater-than , 5, the figure is dropped, and the last retained figure is. raised by 1. As an example,11.446 is rounded off to 11.45. 4.5.3. If the figure following those to be retained is 5, and if there are no figures other than zeros beyond the five, the figure 5 is dropped, and the last-place figure retained is increased by one if it is an odd number or it is kept l ) unchanged if an even number. As an example,11.435 is ' rounded off to 11.44, while 11.425 is rounded off to 11.42. l LO B-3
D-Appendix C i Maximum Permissible Concentrations of Radioactivity in Air and Water have Background in Unrestricted Areas O C-1
p: 9 6 V Table C-1. Maximum' permissible concentrations of radioactivity.in air and water labove natural background in unrestricted areas.a
-Air -
Water Gross alpha 3 pCi/m3 Strontium-89 3,000 pCi/1 Gross beta 100 pCi/m3 Strontium-90 300 pCi/1 Iodine-131b 0.14 pCi/m3 Cesium-137 20,000 pCf /1 Barium-140 20,000 pCi/1 Iodine-131 300 pCi/1 Potassium-40c 3,000 pC1/1 Gross alpha 30 pCi/1 Gross beta 100 pCi/1 Tritium 3 x 106 pCi/1 a Taken from Code of Federal Regulations Title 10, Part 20, Table II and appro-priate footnotes. Concentrations may be averaged over a period not greater than one year. , b From 10 CFR 20 but adjusted by a factor of 700 to reduce the dose resulting from the air-grass-cow-milk-child pathway. c A natural radionuclide. i i 4 !O I i l C-2
NRC-87-25 /j WPSC (414)433-1234. TELEX 5101012698 WPSC GRB TELEcORER (414)433-1297 EASYUNK 62891993 r WIBCONSIN PUBUC_SE5MCE CORPORATION 600 North Adams + RO. Box 19002 + Green Bay, WI 54307-9002 March 2, 1987 10 CFR 50.36 10 CFR 50.59 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
' Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant 1986 Annual Operating Report Enclosed is a copy of the 1986 Kewaunee Nuclear Power Plant (KNPP) Annual Operating Report. This report is being submitted in accordance with Section 6.9.1.b of the KNPP Technical Specifications.
The 1986 KNPP Annual Operating Report also satisfies the reporting requirements of 10 CFR 20.407(a)(2) and 10 CFR 20.407(b) (personnel monitoring), KNPP Technical Specification 4.2.b.5.b (steam generator inspection), and KNPP Technical Specification 6.9.3.a (environmental monitoring). Finally, Section 3 of the report describes those facility changes allowed by 10 CFR 50.59(a)(1). Very truly yours, M/ D. C. Hintz Vice President - Nuclear Power 4 l DJM/jms Enc. cc - Mr. Robert Nelson, US NRC US NRC, Region III h t
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