ML20154B753
| ML20154B753 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 12/31/1985 |
| From: | Hintz D WISCONSIN PUBLIC SERVICE CORP. |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| CON-NRC-86-24 NUDOCS 8603040387 | |
| Download: ML20154B753 (112) | |
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KEWAUNEE NUCLEAR POWER PLANT ANNUAL OPERATING REPORT C)
1985 WISCONSIN PUBLIC SERVICE CORPORATION WlSCONSIN POWER a
L!GHT COMPANY M A DISON GAS S ELECTRIC COMPANY O
i
O TABLE OF CONTENTS Page No.
1.0 Introduction 1
2.0 Summary of Operating Experience 2
3.0 Plant !!odifications, Tests and Experiments 11 4.0 Licensee Event Reports 22 5.0 Fuel Inspection Report 37 6.0 Challenges to and Failures of Pressurizer 38 Safety and Relief Valves 7.0 Steam Generator Tube Inspection 39 8.0 Personnel Exposure and Monitoring Report 49 9.0 Radiological Monitoring Program 53 t
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1.1
1.0 INTRODUCTION
The Kewaunee Nuclear Power Plant is a pressurized water reactor licensed at 1650 MWt.
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 535 MWe net.
The architect / engineer was Pioneer Service and Engineering (PSE) from Chicago.
i 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 commercial,
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Kewaunee has generated 41,641,212 MW hours of electricity as of December i
31, 1985, with a net plant capacity factor of 76.9 (using net DER).
1.1 Highlights During the year, the Kewaunee Nuclear Power Plant was primarily base loaded. The unit was operated at 84.0% capacity factor (using net MDC) with a gross efficiency of 33.3%. The unit and reactor availabi-lity were 82.4% and 83.0% respectively. Table 2.1 is a compilation of the monthly summaries of the operating data, Table 2.2 contains the yearly and total summaries of the operating data, and Figure 1.1 pro-vides a histogram of the average daily electrical output of the-j Kewaunee Plant for 1985.
I On February 8, 1985, the unit was removed from service for its tenth
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annual refueling / maintenance overhaul. Thirty-six fresh fuel assemblies were loaded for Cycle.XI. The unit was returned to service on April 11, 1985.
2.1 2.0'
SUMMARY
OF OPERATING EXPERIENCE January Normal power operation continued through the entire month of January.
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PLANT SHUTDOWNS: There were no plant shutdowns during the month of-January.
February On February 8, the unit was shutdown for refueling / maintenance.
4 PLANT SHUTDOWNS: February 8, scheduled shutdown - 503.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.
Commenced i
- Cycle X-XI refueling outage.
March 1
1 In March, the Cycle X-XI refueling outage continued.
4 i
j PLANT SHUTDOWNS: March 1, scheduled shutdown - 744.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Continued s
Cycle X-XI refueling outage.
April On April 11,-the Cycle X-XI refueling' outage was concluded.
On April 11, the unit was released for operation.
On April 11, a short outage was taken.
l PLANT SHUTDOWNS: April 1, scheduled shutdown - 250.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Continued Cycle X-XI refueling outage. The outage was concluded-1 on April 11.
April 11, a scheduled shutdown - 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Performed the Turbine Overspeed Trip Test.
May On May 12, load was reduced to 72% power for the performance of the Monthly y
Turbine Stop Valve Test. The unit was returned to full load the same day.
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2.2 On May 20, load was reduced to 50% to. permit cleaning of the Condensate pump suction strainers. The unit was returned to full load the next
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morning.
PLANT SHUTDOWNS: There were no plant shutdowns during the month of May.
June On June 16, load was reduced to 72% for performance of the Monthly Turbine Stop Valve Test. The unit was returned to full load the same day.
2 PLANT SHUTDOWNS: There were no plant shutdowns during the month of June.
July On July 14, load was reduced to 72% for the performance of the Monthly Turbine Stop Valve Test. The unit was returned to full load the same day.
PLANT SHUTDOWNS: There were no plant shutdowns during the month of July.
qNJ August On August 6, load was reduced to an average 187 MWE gross for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> for cleanup of Steam Generator chemistry following contamination of the secon-dary plant by lake water.
A short outage was required on August 8 for repair-of a 2-inch excess steam vent line which ruptured during the Monthly Turbine Stop and Governor Valve Test.
PLANT SHUTDOWNS: August 8, Forced Shutdown - 14.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> for repair of a 2-inch excess steam vent line from a MSR which ruptured during the Monthly Turbine Stop and Governor Valve Test.
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2.3 Septenber
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On September 8, load was reduced to 7EE power for the Monthly Turbine
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Stop and Governor Valve Test. The unit was returned to full power the same day.
PLANT SHUTDOWNS: There were no plant shutdowns during the month of September.
October On October 6, load was reduced to 72% power for the Monthly Turbine Stop and Gcvernor Valve Test. The unit was returned to full power the same 4
day.
PLANT SHUTCCVNS: There were no plant shutdowns during the month of October.
4 November
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On November 3, load was reduced to 72% power for the Monthly Turbine Stop and Governce Valve Test. The unit was returned to full power the sar.e day.
On November 13, a structural failure of an a-ir operatcr for one of the feedwater regulating valves lead to a plant trip. The unit was returned to service on the following day.
PLANT SHUTDOWNS: Un November 13, forced shutdown - 22.8 hcurs due to a structural failure of an air operator for cne cf the feedwater regulating valves.
i December On December 8, load was reduced to 72% power for the Monthly Turbine Stop and Governor Valve Tests. The unit was returneo to full power the i
ps same day.
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.On December 12, an instrument bus outage resulted in a reactor trip when i
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the feedwater regulating valve powered from the affected instrument bus i
went to its failed shut position. The unit was returned to operation the i.
same day.
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- j-PLANT SHUTDOWNS: On December 12, forced shutdown - 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> due to an
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O TABLE 2.1 (Page 1 of.2) 2.5 ELECTRICAL POWER GENERATION CATA (1985)
MONTHLY J
January February March April May June Hours RX was critical 744.0 168.1 0.0 504.2 744.0 720.0
)
i RX Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 0.0 1
Hours Generator On-Line 744.0 168.1 0,0 467.6 744.0 720.0-l Unit Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 0.0 t
1 Gross Thermal Energy Generated (MWH) 1,224,279 271,048 0
682,721 1,212,434 1,179,472 i
j h
Gross Elec. Energy Generated (HWH) 406,000 89,800 0
231,600 404,900 395,400' 5
Net Elec. Energy Generated (NWH) 387,721 85,656 0
220,130 385,795 377,013
(
RX Service Factor 100.0 25.0 0.0 70.1 100.0 100.0 RX Availability Factor 100.0 25.0 0.0 70.1 100.0 100.0 t
i Unit Service Factor 100.0 25.0 0.0 65.0 100.0 100.0 I
Unit Availability Factor 100.0 25.0 0.0 65.0 100.0 100.0 i
Unit Capacity Factor (ttsing MDC net) 103.6 25.3 0.0 60.9 103.1 104.1 4
Unit Capacity Factor (using DER net) 97.4 23.8 0.0 57.2 96.9 97.9 1
Unit Forced Outage Rate 0.0 0.0 0.0 0.0 0.0 0.0 Hours in Month 744 672 744 719 744 720 l
Net MDC (Mwe) 503 503 503 503 503-503 i
4 e
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l TABLE 2.1 (Page 2 of 2) 2.6 ELECTRICAL POWER GENERATION DATA (1985)
HONTHLY July August September October November December Hours RX was critical 744.0 735.4 720.0 745.0 701.6 740.2 RX Reserve Shutdown Hours 0.0 0.0 0.0 0.0 0.0 0.0 Hours Generator On-line 744.0 729.3 720.0 745.0 697.2 735.5 linit Reserve Shutdown Hours 0.0 1.0 0.0 0.0 0.0 0.0 Gross Thermal Energy Generated (MWH) 1,218,713 1,148,512 1,178,061 1,222,974 1,121,094 1,181,301 Gross Elec. Energy Generated (MWH) 406,800 379,400 389,300 409,400 375,300 393,300 Net Elec. Energy Generated (MWH) 387,658 360,966 370,728 390,317 357,687 375,505 RX Service Factor 100.0 98.8 100.0 100.0 97.4 99.5 RX Availability Factor 100.0 98.8 100.0 100.0 97.4 99.5 Unit Service Factor 100.0 98.0 100.0 100.0 96.8 98.9 Unit Availability Factor 100.0 98.0 100.0 100.0 96.8 98.9 Unit Capacity Factor (using MDC net) 103.6 96.5 102.4 104.2 98.8 100.3 Unit Capacity Factor (using DER net) 97.4 90.7 96.2 97.9 92.9 94.3 Unit Forced Outage Rate 0.0 2.0 0.0 0.0 3.2 1.1 Hours in Month 744 744 720 745 720 744 Net M00 (Mwe) 503 503 503 503 503 503
2.7 TABLE 2.2 ELECTRICAL POWER GENERATION DATA r
1985 l
YEAR CUMULATIVE Hours RX was critical 7,266.5 86,017.0 RX Reserve Shutdown Hours 0.0 2,330.5 Hours Generator On-Line 7,214.8 84,555.5 i
Unit Reserve Shutdown Hours 0.0 10.0 Gross Thermal Energy Generated (MWH) 11,649,708 132,716,832 Gross Electrical Energy Gen. (MWH) 3,881,200 43,738,500 Net Elec. Energy Generated (MWH) 3,699,176 41,641,212 RX Service Factor 83.0' 85.0 RX Availability Factor 83.0 87.3 l
.I Unit Service Factor 82.4 83.5 Unit Availability Factor 82.4 83.5 Unit Capacity Factor 84.0 79.5 (using MDC net)
Unit Capcity Factor 78.9 76.9 j
(using DER net)
Unit Forced Outage Rate 0.6 3.3 Hours in Reporting Period 8,760 101,209 l
1 J
i l
i lO I
i i
5
3,1 3.0 PLANT MODIFICATIONS, TESTS AND EXPERIMENTS 4
r' This section is provided in accordance with the requirements of Part 50.59 D]
(b) to Title 10, Code of Federal Regulations (10CFR50.59(b)). This regula-tion allows licensees to make changes in the facility as described ili the i
Updated Safety Ar.alysis 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 s change in the Technical Specifications or an unreviewed safety question.
10CFR50.59(b) requires that such changes be reported on An annual basis.
t Plant Modifications, 10CFR50.59 There were no modifications during 1985 which introduced,an unreviewed safety question and, therefore, prior NRC approval was not required.
- i r
The following summary of modifications includes those significant modifica-O tie ce P etee deri
- 198s se eet grevieesix revertee aa x er <*ese i
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 cor,sidered to be of significance, warranting mention in this summary report.
Reactor Control and protection Dedicated power sources were provided for the manual reactor trip relays., ;
These relays were on the same circuit as the Main Steam isolation test relays and Main Steam header isolation valve solenoid valves.
(DCR 1458).
Summary of Safety Evaluation Failure of unrelated equipment can no longer effect the manual reactor trip capability; therefore, safety and reliability are enhanced.
i
3.:2 73 480V Supply and Distribution y/
All continucus duty 440 Volt rated motors that were fed from the safeguards buses were replaced with 460 Volt rated motors, or removed from safeguards power if they were ncnsaf eguards n.otors. This modification allows greater flexibility in setting taps on KNPp Auxiliary transformers to freet system extremes in high and low voltage (DCR 1373 Rev. 1)
Summary of Safety Evaluation This modification will increase notor life and reliability; therefore, plant safety is enchanced.
Safeguards Coo' ling Nodificaticns Additicnal ambient cooling capacity was added in the folicwing areas:
- Auxiliary Building Basement, EL 5863-0" (DCR-1635)
- Auxiliary Building Fan Flocr EL 657'-0" (DOR-1630)
Ci
- Turbine Building Basement., EL 586'-0" (DCR-1631)
This capacity was added to ensure that area temperatures do not exceed 104 F during normal operations and long-term operation of safeguards equip-ment, with or without auxiliary building special ventilation running.
Summary of Safety Evaluatien The additional ambient cooling capacity will prevent area tem-peratures (both norn.al and post-accident) from exceeding qualification temperatures of safety related eauipment in their respective areas, thereby increasing equipnant reliability, and _ enhancing overall safety.
OO m
's 3.3 Reactor Building Vent
~
Two Control Rod Drive Mechanism (CRDM) cooling coils were installed on top of the reactor vessel missile shield. These cooling units use service water)from the discharge of the safeguards Containment Fan Coil Units (CFCU) as their source of cooling water. Containment ambient temperatures have been decreased and the safety function of the service water system M unchanged.
(DCR 1114)
Summary of Safety Evaluation The CRDM cooling coils were installed with one on each train of the service water system, preserving safeguards system separation. The service water flowrate to containment has not changed significantly
.}
during normal operation; and following a safety injection signal the CRDM fan coolers are isolated and the service water flow path to the O
cecu's is es 4t s ar4er te the caoa r ceeier 884 tie #.
8 = ese containment ambient temperctures have been reduced, equipment reliabi-lity is enhanced. There are no adverse safety consequences.
Environmental Qualification Electrical equipment in various systems was upgraded to improve its environmental qualific:.c tons. These upgrades included:
- Replacement of.<.everal limit switches for dampers in the Reactor Building Vent and Shield Building Vent Systems (DCR 1545).
1 0
3.4 1
- Relocation of Auxiliary Building Special Ventilation, Steam Exclusion, and Shield Building Ventilation power contactors, indicating lights, and i
I associated control components to a mild environment, and repowering the filter assembly humidity alarm modules to prevent environmentally induced
.s degradation of the QA 1 power sources. (DCR 1546)
- Repowered non E-Q pilot solenoid valves in the Service Water System to-prevent environmentally induced degradation of control circuits for E-Q Service Water Control Valves. (DCR 1547)
- Separating the power supplies for the control room indicating lights for the Main Steam Isolation Valves in the Main Steam System (DCR 1548).
- Separating the power supplies for the caustic additive valves in the Con-tainment Spray System (DCR 1549).
. Replacement of Pressurizer Power Operated Relief Vaive (PORV) solenoid valves. (DCR 1550)
- Replacement of solenoid valves in the Primary Sampling System (DCR 1258).
- Drilling of two (2) weep holes in the Control Room Ventilation Remote Panel to permit condensate drainage (DCR 1417).
- Replacement of damper. actuators and associated limit switches in the Shield Building Ventilation System (DCR 1433).
- In the Miscellaneous Drains and Sumps System; replacement of sump level switches and solenoid valves and relocation of the Deaerated Drains Tank pump controls to a low radiation area. (DCR 1369)
Summary of Safety Evaluation In each case replacement (equipment upgrade), or relocation to a less harsh environment resulted in a higher degree of component reliability during post-accident operation; plant safety was improved.
O
3.5 4
- Reactor Coolant Piping' downstream of the Pressurizer Safety Relief Valves was modified to accommodate stress induced by valve actuation (DCR-1326).
Summary of Safety Evaluation Stress levels.in the affected piping are now.within the USAR allowable stresses, and the design meets the requirements of NUREG 0737. As a result, overall plant safety is increased.
Reactor Cavity Boot Seal A pressure relief-valve was installed in the air /N2 supply to the reactor cavity boot seal. This relief valve was installed to prevent overpressuri-zation of the cavity seal while the seal is.being inflated.
(DCR 1567) 4 Also, provided backup nitrogen system to the refueling cavity boot seal.and-provided a control room alarm for loss of pr:ssure to the cavity boot seal (DCR 1608).
}
-Summary of Evaluation Plant safety is increased as installation of the pressure relief valve 1
will provide additional assurance that a failure of the reactor cavity i
boot seal will not occur due to overpressurization. The back-up
}
nitrogen system and low pressure alarm provide additional assurance i
that a depressurization will not occur.
i Main Condenser i
~
The admiralty brass tubes in the main condenser were replaced with Type 439 stainless steel tubes.
(OCR 1551)
Summary of Safety Evaluation i
This modification will improve condenser reliability. -Also, the potential for steam generator copper related problems, to which admiralty brass tubing contributes, was reduced.
3.6 Pipe Thinning Portions of the carbon steel piping in the Heater Drain System and the Bleed Steam System were replaced with stainless steel piping..This change was required due to wall thinning, a result of water impingement erosion.
(DCR's 1639, 1640)
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 possibility of an outage due to secondary side unavailability.
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 l
approximately sixty components operational from the panel.
- The installation of required instrumentation on the dedicated shutdown panel is 75% complete.
- The three hour fire wall installation and penetration sealing is 85%
complete throughout the plant.
- Cable pulling required for separation of dedicated and alternate shutdown equipment is approximately 98% complete.
- Fire detection system modifications are complete.
- HVAC system modifications are 98% complete.
(DCR's 1189, 1191, 1192, 1193, 1194, 1195, 1197, 1361).
O
3.7 S_ummary of Safety Evaluation 4
These modifications enhance both automatic and manual control of
(
the plant in the unlikely event of a fire. The Appendix R modifica-tions will preclude a fire from affecting the capability to bring the plant to a safe shutdown.
Security Perimeter detection system zones 7 and 8 were converted from fence mounted to a free standing design. This was done to reduce the large number of false alarms and to eliminate the interference from the security fence (DCR 1685).
Summary of Safety Evaluation This modification will eliminate interference from the security fence, reduce the excessively high number of false alarms, and permit better access to the security system for maintenance and testing.
Security Provide security modifications required by the new Office / Warehouse Annex (OCR 1356).
Summary of Safety Evaluation lhe design and type of equipment used in the plant security system have not changed. All changes are consistent with 10CFR73. The degree of surveillance or access control is not changed as a result of this modification. Plant safety is not affected.
Makeup Water System Demin and Secondary Sampling System Installed on-line Dissolved Oxygen Analyzers at the condensate pump discharge. (DCR 1423) l
. _. _ _. _ _ _, _. ~...
i 3.8 Summary of Safety Evaluation This modification will enable on-line monitoring of water treatment l
system performance to verify that oxygen concentration levels are within Kewaunee chemistry guidelines. Plant reliability is i
enhanced by the ability to detect increasing levels'of dissolved.02 at
[
the condensate pump discharge.
1
(
Chemical and Volume Control The diaphragms were removed from CVC monitor tanks A & B due to problems with pressure equilibration when discharging the tanks.
(DCR'1606) i i
Summary of Safety Evaluation 1
All condensate in the monitor ta;;ks is discharged to. Lake Michigan.
The purpose of the diaphragms was to prevent air fro. being absorbed i
in the water stored in the monitor tanks. Because the monitor tanks are discharged, and the contents are not sent to the reactor makeup system, there are no safety implications with removing the diaphrages.
Solid Waste Processing (RWS) i Modified and installed the necessary equipment in the Rad Waste System-(RWS) solidification system to permit use of High Integrity Containers (HIC) for spent resin shipment (DCR 1414).
Summary of Safety Evaluation l
HIC's have been approved for shipment of spent resin under 10CFR61.
1 This design change has no adverse effect on plant safety.
}
Computer Added auxiliary feedwater flow, wide range containment pressure and wide 4
range containment sump level indication to the Honeywell Computer for display on the Safety Assessement System (SAS).
(DCR 1689) i
-,,. - - -., _, -,.- ~,,, -, _,,.. -.
...-,,,v.,_m%,..,w,_v,,.-
,,,,,mo.,y,.,
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,,v,,w,,,.__,
n,-,,,,,,rywwv,,-y-,,,,-,w,.3%,,w,,w.,-,
3.9 Summary of Safety Evaluation The addition of the parameters listed above enhances the Safety Assessment Systems' capability to monitor and display the five impor-tant plant functions identified in NUREG 0696, and Supplement 1 to NUREG 0737.
l Makeup Water and Secondary Sampling i
l The scale of the demineralizer and secondary analysis system specific con-ductivity recorders was changed from a linear to a non-linear scale. Also, the secondary analysis pH recorder scale was changed from 2-12 pH to 6-13 pH. The above changes were made to improve resolution in the range of nor-mal readings.
(DCR 1424) 4 Summary of Safety Evaluation j
In each case the modification improved resolution and accuracy of readings, thereby improving trend indications.
Buildings - Structures Expanded the relay room to house I/O Cabinets for new computer (DCR 1257).
Summary of Safety Evaluation The expansion was completed following the applicable guidelines in the USAR and 10CFR50 Appendix R requirements. As a result plant safety is unaffected.
Service Water An ultrasonic flow meter was installed which locally indicates the i
flow rate in each of the service water headers. (DCR 1372)
3.10 Summary of Safety Evaluation The service water flow indication was provided to more accurately determine the dilution of plant liquid waste discharges during refueling when both circulating water pumps are shutdown. The flow meter has no safety related function.
Miscellaneous Numerous equipment changes were required as a result of vendors dropping out of the nuclear market or equipment obsolence (DCR's 1289 and 1325).
Summary of Evaluation These changes involved finding equivalent or better replacement equip-ment from qualified suppliers and updating the associated documen-tation. Because the level of quality is equivalent or better, there are no adverse safety consequences.
I O
s 3.2-1
' 1 3.2 Plant Procedures, 10CFR50.59 I ()
There were no procedure revisions during 1985 which introduced an unreviewed safety question or which changed procedures as described in i
the Updated Safety Analysis Report.
i 3.3 Tests And Experiments, 10CFR50.59 i
Core Reload / Physics Testing Thirty-six (36) fresh region M assemblies were loaded for cycle XI.
t Routine start-up physics testing was performed and reported in the Cycle XI start-up report.
i
)
Summary of Safety Evaluation A 10CFR50.59 reload safety analysis was performed and sub-mitted December 7, 1984.
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4.1 4.0 LICENSEE EVENT REPORTS This section is a summary of the 23 Licensee Event Reports (LER) submitted to the NRC in 1985 in accordance with the requirements of 10CFR50.73. None of the events described in 1985s' LER's posed a threat to the health and safety of the public.
LER 85-01 At 1437 on January 22, 1985, during full power operation, there was an inadvertent actuation of the IB Internal Containment Spray System.
The IB pump ran for 1 minute and 40 seconds discharging an estimated 2500 gallons of borated water into the containment building before being secured. The pump start occurred during the performance of SP55-155, " Engineered Safeguards Logic Test".
When the pump start occurred, the operator verified that it was inadver-1 tent, secured the system and reset containment spray. The operators received various battery ground alarms as a result of instrument malfunc-tions in containment.
At 1525 it was discovered that the RWST level was below technical specifi-cation (TS) limits, refilling was started and preparations were made to begin a plant power reduction. The RWST level was above TS setpoint at 1555 hence no reduction in power was initiated.
Immediate actions were taken to assess the situation and identify the cause.
Long term actions were planned to clean the containment interior, and perform an evaluation to identify potential hardware modifications which would prevent reoccurrence.
~
O
4.2 LER 85-02 g
On January 25, 1985, with the plant at full power operation, Wisconsin Public Service Corporation was notified by their Architect Engineer that a seismic analysis for the non-safety-related piping section of the Containment Integrated Leak Rate Test Penetration could not be located.
This was identified during the evaluation of a proposed modification. To correct this deficiency, a design change was completed during the 1985 refueling / maintenance shutdown with a proper seismic analysis of the as-modified design. WPSC considered this item of sufficient signifi-cance to merit reporting to the Commission within the "0THER" category of 10 CFR 50.73.
LER 85-03 On February 8, 1995, a plant operating mode change was in progress from 15 percent reactor power to hot shutdown. Following the transfer of steam generator level control from main feedwater to auxiliary feedwater and the 1
manual opening of the main generator output breaker, the indicated water level in the IB Steam Generator went below the lo-lo level setting, (17%
narrow range level) initiating a reactor trip.
Plant operating pro-cedures were followed to place the plant in the hot shutdown operating mode. This event was initiated by a personnel error. There was no effect on the health and safety of the public. As corrective action the text description of this event was routed to plant reactor opera-tors and the training department.
- O
4.3 LER 85-04 At 1230 on February 10, 1985, with the plant in a refueling shutdown, a control room operator noticed the IB Reactor Coolant Pump running.
4 Subsequent investigations revealed that the pump had inadvertently started due to a grounded condition in the. actuation circuitry associated with the l
4160V switchgear. The ground was caused by water accumulation in a pressure switch as a result of an inadvertent containment spray (reference LER 50-305/85-01). The ground provided enough current to gate the solid state starting circuitry.
An evaluation of the event showed that due to the location of the safety related switchgear in the plant, and the routing of associated cables, no credible single event would result in actuation of redundant trains of switchgear in a manner which could violate the assumption of the safety
[
analysis. Consequently, the event posed no nuclear safety concerns.
i This event was reported under OTHER as an item of general interest.to the industry.
I LER 85-05 On February 11, 1985, with the plant in a refueling shutdown, a control room operator discovered that the 1A Exhaust Fan of the Auxiliary Building Special Ventilation System (ABSV) was running.
Investigation revealed that the coil on the solenoid valve controlling air to the Zone SV Exhaust Filter 1A Inlet Damper had burnt out, failing the three-way solenoid valve to the vent position. This opened the inlet damper which in turn opened the exhaust damper and started the fan on Train A of the ABSV system.
l The system failed in the safe position. There was no impact on the health 1
and safety of the public.
j l
4.4 LER 85-06 On February 20, 1985, while shutdown for refueling and during the Steam Generator tube eddy current' examination, a tube in the 1A steam generator requiring plugging in 1984 was found plugged in the hot leg only. An adja-cent tube, not requiring plugging, was found plugged in the cold leg
'only. The tube that required plugging had a 55% thru-wall indication in 1984 and a 91% thru-wall indication in 1985. The exact-cause of this event remains unknown; however, it is suspected that the cold leg tube sheet was mismarked during the 1984 steam generator tube plugging effort.
To prevent recurrence of this event the tubesheet templates, rather than the tubesheets, are marked to identify the tubes to be plugged. These templates are independently verified prior to tube plugging. The J
installed plugs are verified against the tube plugging list and a video tape is made of the tube sheets for final verification. Twenty-six tubes-in the 1A S/G and 22 tubes in the 1B S/G were removed from service as a result of tube plugging in 1985.
LER 25-07 x
At 1601 on February 25, 1985, while in a refueling shutdown condition, a sur-veillance procedure on pressurizer pressure transmitters was being con-ducted. Luring this procedure the I&C technician performing the calibration asked that the red pressurizer pressure channel be tripped.
Following this, he calibrated the white channel. With one channel tripped and another with an artificial input > 2000.psig, the SI signal was reset.
This, coincident with steam generator pressure < 500 psig caused a safety injection signal. Plant operating procedures were followed to restore the plant to normal refueling shutdown conditions. No equipment or system failures contributed to this event. This event resulted from the I&C tech-nician in the field requesting the wrong channel be tripped. The sur-s
4.5 veillance procedure is being revised to prevent reoccurrence. This q
procedure is only conducted during refueling shutdown. The plant equipment b
lineup at shutdown prevented this event from having any adverse safety implications.
LER 85-08 At 1647 on February 25, 1985, with the plant in a refueling shutdown con-dition, alarm 47001-34, " Condenser Low Vacuum Turbine Trip", momentarily cleared and then alarmed again. Following this, both diesel generators started. After investigating the cause, the diesel generators were I
secured.
Investigation revealed that these two events were caused by maintenance to the turbine trip mechanism located on the turbine pedestal. During this maintenance the turbine manual trip / reset lever was momentarily placed in the latch position allowing turbine auto stop oil pressure to increase to i
the point where the turbine trip pressure switches were reset. This cleared the Condenser Low Vacuum Turbine Trip alarm. The lever was then 4
placed in the trip position allowing the auto stop oil pressure to l
decrease. As the pressure fell below 45 psig, the Condenser Low Vacuum Turbine Trip signal and Diesel Generator Start signals were initiated.
To prevent reoccurence, the diesel generator start signal from a turbine trip will be removed from service as part of the shutdown evolution during i
j extended outages and returned to service prior to unit start-up.
I This event had minimal impact on plant activities and no effect on the public health and safety.
1 4
4.6 LER 85-09 On March 12, 1985, with the plant in refueling outage, information was acquired indicating that a contracted employee, performing maintenance at the plant, had received a whole body occupational dose in excess of the 10 CFR 21.101 standard of 1.25 Rems per calendar quarter prior to the licensee determining the individual's accumulated occupational dose as required by this regulation. The worker received an accumulated whole body dose of 1.46 Rees in the month of February, 1985, as indicated by the i
worker'stwothermoluminescentdosimeters(TLD). However, the worker's self-reading dosimeter (SRD) indicated an exposure of only 0.85 Rems. The accumulated occupational dose is normally determined when the SRD exposures reach about 1.1 Rems.
Investigations have failed to explain the differen-ces in dosimeter readings, and Wisconsin Public Service Corporation (WPSC) has concluded that this is an isolated incident. The worker's accumulated occupational dose was_immediately determined and is well within NRC requirements.
In addition, authorization procedures were initiated allowing the worker to exceed the 1.25 Rem limit. During investigation of-this event, it was determined that procedural violations occurred in two instances because the worker's TLD was not processed at the SRD action dose of 150 mrem per day. Appropriate personnel have been reinstructed on this procedural requirement. This report was submitted pursuant.to the requirements of 10 CFR 20.405 (a)(1)(i).
1 LER 85-10 On April 5, 1985, the plant was in the Hot Shutdown Operating Mode with the Reactor subcritical, following a refueling outage. Shutdown Banks A & B and Control Bank-C were fully withdrawn in preparation for rod drop testing. A reactor trip occurred due to a Steam Flow greater than Feed.
Flow signal coincident with a Lo Steam Generator (S/G) Level Signal.- The
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i 4.7 operators performed the immediate actions prescribed in the Reactor Trip procedure.
Investigation revealed that one transmitter for Feed Flow and one transmitter for Steam Flow were out of calibration'resulting'in a SF >
^
FF trip signal being present. The Balance of Plant operator allowed the j
i level in Steam Generator 1A to drop to the low level setpoint. Because there was a SF > FF signal present this completed the coincidence, and a RX trip occurred. The Feed Flow and Steam Flow instruments were recalibrated prior to continuing with rod drop testing. The operator was reminded j
of the importance of reactor trip signals even when the plant is' 1
i shutdown. No further corrective action is planned. The Reactor Protection System performed as required, and there was no impact on the
{
j health and safety of the public, t
i l
LER 85-11
{
4 At 1100 on April 7, 1985, with the plant in the Hot Shutdown Operating mode following a refueling outage, the Auxiliary Building Special Ventilation System actuated.
The system actuated from a steam exclusion signal caused j
by steam issuing through a blown rupture disk on the boric acid evaporator condenser.
Immediate actions were taken to verify proper actuation of the Auxiliary Building Special Ventilation System and investigate the cause.
This event occurred because the boric acid evaporator, which was out of service, was isolated with a leaking control valve, rather than the manual
{
isolation valves.
Isolating equipment with manual isolation valves, where i
applicable, was discussed in operator training to prevent recurrence of j
this type of event.
1 lO i
4.8 LER 85-12 At 1522 on April 10, 1985, with the plant in the Hot Shutdown operating mode preparing to start up following a refueling outage, a reactor trip occurred during performance of the surveillance procedure to calibrate the intermediate range nuclear instrumentation channels. The trip was caused i
by the P-6 relay chattering, which was introduced through the grounding of the test equipment. The chattering bistable blew the control power fuses for intermediate range channel N35 detector. This completed the l
one out of two actuation logic for an Intermediate Range Hi Flux Reactor Trip. Immediate actions were taken to stop the dilution in progress and verify the reactor trip. The reactor was in the shutdown condition with the control banks inserted prior to the event and the reactor protection I
system performed as designed, hence there was no impact on public health and safety.
i LER 85-13 On May 5, 1985, with the plant at 100% power, while performing SP 56C-093,
" Containment Hydrogen Monitor Operational Test," the heat tracing circuit on the suction line to the 1A containment hydrogen analyzer was discovered inoperable.
Investigation revealed that this condition had existed since i
i April 4, 1985.
On April 4, 1985, the Shift Supervisor issued a work i
i request to repair the inoperable heat tracing circuit. The Shift i
Supervisor was aware of the recently issued Technical Specification (March 3, 1985) regarding hydrogen monitor operability; however, the loss of one train of redundant heat tracing, although degrading the system, did not clearly render the hydrogen monitor inoperable.
Corrective actions were not completed due to the unavailability of spare parts, and on May 5 the
4.9 failed heat tracing circuit was discovered again. At this time Management O
evaluations conservatively concluded that the 14-day LCO_on. hydrogen moni-4 i
tor operability had been exceeded and preparations for an orderly. shutdown commenced. Repairs were completed in three hours and a power reduction was-not required.
Further evaluation concluded that plant operation was within Technical Specifications as the redundant train's heat traced suc-i tion line could have been valved into the 1A H2 monitor. This event was reported under OTHER to identify the significance that auxiliary com-ponents have in determining equipment operability.
Corrective actions included routing this LER to all SRO's for review and providing training to operations personnel on hydrogen monitor operation. Also, SP 56C-093 was revised to include heat tracing operability in the acceptance criteria.
)
LER 85-14 i
At 1730 on June 15, 1985, the Auxiliary Operator found the concrete block, j
which prevents access to the spent resin storage tank room, removed.
f The entry way was barricaded and a high radiation area sign posted. The j
tank was reading 1 to 4 Rem /Hr on contact with a general area background reading of 0.2 to 0.3 Rem /Hr. This was a violation of Technical Specification 6.13.1.b. which requires that each High Radiation Area, where i
the intensity of radiation is greater than 1 Rem /Hr, be maintained under the administrative control of the Shift Supervisor.
The block had been I
removed on the morning of June 13 to allow design change and maintenance work on the tank.
During the day shifts of June 13 and June 14, access was positively controlled by radiation protection personnel providing coverage in the area. During the associated backshifts, when there is i
minimal activity in the Auxiliary Building, the area was barricaded and
___.--,,m.
..,,,.,._,-,,.-~.,r
4.10 posted as a high radiation area. On June 15 when the operations personnel discovered the. situation, the on-shift radiat_ ion protection technician was directed to survey the~ area, and the concrete block was returned to its proper location with the locking device secured at 1835.
To prevent-recurrence, Technical. Specification requirements for securing high radiation areas were discussed during the weekly Radiation Protection staff-meeting on June 18, 1985.
LER 85-15 On June 17, 1985, Fluor Engineers, Inc. notified Wisconsin Public Service Corporation that the seismic qualification of the emergency diesel genera-tor differential relays could not be conclusively determined. The relays are General Electric Model 12CFD22B1A.
With this information and addi-tional details provided.in INPO SER 18-84 Supplement 1, " Diesel Generator Differential Relays Not Seismically Qualified", a management decision was made to defeat the differential relays' trip function. This was completed by 1600 on June 17, 1985.
The long term resolution of this discrepancy is being evaluated, and details will be provided in a supplemental report.
LER 85-16 On July 25, 1985 at 1101 with the plant at 100% power, a control operator observed the position indication lights for several containment isolation val'ves and steam exclusion dampers change position.
Immediately, an investigation into the cause_of the event was initiated and after verifying that plant ccnditions were normal, the operators returned the equipment to its normal operating configuration. At the time of the event-a QC Technicien was verifying wire codes in Relay Rack 170, AC Safeguard Bus 5 Distribution Fuse Panel (RR170) from an approved procedure under cognizance of the Shift Supervisor. The event occurred when power was
4.11 momentarily interrupted to RR170 due to manipulation of an imprcper crimp on the RR170's power lead. The power interruption caused several contain-ment isolation valves to perform their isolation function and a momentary loss of system redundancy. There was no impact on the health and safety of the public. This event is being reported under 10 CFR 50.73(a)(2)(iv) as an automatic initiation of an Engineered Safety Feature. A procedure was developed to allow crimping RR-170's power lead without interrupting power ard was successfully completed on August 2, 1985. A preventative main-tenance procedure is being written to visually inspect wire terminations in selected terminal boxes, relay racks, and other electrical enclosures.
This procedure will be performed during the 1986 refueling outage and at periodic intervals thereafter.
LER 85-17 O
On August 8, 1985 at 0032 a manual reactor trip was required in order to isolate a ruptured excess steam vent line from the IA2 MSR to the 15A feed-water heater.
Immediately after the trip, recovery actions were followed per procedure and a post trip review was performed. All of the equipment necessary to ensure a complete reactor and turbine trip operated properly.
At 0825 on the same day, when attempting a reactor startup, the operator failed to block the source range hi flux signal and. received a reactor trip. A post trip review was performed and all the equipment required to ensure a complete reactor trip functioned normally. A startup was reini-l tiated and the reactor was critical by 0908. Maintenance completed i
replacing the ruptured vent 1ine and at 1513 the plant was synchronized to l
the distribution grid. Cause of the ruptured line was attributed to ero-j sion of the carbon steel piping. An inplace program to examine steamline piping for tube wall thinning and replacement will be continued.
i
4.12 LER 85-18 nQ On August 20, 1985 with the plant at 100% power a management review revealed that the fire hose inspection required by pTarit Technical Specification 4.15.2 had been performed outside the required 18 month, +
25% time interval. The surveillance was completed seven days late.
Exceeding the surveillance period resulted from a management oversight when a procedure, written to satisfy two Technical Specifications with dif-ferent surveillance frequencies, was modified consistent with the longer frequency. As corrective action the surveillance procedure will be clarified and the individuals involved counseled on the importance of following administrative directives for procedure modifications.
Inspection and hydrostatic testing of the removed hoses has shown them to be acceptable for continued use; hence, there were no safety implications.
O)
LER 85-19 This LER is being submitted under OTHER in response to IE Bulletin 85-02.
In accordance with the requirements of IE Bulletin 85-02, a test of the Undervoltage Trip Attachment [UVTA] for each of the Reactor Trip Breakers was made to verify that 20 ounces of force margin exists.
Testing was conducted with the plant at 100% power.
Results of the testing revealed that two breakers failed the force margin test at the 20 ounce level; however, the reactor trip breakers were demonstrated to trip on demand and were at all times capable of performing their safety-related function.
In addition, the reactor trip breakers were proven to pass the force margin test at the 16 ounce level as recommended by Westinghouse, the breaker supplier. Additional measures have been implemented beyond those required in the bulletin until the shunt trip modification is completed.
With this repcrt, WPSC will have met all requirement of IE Bulletin 85-02.
4.13 LER 85-20 n
At 2329 on November 13, 1985, while at 100% power, the A train main feed-water control valve failed shut causing a low-low water level in steam generator IA, and subsequent reactor trip. Two of the four cap screws, which hold the valve actuator to the yoke on the the 1A main feedwater control valve, sheared causing the valve to fail shut. The auxiliary feedwater system started coincident with the low-low steam generator level signal, assuring an adequate heat sink for decay heat removal.
Immediately after the trip, recovery actions were followed per E-0, Reactor Trip or Safety Injection, and a post trip review was performed.
Other than the damaged feedwater regulating valve, which failed closed, al,1 equipment responded as designed to the trip. The reactor was critical again at 1752, November 14, 1985. The failure mechanism of the cap screws f) is suspected to be low cycle stress fatigue. The failed cap screws will v
undergo metallurgical testing in attempt to verify low cycle stress fatigue, or identify any other failure mechanisms.
Long term corrective actions will be developed upon completion of the metallurgical evaluation.
LER 85-21 On November 14, 1985, with the plant at 0% reactor power, the Source Range high flux Level Trip setpoint verification test was being conducted prior to reactor startup from the trip of November 13, 1985 (See LER 85-020).
The High Flux at Shutdown bistable trip value was read and recorded in place of the high flux Level Trip setpoint on both source range channels.
An operations department review on November 26, 1985, with the reactor at 100% power, identified the error when it was noted that the recorded values of the Source Range high flux Level Trips corresponded to those of previous High Flux at Shutdown alarm readings.
During the startup of November 14
4.14 1985 the source range response to increasing flux levels was nernal, the p
overlap data for the source and internedtate ranges was norral, the Source U
Range high flux Level Trip was blocked upon receipt of the P-6 permissive, and the balance of the reactor protection system was cperable. During a subsequent reactor startup on Decenber 12, 1985 the sotrrce range high flux Trip was tested, with the as-found values satisfactory. There was no effect on the health and safety of the public as a result of this ever,t.
This event is considered to be an isolated perscnnel error, and the indivi-duals responsible have been counseled on the impcrtar.ce of procedure per-formance. Also, the associated surveillance procedure will be revised to include a lower bound on the Source Range high flux reactor Trip acceptance criterion, such that the value for High Flux at Shutdown will t:e outside the acceptable range.
LER 85-22 n
v On November 27, 1985 at 100% power, Surveillance Procedure SP 48-046,
" Target Band Determination," (Technical Specification 3.10.b.7) was per-formed outside of the required time interval of each effective full power month.
The surveillance was completed 4 1/2 days after the allowable extension to the surveillance interval.
Due to a reactor trip on November i
13, and restart on November 14, the initial conditions for the procedure, which include no change in Xenon greater than 50 pcm for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the flux map, could not be met. Calibration of the Nuclear Irstrumentation
)
following startup also contributed to the delay in performance of the target band determination.
On November 25, 1985 initial conditions for j
performance of the target band determination were satisfied. As the result of a personnel oversight the Target Band Determination was not perforned until November 27, 1985. Failure to complete the surveillance within the j
required time interval, as soon as plant conditions permit, is reportable per 10 CFR 50.73 (a)(2)(1)(B). The results of Sp48-046 showed the target band to be acceptable, hence there were no safety implications.
4.15 LER 85-23 At 0135 on December 12, 1985 with the plant at 100% power, a loss of power to Instrument Bus I occurred. This res~ulted in a partial loss of instrumentation and various alarms.
18 main feedwater control valve also closed due to loss of positioner power. A reactor / turbine trip then occurred due to a Lo Steam Generator Level signal coincident with a Steam Flow /Feedwater Flow mismatch signal on IB Steam Generator. The control operators performed the recovery actions specified in Emergency Procedure E-0, " Reactor Trip or Safety Injection" and a post-trip review was completed. The Auxiliary Feedwater System started as a result of a lo-lo steam generator level signal, assuring an adequate heat sink for decay heat removal. There was no lapact on the health and safety of the public.
Investigation of the inverter for Instrument Bus I revealed the constant t
voltage transformer had failed. The instrument bus was switched to an alternate power supply and the inverter was deenergized. A plant startup was commenced at 0524 on December 12, 1985. At 0711 on December 12 the constant voltage transformer was replaced and the Instrument Bus was I
returned to its normal power supply.
i j
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I
5.1 5.0 FUEL INSPECTION REPORT Thirty six (36) fresh Region M assemblies were loaded for Cycle XI.
I Startup physics testing was performed and reported in the Cycle XI Startup J
Report.
^A The irradiated fuel inspection was performed with an underwater TV camera, All peripheral fuel rods were examined using one-half face scans. Ten assemblies were inspected, including one each in regions G, H, I and J and two each in regions A, K, and L.
All assemblies exhibited rod slippage to various degrees with the majority having rods in contact with the bottom i
nozzle. Numerous scrapes to the rodlets, grids and top and bottom nozzles were also noted. However, no damage to the cladding or supporting struc-tures was observed. All assemblies exhibited axially varying crud depo-sits. The one Region H assembly showed slight rod bowing. Overall condition of the fuel was very good with no evidence of fuel cladding i
4 degradation on the fuel rods examined. Video tapes were made of all exact-nations.
4 1
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6.0 CHALLENGES TO AND FAILURES OF PRESSURIZER SAFETY AND RELIEF VALVES There were no challenges to or faitares of pressurizer safety or relief valves durir.g 1985.
During the 1985 refueling catage the pressurizer safety and relief valve discharge pipirta was wadified. Two rupture disc /biffle-plate assemblies were installed at the discharge of the pressurizar safety valves to 40f in relieving stress following safety valve actuation.
This modification is i
discussed in section 3 of this report.
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7.0
SUMMARY
CF 1985 STEAM GENERATOR EDDY CURRENT EXAMINATION APPLICABLE DEFIhlTIONS:
i Degraded Tube: A tube with greater than a 20%.thru-wall indication.
i
~
Defective Tube: A tube with greater than a 50% thru-wall indicaticni If w..
significant tube thinning has occurred in the are'a pf the i
indication, the defective tube criterion is reduced to i
greater than 40% thru-wall. Defective tubes requira i
- plugging, i
i l
HISTORICAL SWARY GF TUBES FtuGGEO IN TbE KEVAutfEE STEAM GEERATORS Humber of Tubes l
j Plugged in:
Stean Generator LA Stean Generater IB j
1983 23 43
\\
f 1984 8
17 j
i 1985 27 22 TOTAL 58 88 1
i TOTAL AS PERCENT (3388 tubes / generator) 1.7%
2.6%
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7.2 TABLE 7.1
SUMMARY
GF 1985 STEAM GENERATOR EDDY CURRENT EXAMINATION
,o NE INSPECTION EXTENT Steam Generator IA Extent of Inspection Hot Leg Cold leg Full Length 3179 U-Bend 174
- 1 TSP (l) 4 10 TOTALS 3357 10 Steam Generator 1B Extent of Inspection Hot Leg Cold leg Full length 3129 O
U-Bend 188 V
- 7 ISP(l) 1 1
- 2 TSP 2
- 1 TSP 4
9 TOTALS 3322 12 (1) Tube support plates, counted up from the tube sheet inspected.
o
7.3 TABLE 7.2 (Page 1 of 3) 1A STEAM GENERATOR 1985 EDDY CURRENT EXAMINATION O
% THRU-WALL INDICATION (1)
ROW COLUMN PENETRATION PLUGGED LOCATION 5
1 65 X
H: TSP #1 4
2 44 X
H: TSP #1 7
10 47 X
H: TSP #1 17
-11 93 X
H: TE 6
12 83 X
H: TE + 5.8 8
12 92 X
H: TE + 14.5
/
15 13 88 X
H: TE + 9.5 20 17 89 X
H: TE + 10.9 15 20 86 X
H: TE + 9.8 6
21 SQR X
H: TE + 3.9 1
28 46 X
H: TSP #1 11 31 61 X
H: TS + 0.1 7
33 50 X
H: TE + 4.9 23 36(2) 91 X
H: TE + 5.3 5
47 84 X
H: TE + 3.0 23 47 78 X
H: TS + 0.1 24 47 41 X
H: TS 25 47 (3)
X H: TS 27 47 66 X
H: TS + 0.3 28 47 58 X
H: TS + 0.4
()
7.4 TABLE 7.2 (Page 2 of 3) 1A STEAM GENERATOR 1985 EDDY CURRENT EXAMINATION
()
% THRU-WALL ROW COLUMN PENETRATION PLUGGED LOCATION 32 47 78 X
H: TS + 0.2 6
50 89 X
H: TE + 3.5 17 50 57 X
H: TS + 2.3 33 58 84 X
H: TE + 6.0 4
59 SQR X
H: TE + 8.1 3
71 74 X
H: TS + 0.2 13 4
38 H: TSP #1 1
5 14 31 H: TE + 0.8 32 16 31 H: TS 8
19 28 H: TS + 1.1 8
23 24 H: TS + 0.2 11 23 21 H: TSP #1 3
32 23 H: TS + 0.8 36 34 25 H: TS + 45.4 13 42 28 H: TS + 2.6 24 48 21 H: TS + 0.1 8
49 28 H: TS + 0.4 13 50 25 H: TS + 1.5 23 51 22 H: TS + 1.6 18 52 24 H: TS + 1.0 10 56 37 H: TS + 0.6 11 59 30 H: TS + 0.6 11 60 20 H: TS + 0.5 23 65 33 H: TS + 0.9 O
7.5 TABLE 7.3 (Page 3 of 3) 1A STEAM GENERATOR 1985 ED0Y CURRENT INSPECTION
% THRU-WALL INDICATION (1)
R0W COLUMN PENETRATION PLUGGED LOCATION 25 65 30 H: TS + 0.8 8
68 32 H: TS + 0.8 19 70 28 H: TS + 0.5 6
72 36 H: TS + 0.4 18 74 25 H: TS + 0.7 6
75 28 H: TS + 0.2 2
77 39 H: TS + 0.4 6
77 21 H: TS + 0.5 6
80 36 H: TS + 0.6 i
7 83 21 H: TS + 14.0
]
)
24 77 20 1
(1)H - Inspected From Hot Leg C - Inspected From Cold Leg TSP - Tube Support Plate TS - Tube Sheet TE - Tube End SQR -- Squirrel Note that numbers added to TSP, TE, ect., are distances in inches above the indicated landmark in the indicated leg.
(2) Tube R23, C36 was plugged in the hot leg in 1984, the cold leg was plugged in 1985.
(3) Plugged based on 8 x 1 response and bobbin response.
O
7.6 TABLE ^7.3 (Page 1 of 5)
()
IB STEAM GENERATOR 1985 ED0Y CURRENT INSPECTION
% THRU-WALL INDICATION (1)
ROW COLUMN PENETRATION PLUGGED LOCATION 9
19 23 H: TS + 22.9 11 19 46 X
H: TE + 5 l
5 20 37 H: TS + 0.5 8
22 23 H: TS + 0.7 i
9 22 32 H: TS + 0.6 3
23 25 H: TS + 0.3 i
9 23 25 H: TS + 0.5 I
11 25 33 H: TS + 0.6 i
16 26 25 H: TS + 0.8
( )
16 28 30 H: TS + 0.9 16 30 SQR X
H: TE + 10.6 24 30 24 H: TE + 2.9 25 30 48 X
H: TE + 3.1 16 31 44 X
H: TS + 1.3 l
22 31 52 X
H: TE + 4.7 28 31 37 H: TS + 1.1 23 32 25 H: TE + 3.5 24 32 26 H: TE + 3.5 27 32 21 H: TE + 3.1 28 32 31 H: TE + 3.6 5
33 29 H: TE + 4.3 16 34 20 H: TS + 1.9
()
23 34 29 H: TE + 2.9 30 34 25 H: TE + 4.5
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TABLE 7.3 (Page 2 of 5)
^
18 STEAM GENERATOR 1985 E00Y CURRENT INSPECTION i
% THRU-WALL INDICATION (1)
R0W COLUMN PENETRATION PLUGGED LOCATION 16 36 24 H: TS + 2.3 3
37 23 H: TE + 3.2 l
23 37 28 H: TE + 5.4 24 38 51 X
H: TE + 5.0 I
30 38 37 H: TS + 0.8 l
31 39 22 H: TS + 0.2 32 39 27 H: TE + 3.9 15 40 39 H: TS + 4.7 1
l 30 40 70 X
H: TS + 0.8 8
41 73 X
H: TE +'3.8 25 41 28 H: TE + 3.5 8
42 30 H: TS + 40.9 13 42 SQR X
H: TE + 4.4 l
15 42 29 H:
TS'+ 4.3 16 42 33 H: TS + 4.6 i
l 23 42 83 X
H: TE + 3.1 i
l 30 42 31 H: TS + 2.6 10 43 23 H: TS + 0.4 16 43 31 H: TE + 4.2 21 43 22 H: TS + 0.5 32 43 22 H: TS + 0.9 5
13 44 29 H: TS + 2.6 0
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r c' e---*****w
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TABLE 7.3 (Page 3 of 5)
Os 1B STEAM GENERATOR 1985 EDDY CURRENT INSPECTION
% THRU-WALL INDICATION (1)
RDW COLUMN PENETRATION PLUGGED LOCATION 26 44 24 H: TE + 4.8 14 45 25 H: TS + 2.6 27 45 59 X
H: TE + 3.4 28 45 29 H: TE + 5.0 32 45 27 H: TS + 0.0 15 46 SQR X
H: TE + 4.4 30 46 40 X
H: TE + 3.5 6
47 37 H: TS + 0.9
()
10 47 35 H: TS + 1.7 14 47 28 H: TS + 4.6 24 48 29 H: TS + 3.5 27 48 52 X
H: TS + 5.0 30 48 23 H: TS + 3.3 33 48 33 H: TS + 0.8 15 49 21 H: TE + 3.5 17 49 25 H: TS + 4.4 31 49 30 H: TS + 2.8 33 49 22 H: TS + 0.9 25 50 20 H: TE + 5.3 31 50 21 H: TE + 4.5 1
51 85 X
H: TE + 4.2 O
.- - _ ~. --
7.9 1
TABLE 7.3 (Page 4 of 5) l 1B STEAM GENERATOR 1985 ED0Y CURRENT INSPECTION i
% THRU-WALL
-INDICATION (1)
. ROW COLUMN PENETRATION PLUGGEO LOCATION i
f 27 51 44 X
H: TE + 4.2 l
31 51 40 X
H: TE + 3.4 j
29 52 52 X
H: TS + 5.3 l
32 52 29 H: TS + 0.0 3
53 30 H: TS + 0.9 i
1 21 53 22 H: TE + 4.9 I
~23 53 29 H: TE + 4.1 33 54 28 H: TS + 0.4 33 56 28 H:
'S + 0.0 14 58 39 H: TS + 1.1 l
23 58 24 H: TE + 5.0 i
27 58 29 H: TE + 3.5 33 58 21 H: TS + 1.5 j
25 59 35 H:
TE.+ 2.8 33 59 22 H: TS + 2.6 2
60 75 X
H: TE + 3.6 33 60 26 H: TS + 1.7 26 61 31 H: TE + 4.1 i
23 63 24 H: TS + 1.6-27 63 25 H: TS + 0.2 17 64 20 H: TS + 1.0 j
37 64 34 H: V-4 + 0 15 72 30 H: TE + 4.5 1
i
7.10 TABLE 7.3 (Page 5 of 5)
'x IB STEAM GENERATOR 1985 E00Y CURRENT INSPECTION
% THRU-WALL INDICATION (1)
ROW COLUMN PENETRATION PLUGGED LOCATION 13 73 24 H: TS + 1.4 4
75 23 H: TS + 0.9 33 76 36 H: TSP #1 + 0.0 15 4
28 CL: TSP #2 + 0.0 20 7
28 CL: TSP #2 + 0.0 44 44 21 CL: TSP #6 + 0.0 8
45 26 CL: TSP #3 + 5.4 l
8 62 44 X
CL: TSP #6 + 0.0 39 64 53 X
CL: TSP #5 + 0.0
()
39 65 24 CL: TSP #6 + 0.0 35 72 29 CL: TSP #6 + 0.0 33 73 38 CL: TSP #7 + 0.0 35 73 20 CL: TSP #7 + 0.0 33 76 30 CL: TSP #7-e 0.0 16 88 30 CL: TSP #2 + 0.0 19 89 28 CL: TSP #1 + 0.0 (1)H - Inspected from Hot Leg C - Inspected from Cold Leg TSP - Tube Support Plate TE - Tube END TS - Tube Sheet SQR - Squirrel Note: Numbers added to TSP, TE, etc., are distances in inches above the indicated landmark, in the indicated leg.
8.1 8.0 PERSONNEL EXPOSURE AND MONITORING REPORT Pursuant to 1C FR20.407(a)(2), and 20.407(b), a tabulation of the number of
.O 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 1985 Exp. Range (mR)
No. of Personnel No Measurable 331 100 162 100 - 249 113 250 - 499 99 500 - 749 75 750 - 999 39 1000 - 1999 30 2000 - 2999 0
3000 - 3999 1
4000 - 4999 0
5000 - 5999 0
6000 - 6999 0
7000 - 7999 0
8000 - 8999 0
9000 - 9999 0
10000 - 10999 0
11000 - 11999 0
Grand Total 850 GV T
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8.2 Table 8.2 4
-TOTAL NUMBER OF CONTRACTORS PROVIDED WITH PERSONAL DOSE MONITORING DEVICES O
Exp. Range (mR)
No. of Personnel I
No Measurable 231 1
100 85 i
100 - 249 73 1
5 l
250 - 499 55
)
i 500 - 749 51 i
750 - 999 27 1000 - 1999 22 2000 - 2999 0
i j
j 3000 - 3999 0
l Total 544 1
i i
Table 8.3 i
TOTAL NUMBER OF WPSC PLANT STAFF PROVIDED WITH PERSONAL DOSE MONITORING DEVICES i
j Exp. Range (mR)
No. of Personnel l
No Measurable 59 i
i 100 42 i
l 100 - 249 32 250 - 499 36 500 - 749 21 j
750 - 999 11 1000 - 1999
'7 2000 - 2999 0
3000 - 3999 1
j Total 209 4
1 l
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8.3 Table 8.4 TOTAL NUMBER OF PERSONNEL (WPSC NON-PLANT STAFF) PROVIDED WITH PERSONAL DOSE MONITORING DEVICES 4
Exp. Range (mR)
No. of Personnel No Measurable 41 100 35 100 - 249 8
250 - 499 8
500 - 749 3
750 - 999 1
1000 - 1999 1
2000 - 2999 0
3000 - 3999 0
Total 93 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 Kewaunee Nuclear Power Plant Technical Specification. The table shows the total man-rem exposure for the year was 175.995.
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9.1 i
9.0 RADIOLOGICAL MONITORING PROGRAM i
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j Attached is the report from Teledyne Isotopes on the Radiological 4
Monitoring Program for Kewaunee Nuclear Plant for 1985.
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T TELEDYNE ISOTOPES MIDWEST LABORATORY 1509 FRONTAGE RD.
NORTHBROOK, IL 600624197 (312)564 4700 REPORT TO WISCONSIN PUBLIC SERVICE CORPORAiiON WISCONSIN POWER AND LIGHT COMPANY MADISON GAS AND ELECTRIC COMPANY RADIOLOGICAL MONITORING PROGRAM FOR THE KEWAUNEE NUCLEAR POWER PLANT KEWAUNEE, WISCONSIN ANNUAL REPORT - PART I
SUMMARY
AND INTERPRETATION g
January - December 1985 PREPARED AND SUBMITTED BY TELEDYNE IS0 TOPES MIDWEST LABORATORY PROJECT N0. 8002 Approved by:
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L. 6. Huebner U
General Manager O
7 February 1986
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l 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.
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'd TABLE OF CONTENTS Page Preface............................
ii List of Figures........................
iv List of Tables v
- 1. 0 INTRODUCTION 1
2.0 S UPN AR Y............................
2 3.0 RADIOLOGICAL SURVEILLANCE PROGRAM 3
3.1 Methodology 3
3.1.1 The Ai r Program..................
3
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3.1.2 The Terrestrial Program..............
4 3.1. 3 The Aquatic Program................
5 3.1.4 Program Execution..................
6 3.1.5 Program Modifications...............
6 3.2 Results and Discussion..................
7 3.2.1 Atmospheric Nuclear Detonation 7
3.2.2 The Air Environment...
7 3.2.3 The Terrestrial Environment............
9 3.2.4 The Aquatic Environment.
12 4.0 FIGURES AND TABLES 15
5.0 REFERENCES
32 APPENDICES A.
Interlaboratory Comparison Program Results.........
A-1 j
B.
Statistical Notations...................
B-l' C.
Maximum Permissible Concentrations of Radioactivity in Air and Water above Natural Background in Unrestricted Areas C-1 O
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LIST OF FIGURES No.
Caption Page 4-1 Samplirig locations, Kewaunee Nuclear Power Plant.......
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(>T LIST OF TABLES No.
Title Page 4.1 Sampling locations, Kewaunee Nuclear Power Plant.
17 4.2 Type and f requency of collection...............
18 4.3 Sample codes used in Table 4.2................
19 4.4 Samp l i ng Sumary.......................
20 4.5 Environmental Radiological Monitoring Program Sumary.....
21 In addition, the following tables are in the Appendix:
Appendix A A-1 Interlaboratory Comparison Program Results, milk, water, air filters, and food sanples, 1982-1985 A-3 A-2 Interlaboratory Comparison Program Results, thermoluminescent 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
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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 i
generation was achieved on April 8,1974, and the Plant was declared commer-cial on June 16, 1974.
This report sumarizes the environmental operation data collected during the period January - December 1985.
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 dp them.
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SUMMARY
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Results' of. sample analyses during the period January - December 1985 are s
summarized in Table 4.5.
Radionuclide concentrations measured at indicator l
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 1.
Trace amounts of. cobalt-58, cobalt-60, and cesium-137 were detected in several bottom sediment samples collected in the discharge area (K-1c, K-1d, K-lj)'
i and Two Creeks Park (K-14), averaging 0.062, 0.062, and 0.036 pCi/g dry weight above background level for j
cobalt-58, cobalt-60, and cesium-137, respectively. At i
the same time, the gross beta concentration averaged 2.8 pCi/1.. Assuming that all gross beta activity was
}
due to these three isotopes, cobalt-58 and cobalt-60 would contribute 1.1 pCi/l each and cesium-137 would i
j contribute 0.6 pCi/l of gross beta concentration.
1 These concentrations constitute only 0.0011%, 0.0022%,
and 0.003% of the maximum permissible concentrations of 4
100,000 pCi/1, 50,000 pCi/1, and 20,000 pCi/l for t
cobalt-58, cobalt-60, and cesium-137, respectively, established in 10 CFR 20 Document.
The presence of these isotopes in-bottom sediment samples is probably plant-related, but the levels are j
insignificant.
I 2.
Nine samples collected at discharge -(K-ld) and six l
samples collected at Two Creeks Park (K-14) ~ had elevated tritium levels.
The annual mean tritium concentration at the discharge was 540 pC1/1 above j
background level.
The highest ' concentration was i
~
measured in the sample collected at discharge (K-1d) on February -2,1985 and measured 1920 pC1/1 above background level.
The presence of tritium in the discharge water is. attributable to the Kewaunee Nuclear Plant operation, but the highest discharge rate measured constitutes only 0.07% of the maximum permissible concentration-of 3,000,000 pCi/l estab-j lished in the 10 CFR 20 Document.
The annual mean tritium concentration in lake water j
collected at Two Creeks Park 'was 50 pCi/1 above back-ground level and the maximum was measured in the sample collected on February-4,.1985 (130 pCi/1, which constitutes about 0.004% of the permissible level).
However, because of the associated counting error, a concentration of this low magnitude is
)
indistinguishable from the background level.
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- 3. 0 RADIOLOGICAL SURVEILLANCE PROGRAM Following is a description of the Radiological Surveillance Program and its execution.
4 1
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.
1 The sampling program monitors the air, terrestrial, and aquatic envi-l 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.
1 3.1.1 The Air Program Airborne Particulates w
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-lf, K-2, K-7, K-8, K-15, and K-16), and dis-patched by mail to TIML for radiometric analysis.
The material on the filter is counted for gross alpha and 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.
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Airborne Iodine Charcoal filters are located at locations K-1f, K-2, K-7, K-8, K-15, and K-16.
The filters are changed bi-weekly and analyzed for iodine-131 immediately after arrival at the laboratory.
Ambient Gama Radiation - TLDs lne integrated gama-ray background is measured at six air sam-pling locations (K-lf, K-2, K-7, K-8, K-15, and K-16) and at four milk sampling locations (K-3, K-4, K-5, and K-6) with thermo-luminiscent dosimeters (TL0s),
CaF :Mn bulb TLDs are exchanged 2
quarterly and annually.
Precipitation Monthly composites of precipitation sagles collected at K-ll are analyzed for tritium activity by liquid scintillation technique.
3.1.2 The Terrestrial Program Milk Milk sa@les are collected weekly (one gallon from each location) i 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, K-5, K-12, and K-19) and from two herds that graze between four and ten miles from the reactor site (K-3 and K-6).
The milk samples are analyzed for iodine-131, strontium-89 and
-90, ce si um.-137, barium-140, potassium-40, calcium, and stable potassium.
Well Water One gallon water samples are collected quarterly from four off-site wells located at K-10, K-11, K-12, and K-13.
Monthly one-gallon water samples are collected from two on-site wells located at K-1g and K-lh.
The-gross alpha and beta activities are determined on the total.
residue of each water sample... The concentration of potassium-40 is calculated from total potassium, which is determined by flame photometry on all samples.
The tritium levels in quarterly composites of monthly on-site samples from K-lg are determined by I
liquid scintillation technique.
Quarterly composites of monthly grab samples of water from one on-site well (K-1g) are analyzed for strontium-89 and strontium-90.
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Domestic Meat (v]
Domestic meat samples (chickens) are obtained annually (in the third quarter) at locations K-20, X-24, K-25, and K-27.
The flesh is separated from the bones, gama scanned, and analyzed for gross alpha, gross beta, strontium-89, and strontium-90 activities.
5 1 9. 9. 5 Eggs are collected quarterly at Location K-27.
The samples are gama scanned and analyzed for gross alpha, gross beta, strontium-89, and strontium-90 activities.
Vegetables Vegetable samples (5 varieties) are collected at locations K-17 and K-26, and two varieties of grain, if available, at location K-23.
The samples are gamma scanned and analyzed for gross alpha, 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 p
six dairy farms (K-3, K-4, K-5, K-6, K-12, and K-19).
The h
samples are gama scanned and analyzed for gross alpha, gross beta, strontium-89, and strontium-90 activities.
During the first quarter cattle feed is collected from the same six dairy f arms, and the same analyses are perforned.
Soil Soil samples are collected twice a year on-site at K-1f and f rom 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 f rom 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, p
K-lb) are collected near the mouth of each creek.
Samples from Q
the South Creek (K-le) are collected about ten feet downstream from the point where the outflows from the two drain pipes meet.
5
i The water sagles are analyzed for gross alpha and gross beta activity in the total residue, dissolved solids, and suspenoed solids.
The concentration of potassium-40 is calculated from j
total potassium, which is determined by flame photometry.
The tritium activity in the take Michigan samples is determined by liquid scintillation technique.
Quarterly conposites of monthly grab sa@les from Lake Michigan are also analyzed for strontium-89 and strontium-90.
D Fish samples are collected in the second, third, and fourth quarters 'at Location K-Id.
The flesh is separated from the bones, gama scanned and analyzed for gross alph3 and gross beta activity.
Ashed bone samples are analyzed for gross alpha, gross beta, strontium +89 and strontium-90 activities.
Slime 4
Slime samples are collected during the second and third quarters from three Lake Michigan locations (K-Id, X-9, and K-14), and from three creek locations (K-lo, K-lb, and K-le), if av611able.
The samples are analyzed for gross alpha and gross beta activi-ties. If the quantity is sufficient, they are also gama scanned and analyzed for strontium -89 and strontium.90 activ-ities.
k Bottom Sediments Bottom sediments are collected four times a year from five locations (K-Ic, X-Id, K-lj, K-9, and K-14).
The samples are analyzed for gross alpha and gross beta activities and for strontium-89 and strontium-90.
Each sample is also gamma 1
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.
s 3.1.4 Pr[igramExecution Program execution is summarized in Table 4.4, The program was executed as described in the preceding sections with 'one exception.
There were no TLD data for Location K-5 for the
+
fourth Quarter of 1985 because the TLD holder was lost in the snow.
An attempt was made to find it but was not successful.
J 3.1.5 Procram Modifications There were no program modifications during 1984.
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34 2 Results and Discussion 1
The results for the reporting period January to December 1985 are presented in summary form in Table 4.5.
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.
i The discussion of the results has been divided into three broad cate-gories:
the air, terrestrial, and aquatic environments.
Within each category, sanples 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 Teledyne Isotopes Midwest Laboratory or its predecessor, Hazleton Environmental Sciences.
The tabulated results of all measurements made in 1985 are not included in this section, although references to these results will be made in the discussion.
The complete tabulation of ' the 1985 results is contained in Part II of the 1985 annual report on the Radiological i
Monitoring Program for the Kewaunee Nuclear Power Plant.
l 3.2.1 Atmospheric Nuclear Detonations There were no reported atmospheric nuclear tests in 1985.
The l
last reported test was conducted by the People's Republic of China on October 16, 1980.
The reported yield was in the 200 j
kiloton to 1 megaton range.
4 l
3.2.2 The Air Environment 1
i Airborne Particulates
}
For air particulates, both gross alpha and gross beta measure-l ments yielded annual means that were either identical (gross alpha) or nearly identical (gross beta) for the indicator. and j
control locations.
Mean gross alpha and gross beta concentra-tions were slightly lower than in 1984.
The highest annual l
means, which were close to the average means, for gross alpha and gross beta were measured at control location K-16, 26 miles NW of the station, and at indicator location K-1f, 0.12 miles S l
of the station, respectively.
Gross alpha and beta concentrations at all -lecations were also i
analyzed by ouarters.
The activity was higher in the first quarter, declined during the second quarter, and rose slightly i
during the third and fourth quarters.
There was no clear cut l
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/U evidence of the spring peak, which has been observed alcost annually (1976 and 1979 were axceptions) for many years (Wilson et al.,1969). The spring peak has been attributed to f allout of nuclides from the stratosphere (Gold et al.,1964).
Gamma spectroscopic analysis of quarterly cociposites of air particulate filters yielded similar results for indicator and control locations.
Beryllum-7, which is produced continously in the upper atmosphere by cosmic radiation ( Arnold and Al-Salt h, 1955), was detected in seventeen of twenty-four samples and was the only gama-emitting isotope detected.
There was no indica-tion of a station effect on the data.
All other gama-emitting isotopes were below their respective LLO liriits.
Airborne Iodine Bi-monthly levels of airborne tooine-131 were below the lower limit of detection (LLO) of 0.01 pCi/m3 at all locations.
Thus, there is no indication of an effect of the plant operation on the local air environment.
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Ambient Gamma Radiation - TLDs Ambient gama radiation was monitored by TL0s at ten locations:
four indicator and six control.
The quarterly TL0s at the indicator locations measured a mean dose equivalent of (64.518.2)* mR/365 days, in agreement with the mean at the control locations of (62.016.4) mR/365 days, and were slightly higher than the means obtained in 1984 (53.8 and
- 51. 7 mR/365 days, respectively).
The quarterly measurements agreed within the error with the annual measurements which were (56.816.5) mR/365 days, for the indicator and (59.014.8) mR/365 days for the control locations.
All these values are slightly lower than the United States average value of 78 mR/ year due to natural background radiation (National Council on Radiation Protection and Measurements, 1975).
The highest means for the quarterly and annual TL0s were 75.4 and 66.3 mR/365 days and occured at indicator location K-7.
f
- Unless otherwise indicated, uncertainties of average values are standard n
deviations of the individual measurements over the period averaged.
Uncer-U tainties of individual measurements represent probable counting errors at the 95% confidence level.
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Precipitation Precipitation was monitored only at an indicator location, K-11.
Tritium was detected in five samples and averaged 170 pCi/1.
This level of activity is expected-in the precipitation and is attributable to the recycling of tritium produced by the previous nuclear tests in the atmosphere.
3.2.3 The Terrestrial Environment Hilk Of the 192 analyses for iodine-131 in milk all were below the LLD d
2 level of 0.5 pCi/1.
i Strontium-89 concentration was below the LLO level of 2.4 pCl/l
^
in all samles.
Strontium-90 was found in samples.
The mean values were nearly identical for indicator and control locations (2.1 pCi/l and 2.2 pCi/1, respectively).
Barium-140 concentration was below th;'LLD of 10 pCi/l in all O
samples.
Cesium-137 concentration was also below the LLD of 10 l
pCi/l in all samples.
Potassium-40 results were nearly identical at both the indicator and control locations and were essentially identical to the i
levels observed in 1978 through 1984.
Due to the chemical similarities between strontium and calcium, and cesium and potassium, oroanisms. tend to deposit cesium-137 in the sof t 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
'i environmental accomulation of these radionuclides.
No statis-tically significant variations in the ratios were observed.
The neasured concentrations of stable potassium and calcium are in agreement with previously determined values of 1.50f0.21 g/l and 1.1610.08 g/1, respectively (National Center for Radiological i
Health,1968).
s Well Water
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Gross alpha concentration in well water was' below the LLD level of 2.9 pCi/l in all samples.
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Gross beta concentration in well water was l'.6 pC1/1 in samples 4
from the control location.
The mean value for all indicator locations was 2.'6 pC1/1 and was nearly identical to the values observed. in.1977, 1978, 1979, 1980,1981,1982,1983, and 1984 i
( 3. 3 p C1/1, 3. 4 p e i /1, 3. 0 p C l /1, 3. 0 p Ci /1, 3,6 p Ci /1, 3. 2 j
pCi/1, 2.9 pCi/1, and 2.3 pCi/1, respectively).
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Tritium concentration in the on-site well (K-Ig) was below the LLO of 100 pCi/1 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 3.1 pCf/1), in agree-ment with the previously meas.ured values.
j Domestic Neat In meat (chickens), gross alpha concentration was similar at both i
indicator and gootrol locations (0.15 and 0.12 pCi/g wet weight, respectively).
Gross beta concentration averaged 2.69 pCi/g wet weight for indicator locations and 2.59 pCi/g wet weight for control locations. Gama-spectroscopic analysis showed that most of the beta activity was due to naturally occurring potassium-40.
i All other gamma-emitting isotopes were below their respective LLD limits.
I i
Si.9.s, In egg samples, the gross alpha concentration averaged 0.07 pCi/g wet weight.
Gross beta concentration averaged 1.04 pCi/g _
wet weight, about equal to the concentration of the naturally-4 occurring potassium-40 observed in tr.e - saaples (1.07 pC1/g).
3 All other gamma-emitting isotopes were below their respective LLO's.
The level of strontium-89 was below the LLD of 0.004
+
pCi/g wet weight.
Strontium-90 was detected in one sample and was at the LLD level of 0.002 pCi/g wet weight.
Vegetables i
In vegetables, alpha concentration averaoed 0.25 and 0.15 pC1/g wet weight in indicator and control samples, respectiYely.
Gross beta concentration was slightly higher at the indicator location than at the control location ard was due primarily to.
l the potassium-40 activity.
Strontium-89 activity was below the <
LLO of 0.004 pCi/g wet weight in all samples.
Strontium-90 activity was nearly identical at the control location and the indicator location (0.005 pCi/g wet weight and 0.004 pCi/g _ wet 4
g weight, respectively).
All other gama-emitting isotopes were below. their+ respective LLO levels.
i 10 i
s y
6 v-.--.
--y,,
,-,,,,,,,y--
p
..,.,..y,---
,__,m
,-g,,,,,m,.v-,,
,,,c-,w7.---7,,,..--.
,w
O)
I below their respective LLD levels.,^ 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 the grain in comparison with the vegetables.
Grass and Cattle Feed In grass, gross alpha concentration was essentially identical at both indicator and control locations (0.29 and 0.27 pCi/g wet weight, respectively).
Gross beta concentration was slightly higher at indicator ; locations (6.09 pCi/g wet weight) than at the control locations (5.[30 pCi/g wet weight) 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 0.02 pCi/g wet weight in all samples..: Strontium-90 activity was detected in all samples and was slightly higher at indicator than at control locations' (0.028 and 0.025 pCi/g wet weight, respectively).
Presence of radiostrantium in grass samples is attributed to the
+
fallout from the previous nuclear tests.
For cattlefeed, the mean gross alpha en 6tration was identical m
at both indicator and control loc 9tbns (0.9 pCi/g wet weight).
Mean gross beta concentration was slightly higher at control locations (8.0 pCi/g wet weight) than at indicator locations (7.6 pCi/g wet weight).
The highest gross beta level was in the sample. from indicator location K-4'(16.5 pCi/g wet weight), and reflected the high potassium-40 level (17.2 pCi/g wet weight) observed in the sample.
The pattern was similar to that observed in 1978 through 1984.
Strontium-89 levels fere -below the LLD level of 0.14 pCi/g wet weight in all sam les.
Strontium-90 activity was lower at indicator locations than at control loca-tions (0.075 and 0.141 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.
1 Soil No significant differences were found between indicator and control values in soil samples..The difference of 0.7 pCi/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 (23.2 and 22.7 h.U) pCi/g dry weight, respectively), and is primarily due to the potassium-40 activity.
Strontium-89 was below the LLD level of s
(
11
,y
~
w
-.3c,
.ye--
i l
O 0.11 pCi/g dry weight in all samples.
Strontium-90 was detected in thirteen of fourteen samples and was higher at control than at indicator locations (0.33 and 0.09 pCi/g dry weight,respectively). Cesium-137 was detected in twelve samples and was higher at control locations than at indicator locations (0.46 and 0.29 pCi/g dry weight, respectively).
All other gansna-emitting isotopes were below their respective LLD's.
The levels of detected activities were similar to those observed in 1979 through'1984.
3.2.4 The Aquatic Environment Surface Water In surface water, the gross alpha concentration in suspended solids was below the LLD of 1.0 pCi/l in all samples.
In dis-solved solids, gross alpha activity was detected in two of seventy-two samples and averaged 4.5 pCi/1.
Mean gross beta activity in suspended solids was detected in four samples and averaged 0.6 pCi/1, barely above the detec-tion limit of 0.5 pCi/1.
Mean gross beta concentration in dissolved solids was higher by a factor of two at indicator locations (5.6 pCi/l) as compared to the control. locations (2.5 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 2.4 pCi/l),
1983 (5.1 and 2.6 pCi/1), and 1984 (5.0 and 2.7 pCi/1).
The control sample is the Lake Michigan water which varies very little in concentration during the year,~while indicator samples include two creek locations (K-la and K-le) which are much higher in concentrations and exhibit large month-to-month variations in gross beta concentration.
The K-la creek drains its water from the surrounding fields which are heavily fertilized and K-le creek draws its water mainly from the Sewage Treatment Pond No.
1.
In genaral, gross beta concentration levels' were high when potassium-40 levels were high and low when potassium-40 levels were low indfcating that the fluctuations in beta concentration were due to variations in potassium-40 concentrations and not to plant operations.
The f act 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 500 pCi/l at indicator locations and was below LLD of 220 pC1/1 at control locations.
The mean concentration at the discharge (K-Id) was 540 pCi/l above the background level of 220 pCi/l and 50 pCi/l above the background level at Two Creeks Park, located 2.5 miles south O
of the plant.
The elevated annual mean of 540 pC1/1 above 1
V background in the discharge water is attributable to the plant 12
operation, but constitutes less than 0.02% of the maximum per-missible concentration of 3,000,000 pCi/l established in the 10 CFR 20 Document.
The highest level of 1920 pCi/l above background level detected in the sample collected February 2, 1985 at the condenser discharge constitutes less than 0.07% of the permissible level.
The highest level measured at Two Creeks Park was 130 pCi/l above background level and constitutes about 0.004% of the permissible level.
However, because of the associated counting error, a corcentration of this low magnitude is indistinguishable from the bnkground level.
Strontium-89 activity was below the LLD of 1.8 pCi/l in all samples.
Strontium-90 activity was detected in one of twelve samples and was 1.2 pCi/1.
Fish In fish samples, gross alpha concentration averaged 0.11 pCi/g wet weight in muscles and was below detection limit in all bone fractions.
In muscle, gross beta concentration was primarily due A
to potassium-40 activity.
The average beta concentration of 2.61 h
pCi/g wet weight was near the everage of the 1973 range of 2.26 to 3.62 pCi/g wet weight.
The cesium-137 concentration in muscle averaged 0.14 pCi/g wet weight and was nearly identical to the level observed in 1979 and 1980 (0.12 pCi/g wet weight in both years),1981 (0.15 pCi/g wet weight), in 1982 (0.17 pCi/g wet-weight),in 1983 (0.14 pCi/g wet weight), and in 1984 (0.10 pCi/g wet weight).
The strontium-89 was below the LLD of 0.10 pCi/g wet weight in all but one sample Strontium-90 was detected in all samples and averaged 0.18 pCi/g wet weight.
Periphyton (Slime)
In periphyton (slime) samples, gross alpha concentration was nearly identical at both indicator and control samples (0.97 and 1.18 pCi/g wet weight, respectively).
Mean gross beta concentration was lower at indicator than at control locations (1.69 and 2.87 pCi/g wet weight, respectively).
Strontium-89 concentration was below the LLD level of 0.046 pCi/g wet weight in all samples. Strontium-90 concentrations. was nearly identical at both endicator and control locations averaging 0.044 and 0.046 pCi/g wet weight, respectively.
-A trace quantity of Co-58 (0.078 pCi/g wet weight) was detected in one sample and trace quantities of Co-60 (mean 0.096 pCi/g wet weight) were detected in three samples collected at indicator location K-14.
All other gamma-c I
l emitting isotopes, except naturally-occurring beryllium-7 and d
potassium-40, were below their respective LLDs.
13
O Bottom Sediments In bottom sediment samples, gross alpha levels were below the LLD of 3.4 pCi/g dry weight in all samples.
The mean gross beta concentration was slightly higher at indi-cator locations than at the control location (6.1 and 6.0 pCi/g dry weight, respectively) and was due mostly to potassium-40.
The difference is not statistically sigrfificant.
Cesium-137 was detected in eighteen of twenty samples and aver-aged 0.05 pCi/g dry weight.
The level was. slightly lower than i
the levels observed in 1979 (0.12 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 pCi/g dry weight),.and in 1984 (0.07 pCi/g dry weight).
Strontium-89 and strontium-90 levels were below their respective LLDs (0.032 and 0.015 pCi/g dry weight, respectively) in'all samples.
Trace amounts of cobalt-58 (six samples, mean 0.062 pCi/g dry weight) and cobalt-60 (ten samples, mean 0.062 pCi/g dry weight) were detected near the condenser discharge.
The presence of trace amounts of these activation products in bottom sediments is probably plant related.
O
$ 0 14
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4.0 FIGURES AND TABLES i
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Kewaunee
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Green Eey (K-I6)
'. $'y K-15
- 25 Miles K-24M
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i K-22H I Kawaunce Co.
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\\d SCALE IN MILES O
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3 4
Figure 1.
Sampling locations, Kewaunee Nuclear Power Plant.
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O Table 4.1 Sampling locations, Kewaunee Nuclear Power Plant.
Distance (miles)b Code Typea and Sector Location K-1 Onsite la I
0.62 N North Creek lb I
0.12 N Middle Creek Ic I
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 lg I
0.06 W South Well lh I
0.12 NW North Well lj I
0.10 S 500' south of condenser discharge K-2 C
9.5 NhE WPS Operations building in Kewaunee K-3 C
6.0 N Lyle and John Siegmund farm, Route 1, Kcwaunee K-4 I
3.0 N Dan Stangel farm, Route 1, Kewaunee K-5 I
- 3. 5 NNW Ed Paplham farm, Route 1, Kewaunee K-6c C
6.5 WSW Leonard Berres f arm, Route 1, Denmark K-7 I
2.75 SSW Earl Bruemmer f arm, Route 3, Two Rivers K-8 C
- 5. 0 WSW Saint Mary's Church, Tisch Mills q
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 f arm, Kewaunee site K-11 I
1.0 NW Harlan Ihlenfeld f arm K-12 I
1.5 WSW Lecaptain f arm, 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 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 riles south of "BB")
K-19 I
1.75 NNE Wayne Paral f arm, Route 1, Kewaunee K-20 I
2.5 N Carl Struck farm, Route 1, Kewaunee K-23 I
0.5 W 0.5 miles west of plant, Kewaunee Site
.K-24 I
5.45 N Fectum f arm, Route 1, Kewaunee K-25 C
2.75 WSW Wotachek farm, Route 1, Denmark K-26d 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 a I = indicator; C = control.
b Distances are measured from reactor stack.
c The K-6 sampling location was changed on October 17, 1980 because the operator of Berres Farm retired.
Berres Farm has been replaced by feviteki Farm, located 0.2 miles West of Berres Farm.
d Location K-18 was changed because the Schmidts Food Stand went out of business and was replaceu by Bertler's Fruit Stand (K-26).
.1
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O Table 4.2 Type and frequency of collection.
Frequency Location Weekly Bi-weekly Monthly Quarterly Semi-Annually Annually K-1 K-la SW SL K-lb SW GRa SL K-lc BSb K-ld SW BSb FIa SL K-le SW SL K-lf AP AI GRa TLD.
50 TLD K-lg WW K-lh WW K-lj BSb K-2 AP AI TLD TLD K-3 Mic GRa TLD CF '
50 TLD K-4 MIc GRa TLD CF-50 TLD K-5 MIc GRa TLD CFd S0 TLD K-6 MIc GRa TLD CFd 50 TLD K-7 AP Al TLD TLD K-8 AP Al TLD TLD K-9 SW BSb SL m
K-10 WW K-ll PR
-WW K-12 Mic GRa CFd wy so K-13 WW K-14 SW BSb
-SL K-15 AP Al TLD TLD K-16 AP AI TLD TLD K-17 VE K-18e K-19 MIc GRa CFd 50 K-20 DM K-23 GRN K-24 DM K-25 DM K-26 VE K-27 EG DM a Three times a year, 2nd ( April, May, June), 3rd (July, Aug., Sept.), and 4th (Oct., Nov., Dec.) quarters, b To be collected in May, July, Sept., Nov.
c Monthly from November through April; weekly from May through October.
d First January February, March e Replace (d by K 26 in summer of 1g8parter only, w
4 4
O
-Table 4.3 Sample codes used in Table 4.2.
Code Description f
l AP Airborne Particulate AI Airborne Iodine i
I TLD Thermoluminescent Dosimeter 1
j PR Precipitation MI Milk
[
WW Well Water s-
]
DM Domestic Meat 3
i EG Eggs l
VE Vegetables
[
GRN Grain l
)
GR Grass i
i' i
i SW Surface Water F1 F> s::
SL Slime s
3 i
i BS Bottom Sediments.
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i 3
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19 i
1
,-,.-,cvm.-,,,-....-y-,--c.,
,,-,v,,,,,,m-y,mm...,
m,,,,,w-,,y%_.m,w,.e,,_._,,,w,,,r.y,.,_%,,_,,.
.m.wyw.,.,-w_,,_,
y.,,,,,
, _. ~.. -
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0:
. Table 4.4.
Sampling summary, January - December 1984.
t Collection Number of Nunber of Sample Type and Number of Samples Samples-Type Frequencya Locations Collected Missed Remarks i
Air Environment Airborne particulates C/W 6
312 0
Airborne iodine C/8W 6
156 0
TLD's C/Q 10 39 1
See text Page 6.
C/A 10 10 0
Precipitation C/M 1
12 0
Terrestrial Environment Milk (May-Oct)
G/W 6
156 0
i (Nov-Apr)
G/M 6
36 0
l Well water G/M 2
24 0
l G/Q 4
16 0
l Domestic meat G/A 4
4 0
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 i
Surf ace wa?er G/M 6
72 0
Fish G/TA 1
5 0
l Slime G/SA 6
12 0
l Bottom sediments G/FA 5
20 0
a Type of collect' ion is' coded as follows:
C = continuous; G = grab.. Frequency is coded as follcws:
W = weekly; M = monthly; Q =. quarterly; SA '= semi-annually; TA = three times per year; FA = four times.
per year; A = annually; BW = bi-weekly.
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J Table 4.5 Environmental Radiological Monttoring Program Sumary.
Name of facility Kewaunee Nuclear Power Plant Docket No.
50-305 Location of facility kewaunee Count y, Wisconsin Reporting Period January - December 1985 (County, State) a Indicator Location with Highest Control Sample Type and Locations Annual Mean Locations Nunt>er of Type humber of Mean (F)C Mean (F)
Mean (F)
Non-routine (Units )
Analyses 8 LLDb Range Locationd Range Range Resultse Airborne GA 312 0.0027 0.0035 (88/104)
K-16, Green Bay 0.0042 (41/52) 0.0035 (167/208)
O particulates.
(0.0009 0.0112) 26 mi NW (0.0012-0.0120)
(0.0008-0.0120)
(pCi/m3)
GB 312 0.003 0.016 (103/104)
K-If, Met Tower 0.017 (52/52) 0.015 (199/208) 0
( 0. 003-0.05 M Onsite, 0.12 mi 5 (0.003-0.052)
(0.002-0.045)
G5 24 Be-7 0.022 0.081 (5/8)
K-7, Bruemer Farm 0.095 (3/4) 0.059 (12/16) 0 (0.052-0.143) 2.75 mi SSW (0.063-0.143)
(0.041-0.080)
N Nb-95 0.0051
<LLD (LLD 0
Zr-95 0.0045 (LLD
<LLD 0
Ru-103 0.0037
<tLD
<LLD 0
Ru-106 0,013
'LLD
<LLD 0
Airborne I-131 312 0.01
<LLD 0
Iodine (p Ci/m3)
TLD -Quarterly Gama 39 5
16.1 (15/15)
K-7, Bruemer Farm 18.8 (4/4) 15.4 (24/24) 0 (mR/91 days)
(13.1-22.1) 2.75 mt SSW (15.5-22.1)
(12.7-19.2) 0 TLD-02 arterl y Gama 10 5
64.5 (4/4)
K-7, Bruemer Farm 15.4 (1/1) 62.0 (6/6) 0 (mR/365 days).
(56.2-75.4) 2.75 mi SSW
( 57. 0- 72.2)
TLD Annual Gama 10 5
56.8 (4/4)
K-7, Bruemer Farm 66.3 (1/1) 59.0 (6/6).
0 (mR/365 days)
(52.7-66.3) 2.75 mi SSW (52.5-65.4)
_______._=__.__________=_...m
_ _ _ _ = _ _
m._m_=_,.
.m__._
O O
Table 4.5 Environmental Radiological Monitoring Program Suninary (continued)
Name of Facility Kewaunee Nuclear Power Plant Docket No.
50-305 Location of Facility Kewaunee County, Wisconsin Reporting Period January - December 1985 (County, State)
Indicator Location with Highest Control Sample Type and Locations Annual Mean Locations Number of Type Number of Mean (F)C Mean (F)
Mean (F)
Non-routine (Lttits )
Analysesa Ltpb Range Locationd Range Range Resultse Precipitation H-3 10 100 170 (5/12)
K-11, Inlenfeld Farm 170 (5/12)
None 0
(ptt/1)
(120-230) 1.0 mi NW (120-230) i Milk I-131 192 0.5 (LLD (pCi/1)
(LLD 0
Sr-89 72 2.4 (LLD
<LLD 0
-Sr-90 72
- 0. 6 2.1 (48/48)
K-6, Novitsky Farm 2.5 (12/12) 2.2 (24/24) 0 (0.9-3.7) 6.7 mi WSW
( 1. 7-3. 4 )
(1.4 3.4)
K-12 Lecaptain Farm 2.5 (12/12) 1.5 mi W5W (1.0-3.7)
G5 72 m
K-40 50 1320 (48/48)
K-3, Stangel Farm 1430 (12/12) 1340 (24/24) 0 N
(1050-1540) 3.0 mi N (1330-1550)
(1130-1550)
.10
<tLD
<LLD 0
Ba-140 10 (LLD
<LLD 0
o (g/1)
K-stable 72 1.0 1.51 (48/48)
K-3, Stangel Farm 1.63 (12/12)
'l.53 (24/24) 0 (1.19-1.75)~
3.0 mi N (1.51-1.76)
(1.28-1.75)
(c/1)
Ca 72
- 0. 5 0.8 (48/48)
K-3.-Stangel Farm 0.9 (12/12) 0.9 (24/24) 0 (0.7-1.4) 3.0 mi N (0.7-1.6)
(0.6-1.6)-
K-6, Novitsky Farm 0.9 (12/12)
.6.7 mi W5W (0.6 1.2) i I
i Well Water GA 40
- 2. 9
<tLD -
(LLD 0
G8 40
- 2. 9 2.6 (26/ 36)
K-lh, North Well 4.0 (12/12) 1.6 (4/4) 0 (1.0-5.0)
Chsite, 0.12 mi NW (1.9-5.0)
(1.0-2.1)
H-3 4
100
<LLD None 0
K-40 40 0.10 2.0(36/36)
K-Ig,SouthWeli 2.5'(12/12) 1.2 (4/4) 0
~
(f lame)
(0. 6-3.1 )
Onsite, 0.06 mi W (2.1-3.1)
(1.0-1.4)
Sr-89 4
- 0. 5
<tLD None 0
Sr-90 4
' O. 3
<L L D a
None 0
i e
t-
p s
V v
v l
i i
Table 4.5 Environmental Radiolcgical Monitoring Program Suninary (continued)
Name of Facility Kewaunee Nuclear Power Plant Docket No.
50-305 i
Location of Facility Kewaunee county, Wisconsin Reporting Period January - December 1985 (County, State)
Indicator Location with Highest Control-Sanple Type and Locations Annual Mean Locations Nunter of Type Number of Mean (F)C Mean (F)
Mean (F)
Non-routine (Units)
Analysesa LLDb Range Locationd Range Range Resultse Domestic Meat GA 4
0.04 0.15 (3/3)
K-27, Schlies Farm 0.19 (1/1) 0.12 (1/1) 0 (chickens)
(0.12-0.19) 1.5 mi NW (pct /g wet)
G8 4
0.03 2.69 (3/3)
K-27, Schites Farm 2.79 (1/1) 2.59 (1/l) 0 (2.57-2.79) 1.5 mi NW GS 4
Be-7 0.01
<tLD
<LLD 0
K-40 0.5 3.04 (3/3)
K-27, Schlies Farm 3.20 (1/1) 2.56 (1/1) 0 (2.63-3.20) 1.5 mi NW U
Nb-95 0.011
<LLD
<LLD 0
Zr-95 0.018 (LLD
<LLD 0
Ru-103 0.010
<tL D (LLD 0
Cs-134 0.008 (LLD
<LLD 0
Cs -137 0.007
<tLD
<LLD 0
Ce-141 0.015
<tLD
<LLD 0
Ce-144 0.032 (LLD
<tLD 0
Eggs GA 4
0.03 0.07 (3/4)
K-27 Schlies Farm 0.07 (3/4)
None 0
(pCi/a wet)
(0.05-0.09) 1.5 mi NW (0.05-0.09)
GB 4
0.01 1.04 (4/4) k-27 Schlies Farm 1.04 (4/4)
None 0
(0.86-1.14) 1.5 mi NW (0.86 1.14)
Sr-89 4
0.004 (LLD None O
Sr-90 4
0.002 0.002 (1/4)
K-27, Schlies Farm 0.002 (1/l)
None 0
1.5 mi NW GS 4
Be-7 0.077 (LLD None 0
K-40 0.01 1.07(4/4)
K-27, Schlies Farm 1.07 (4/4)
None 0
(0.84-1.50) 1.5 mi NW (0.84-1.50)
p
,y l
l Table 4.5 Environmental Radiological Monitoring Program Summary (continued)
Name of facility Kewaunee Nuclear Power Plant Docket No.
50-305 l
Location of Facility kewaunse County, Wisconsin Reporting Period January - December 1985 (County, State) l Indicator Location with Highest Control Sample
. Type and Locations Annual Mean Locations Number of Type Number of Mean (F)C Mean (F)
Mean (F)
Non -ri,a u ne (Units)
Analyses 8 LLDb Range Locationd Range Range Resultse Eggs Nb-95 0.016
<L L D None 0
(pCl/a wet)
(con t 'd) 2r-95 0.020 (L LD 9
None O
Ru-103 0.012 (LLD None 0
Ru-106 0.091 (LLD None O
Cs-134 0.011 (LLD None O
Ce-141 0.020
<LLD None O
Ce-144 0.061 (LLD None 0
vegetables GA 7
0.02 0.25 (1/1)
K-17, Jansky Farm 0.25 (1/1) 0.15 (6/6) 0 (pCi/o wet) 4.25 mi W (0.07-0.23)
GB 7
0.1 3.95 (1/1)
K-17 Jansk y Farm 3.95 (1/1) 2.51 (6/6) 0 4.25 mi W (1.47-3.85)
Sr-89 7
0.004 (LLD
<LLO O
Sr-90 7
0.001 0.004 (1/1)
K-26. Bertler's Fruit 0.005 (5/5) 0.005 (5/5) 0 Stand,10.7 mi SSW 0.003-0.008 0.003-0.008 G5 7
Be-7 0.035
<LLD (LLO 0
K-40 0.75 2.02 (1/1)
K-17. Jansky Farm 2.02 (1/1) 1.91 (6/6) 0 4.25 mi W (0.39-3.34)
Nb-95 0.0040 (LL D (LLO O
Ru-103 0.0043
<Lt0 (LLD 0
Ru-106 0.033
Cs-137 0.0042
<LLD
<LLD 0
Ce-141 0.0077 (LLD
<LLD 0
Ce-144 0.0012 (LLD
<tLD 0
i
\\j C
V Table 4.5 Environmental Radiological Monitoring Program Sumary (continued)
Name of Faci 1ity Kewaunee NucIear Power P1 ant Dochet No.
50-305 Location of Facility Aewaunee County, Wisconsin Reporting Period January - Decen.ber.1985 (County, State)
Indicator location with Highest Control Samole Type and Locations Annual Mean Locations Nunt)er of Type Number of Mean (F)C Mean (F)
Mean (F)
Non-routine (Units)
Analyses 8 LLDb Range Location 0 Range Range Resultse Grain - Oats GA 2
0.18
<LLO None 0
(pct /g wet)
GB 2
0.1 3.62 (2/2)
K-23 Kewaunee Site 3.62 (2/2)
None.
0 (1.90-5.33) 0.5 mi W (1.90-5.33)
Sr-89 2
0.005 (LLD None 0
Sr-90 2
0.002 0.017 (1/2)
K-23 Kewaunee Site 0.01/ (1/2)
None 0
(0.004-0.030) 0.5 mi W (0.004-0.030)
G5 2
0.026 None 0
Be-7 0.2 1.20 (2/2) 1.20 (2/2)
None 0
- (0.49-1.90)
(0.49-1.90) l K-40 0.1 I 3.79 (2/2)
K-23, Kewaunee Site 3.79 (2/2)
~
None 0
(3.46-4.12) 0.5 ni W (3.46-4.12)
Nb -95 0.010 (LLD None 0
Zr-95 0.015 (LLD None O
Ra-103 0.007 (LLD None O
Ru-106 0.061
<LLD None O
Cs-137 0.007
<tLD None O
Ce-141 0.013
<tLO None O
Ce-144 0.044
<LLD None 0
l Catt lef eed GA 6
- 0. 2 (oC /g wet) l0.9(4/4)
K-4, Stangel Farm 2.0 (1/1) 0.9 (2/2) 0 i
(0.3 2.0) 3.0 mi N (0.4-1.4) i GB 6
0.1 7.6 (4/4)
K-4, Stangel Farn 16.5 (1/1) 8.0 (2/2) 0 l(2.6-16.5) 3.0 mi N (4.3-11.7)
Sr-89 6
0.14
<LLD (LLO O
Sr-90 6
0.039 0.075 (1/4)
K-6, Novitsay Farm 0.235 (1/1) 0.141 (2/2) 0 6.7 mi WSW (0.047-0.235; GS 6
Be-7 0.15 0.79 (4/4)
K-4, Stangel Farm 1.96 (1/1) 0.40 (2/2) 0 (0.29-1.96) 3.0 mi N (0.36-0.43)
K-40 1.0 7.46 (4/4)
K-4, Stangel Farm 17.2 (1/1) 7.57 (2/2) 0 l
(3.30-17.20) 3.0 mi N (4.17-10.97) 1 3
l
n A
/"N N
/
Table 4.5 Environmental Radiological Monitoring Program Suninary (continued)
Name of Facility Kewaunee Nuclear Power Plant Docket No.
50-305 Location of Facility Kewaunee County, Wisconsin Reporting Period January - December 1985 (County, 5 tate )
Indicator Location with Highest Control Sample TWe and Locations Annual Mean Locations Nuntaer of Type Number of Mean (F)C Mean (F)
Mean (F)
Non-routine (Units)
Analyses 8 LLDb Range Locationd Range Range Resultse Cattlefeed Nb-95 0.015 (LLD
<LLD 0
d Ir-95 0.024
<L L D
<LLD 0
Ru-103 0.013
<tLD (LLD 0
Ru-106 0.12
<tLD
<LLD 0
Cs-134 0.012
<tLD (LLD 0
Cs-137 0.013
<tLD
<LLD 0
4.
Ce-144 0.090
<tLD
<LLD 0
Grass GA 24 0.11 0.29 (17/18)
K-12, LeCaptain Farm
- 0. ad (3/ 3) 0.27 (6/6) 0 ro (pCi/g wet)
(0.11-0.50) 1.5 mi WSW (0.30-0.42)
(0.15-0.45) cn GB 24 0.1 6.09 (18/18)
K-12, LeCaptain Farm
- 7. 7E (3/3) 5.30 (6/6) 0 (2.94 9.68) 1.5 mi W5W (6.70-8.95)
(3.02-8.55)
Sr-89 24 0.020
<L L D (LLb 0
Sr-90 24 0.006 0.028 (18/18)
K-lb, Middle Creek 0.043 (3/3) 0.025 (6/6) 0 (0.010-0.057)
On site, 0.12 mi N (0.034-0.057)
(0.007-0.069)
G5 24 Be-7 0.3 2.20 (18/18)
K-lb, Middle Creek 2.93 (3/3) 1.50 (6/6) 0 (0.57-5.73)
On site, 0.12 mi N (0.60-5.73)
(0.43-2.75)
K-40 0.1 6.88 (18/18)
K-5, Paplham Farm 7.74 (3/3) 7.54 (6/6) 0 (4.50-8.96) 3.5 mi NNW (6.85-8.20)
(6.32-8.28)
Mb-95 0.1
<tLD (LLD 0
Zr-95 0.1 (LLD
<LLD 0
Ru-103 0.1
<tLD
<tLD 0
0.1
<tLD
<LLD 0
Ru-106 j
K
\\
i
%)
Q,l Table 4.5 Environmental Radiological Monitoring Program Sumary (continued)
Name of Facility Kewaunee Nuclear Power Plant Docket No.
50-305 Location of Facility Kewaunee County, Wisconsin Reporting Period January - December 1985 (County, 5 tate)
Indicator Locat' ion with Highest Control Sample Type and Loc ations Annual Mean Locations hutter of Type Nunter of Mean (F)c Mean (F )
Mean (F)
Non-routine (Units )
Analyses 4 tiDb Range Locationd Range Range Resultse Grass Cs-137 0.02
<tLO K-6, Novitsky Farm 0.066 (1/3) 0.0t>6 (1/6) 0 (pCi/g dry)
- 6. 7 mi WSW (cont'd)
Ce-141 0.1 (LLD
<tLD 0
Ce-144 0.1
<tLD (LLD 0
Soil
' GA 14
- 2. 7 6.3 (10/10)
K-5, Paolham farm 7.8 (2/2) 7.0 (3/4) 0 (pCi/q dry)
( 3. 5-9. 6 )
3.5 mi WSW (6.0-9.6)
(5.7-9.0)
GB 14
- 2. 0 23.2 (10/10)
K-5, Paplham Farm 25.6 (2/2) 22.7 (4/4) 0 (17.1-29.5) 3.5 mi WSW (24.6-26.7)
(20.2-25.9)
,' Sr-89 14 0.11
<LLD
<LLD 0
m i
N Sr-90 14 0.01 0.09 (9/10)
K-6, Novitsky Farm 0.55 (2/21 0.33 (4/4) 0 (0.01-0.17)
- 6. 7 mi WSW (0.22-0.89)
(0.06-0.89)
! G5 14 Be-7 0.41
<t L D
<tLD 0
K-40 1.4 16.7 (10/10)
K-5, Paolham farm
- 19. 3 (2/2) 17.4 (4/4)
J (11.4-20.7) 3.5 mi hNW (17.8-20.7)
(15.1-19.6)
Nb-95 0.15
<LtD
<LLD 0
f Zr-95 0.07 (LLD
<LLO O
i j
Ru-103 0.09 (LLD
<tLD 0
i Ru-106 0.11 (LLD
<tLD 0
t Cs-137 0.01 0.29 (8/10)
K-6, Novitsk y Farm 0.59 (2/2) 0.46 (4/4) 0 (0.05-9.49) 6.7 mi WSW (0.39-0.78)
(0.27-0.78) 3 Ce-141 0.07 (LLD (LLD 0
Ce-144 0.08
<t L D
<LLD 0
0) f t
o v)
Table 4.5 Environmental Radiological Monitoring Program Samary (continued)
Name of Facility Kewaunae Nuclear Power Plant Docket No.
50-305 Location of Facility newaunee County, Wisconsin Reporting Perica January - December 1985 (County, State)
Indicatar Location with Highest Control Sanple Type and Locations Annual Mean Locations Number of Type Number of Mean (F)c Mean (F)
Mean (F)
Non-routine (Units)
Analyses 8 LLDb Range Locationd Range Range Resultse Surface Water GA(55) 72
- 1. 0 (LLD
<LLD 0
(pCi/l)
GA(DS) 72
- 2. 7f 4.5 (2/60)
K-la, North Creek, 5.7 (1/12)
<LLD 0
(3.3-5.7)
Disite, 0.62 mi N GA(TR) 72
- 3. 7f 5.7 (1/60)
K-la, North Creek, 5.7 (1/12)
(LLO O
Onsite, 0.62 mi N GB(55) 72
- 0. 5 0.6 (4/60)
K-14 Two Creeks 0.8 (1/12)
<LLD 0
(0.5-0.6)
Park, 2.5 et 5 GB(DS) 72 0.99 5.6 (58/60)
K-la, North Creek, 11.4 (12/12) 2.5 (12/12) 0 (0.9-31.2)
Onsi te, 0.62 mi N (4.0-31.2)
(1.6-3.6)
GB(TR) 72 1.39 5.8 (57/60)
K-la, horth Creek, 11.4 (12/12) 2.5 (12/12) 0 (1.4-31.2)
Dist te, 0.62 mi N (4.0-31.2)
(1.6-3.6) to H-3 36 220 500 (9/24)
K-Id Condenser 760 (4/12)
<L L D 0
CD (230-2140)
Discharge Onsite (230-2140) 0.10 mi E Sr-89 12
- 1. 8
<tLD (LLD 0
Sr-90 12 1.1 1.2 (1/8)
K-Id, Condenser 1.2 (1/4)
<LLD 0
Discharge Onsite 0.10 mi E K-40 72 0.5 4.4 (60/60)
K-la, North Creek 8.7 (12/12) 1.4 (12/12) 0 (f lame)
(1.0-29.3)
Onsite, 0.62 mi N (4.1-29.3)
(0.6-2.2)
Fisn-Muscle GA 5
0.05 0.11 (5/5)
K-Id Condenser 0.11 (5/5)
None 0
(pCi/q wetl (0.09-0.13)
Discharge Onsite (0.09-0.13) 0.10 mi E GB 5
- 1. 0 2.61 (5/5)
K-Id, Condenser 2.61 (5/5)
None 0
(2.42-2.85)
Discharge, Onsite (2.42-2.85) 0.10 mi E l
1 O
O O
Table 4.5 Environmental Radiological Monitoring Procram Sumary (continued) kame of Facility Kewaunee Nuclear Power Plant Docket No.
50-305 Location of Facility kewaunee County, Wisconsin Reporting Period January - Dece=t:er.1985 (County, State)
Indicator Locatinn with Highest Control Sanp le Type and Locations Annual Mean Locations Nurrcer of Type Number of Mean (F)c Mean (F)
Mean (F)
Non -rout ine (Units)
Analysesa LL0b Range Locationd Range Range Results' Fish-Muscle G5 5
i i
td Be - 7 0.41 (LLO None 0
K-40 1.05 2.73 (5/5)
K-Id, Condenser 2.73 (5/5) hone 0
(2.22-3.04)
Discharge, Onsite (2.22-3.04) 0.10 mi E Nb-95 0.061 (LLO None 0
2r-95 0.045 (LLO None 0
Ru-103 0.045
<LL O None O
Ra-106 0.12 (LLO None O
Cs-137 0.012 0.14 (4/5)
K-lo, Condenser Dis-0.14 (4/5)
None 0
(0.09-0.18) charge, Onsite (0.09-0.18) 0.10 mi E Ce-141 0.081 (LLO None O
Ce-144 0.084
'LLO None 0
Fish-Bones GA 5
0.92
<tLD None 0
GB 5
- 0. 5 1.15 (5/5)
K-Id, Condenser Dis-1.15 (5/5)
None 0
(0.86-1.44) charge, Onsite (0.86-1.44) 0.10 mi C Sr-89 5
0.10 0.35 (1/5)
K-Id, Condenser Dis-0.35 (1/5)
None O
charge, Onsite 0.10 mi E Sr-90 5
0.10 0.18 (5/5)
K-Id, Condenser Dis-0.18 (5/5)
Noce 0
(0.11-0.26) charge, Onsite (0.11-0.26) 0.10 mi E Perip hyton GA 12 0.66 0.97 (2/10)
K-le, South Creek 1.20 (1/2) 1.18 (1/2) 0 (slime)
( 0. 74-1.20)
Onsite, 0.12 mi 5 GB 12 0.1 1.69 (10/10)
K-la, North Creek 3.14 (2/2) 2.87 (2/2) 0 l
(0.20-3.51)
Oiri te, 0.62 mi N (2.76-3.51)
(0.64-5.10)
Sr-89 12 0.046
<t L D
<LLO O
Sr-90 12 0.005 0.044 (9/10)
K-la, North Creek 0.097 (2/2) 0.046 (2/2) 0 (0.006-0.186)
Onsite, 0.62 mi N (0.008-0.186)
(0.020-0.071) l l
l
l f
f~ '%.
(
v]
\\
(
\\
Table 4.5 Environmental Radiological Monitoring Program Summary (continued)
Name of f acility llewaung Actear Power Plant -
Dodet No.
50-305 Location of Facility kewaunee County, % rontsj Reporting Period January - Decemt,er 1985 (County, State) 1 Indi;.!a~
Location with Highest Control Sarrele Type and Locctions Annual Mean Locations Anter of Type Number of nean (O C Mean (F)
Mean (F)
Non-routine (Units)
Analyses 8 LLLA Range locationd Range Range Resultse Periphyton GS 12 (Slime)
(pCl/g wet)
Be -7 0.84 1.02 (1/10)
K-9, Rottok Water 1.11 (1/2) 1.11 (1/2) 0 (cont'd)
Intake,11.5 mi NNE K-40 0.50 1.85 (10/10)
K-la, North Creek, 3.95 (2/2) 1.83 (2/2) 0 (0.63-5.37)
(hsite, 0.62 mi N (2.52-5.37)
(1.38-2.21)
<tLD 0
Co-58 0.051 0.078 (1/10)
K-14. Two Creeks 0.078 (1/2)
<LLD 4
Park, 2.5 mi 5 Co-60 0.031 0.096 (3/10)
K-14, Two Creeks 0.130 (1/2)
<LLD 0
(0.072-0.130)
Park, 2.5 mi 5 ND-95 0.048
<LLD (LLD 0
Zr-95 0.077
<tLD
<tLD 0
Ra-l')3 0.058
<t L D
<LLD 0
Ru-106 0.16
<t LD (LLD 0
Cs-134 0.030 (LL D
<LLD 0
Cs-137 0.025
<LLD
<LLD 0
Ce-141 0.13
<LLD
<LLD 0
Ce-144 0.16
<LLD
<tLD 0
Bottom GA 20
- 3. 4 (LLD
<LLD 0
Sediments (pCi/q ory)
GR 20
- 1. 0 6.1 (16/16)
K-!c, Condenser 6.6 (4/4) 6.0 (4/4) 0 (2.3-10.4)
Discharge, Onsite (5.1-8.4)
(5.3-7.7) 0.10 mi 4 Sr-89 20 0.032
<t te
<tLD 0
Sr-90 20 0.015
<t L D
<tLD 0
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5.0 REFERENCES
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R.
and H.
A.
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1955.
Beryllium-7 produced by cosmic rays.
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Environmental Radioactivity, McGraw-Hill, New York, New York, pp. 213, 275, and 276.'
Gold, S., H. W. Barkhau, B. Shlein, and B. Kahn,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 1978.
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Annual Report.
Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report Part II, Data Tabulations and Analysis, January - December 1979.
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Annual Report.
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Annual Report.
Radiological Monitoring Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report Part II, Data Tabulations and Analysis, January - December 1981.
1983.
Annual Report.
Radiological Monitoring Progran for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report Part II, Data Tabulations and Analysis, January - December 1982.
Industrial BIO-TEST Laboratories, Inc.1974.
Annual Report. Pre-operational Radiological Monitoring Program for the 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 OL'!
Program for the Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, January - December 1976.
32
.-==-
1 l
1978.
Annual Report.
Radiological Monitoring Program for the i
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 i
Radiation Background in the United States.
NCRP Report No. 45.
i Solon, L. R., W. M. Lowder, A. Shambron, and H. Blatz.
1960.
Investigations of Natural Environmental Radiation.
Science. 131: 903-906.
j Teledyne Isotopes Midwest Laboratory.
1984.
Annual Report.
Radiological i
Monitoring Program for the ' Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II, Data Tabulations and Analysis, January - December 1983.
I 1985.
Annual Report.
Radiological Monitoring Program for the.
1 Kewaunee Nuclear Power Plant, Kewaunee, Wisconsin, Final Report, Part II,-
j Data Tabulations and Analysis, January - December 1984, j
1986.
Annual Report.. Radiological Monitoring Program'for the Kewaunee Nuclear Fower Plant, Kewaunee, Wisconsin, Final Report, Part II, Data Tabulations and Analysis, January - December 1985.
b l
Wilson, D. W., G. M. Ward, and J. E. Johnson,1969.
In Environmental Contamina-j tion by Radioactive Materials, International Atomic Energy Agency, p.
125.
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Appendix A 1
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Interlaboratory Comparison Program Results L
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- O Appendix A 4
'I Interlaboratory Comparison program Results 7
i l
Teledyne Isotopes Midwest Laboratory (formerly Hazleton Environmental Sciences) has participated in interlaboratory comparison (crosscheck) programs since the 1
formulation of its quality control program in December 1971.
These programs are operated by agencies wnich supply environmental-type samples (e.g., milk or water) containing concentrations of radionuclides known to the issuing agency j
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 f
alert it to any possible problems, j
Participant laboratories measure the concentrations of specified radionuclides and report them to the issuing agenc/.
Several months later, the agen;;y reports the known values to the participent laboratories and specifies control limits.
Results consistently higher or Icwer than the known values or outside r
the control limits indicate a need to check the instruments,or procedures j
used.
i
)
The results in Table A-1 were obtained through participation in the environ-mental sample crosscheck program for milk, water, air filters, and food samples during the period 1982 through October 1985.
This program has been 4
conducted by the U.
S.
Environmental Protectio 1 Agency Intercomparison and Calibration Section, Quality Assurance Branch, Environmental Monitoring and Support Laboratory, Las Vegas, Nevada.
1 The results in Table A-2 were obtained for thermoluminescent dosimeters
(
(TLD's) during the period 1976, 1977, 1979, 1980, and 1981 through parti-cipation in the Second, Third, Fourth, and Fif th International Intercomparison of Environmental Dosimeters under the sponsorships listed in Table A-2.
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,)
A-2 i
--.,-,--_-.,___-,-,.,-.,.-.,__.-,n-,
, n -,- - _
.-n
.-._.--._.n,.,---,,,-_---
(3 Table A-1.
U.S. Environmental Protection Agency's crosscheck prograrn, com-parison of EPA and Teledyne Isotopes Midwest Laboratory results for milk, water, air filters, and food samples,1982 through 1985.a Concentration in pCi/lb Lab Sanple Date TIhL Result EPA Result Code Type Collected Analysis t2cc t3o, n=1d STW-270 Water Jan. 1982 Sr-89 24.312.0 21.0 5.0 Sr-90 9.4 0.5 12,011.5 STW-273 Water Jan. 1982 1-131 8.6t0.6 8.4tl.5 STW-275 Water Feb. 1982 H-3 1580t147 1820t342 STW-276 Water Feb. 1982 Cr-51
<61 0
Co-60 26.0 3.7 20t5 Zn-65
<13 15t5 Ru-106
<46 20t5 Cs-134 26.8t0.7 22t5 Cs-137 29.7tl.4 23t5 STW-277 Water Mar. 1982 Ra-226 11.9!1.9 11.611.7 g
STW-278 Water Mar. 1982 Gross alpha 15.621.9 1915 Gross beta 19.2 0.4 19k5 STW-280 Water
/cr. 1992 H-3 2690280 28601360 STW-281 Water Apr. 1982 Gross alpha 75t7.9 85t21 Gross beta 114.115.9 106 5.3 Sr-89 17.4tl.8 24t5 Sr-90 10.5t0.6 1211.5 Ra-226 11.4t2.0 10.9tl.5 Co-60
<4.6 0
STW-284 Water May 1982 Gross alpha 31.516.5 27.5t7 Gross beta 25.913.4 2915 STW-285 Water June 1982 H-3 1970 1408 1830t340 STW-286 Water June 1982 Ra-226 12.621.5 13.4t3.5 Ra-228 11.122.5 8.712.3 STW-287 Water June 1982 1-131 6.510.3 4.4 0.7 STW-290 Water Aug. 1982 H-3 3210 140 2890t619 O
A-3
t Table A-1.
(continued)
Concentration in pCl/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis i2ac 130, n=1d STW-291 Water Aug. 1982
' I-131 94.6i2.5 87t15 STW-292 Water
. Sept [ 1982 Sr-89 22.713.8 24.St8.7 l,
Sr-90 10.910.3 14.512.6
~
STW-296 Water Oct. 1982 Co-60 20.0il.0 20i8.7 4
Zn-65 32.315.1 2418.7 Cs-134 15.3tl.5 19.0*8.7 l
Cs-137 21.011.7 20.0i8.7 STW-297 Water Oct. 1982 H-3 2470t20 2560i612 STW-298 Water Oct. 1982 Gross alpha 32i 30 55124 Gross beta 81.716.1 8118.7 Sr-89
<2 0
Sr-90 14.110.9 17.212.6 Cs-134
<2 1.8*8.7 i
Cs-137 22.710.6 20i8.7 Ra-226 13.6t0.3 12.513.2 Ra-228 3.9tl.0 3.6i0.9 i
STW-301 Water Nov, 1982 Gross alpha 12.0tl.0 19.018.7 Gross beta 34.0*2.7 24.0i8.7 STW-302 Water
, Dec.1982 1-131 40.0i0.0 37.0110 STW-303 Water Dec. 1982 H-3 1940120 19901345 STW-304 Water Dec. 1982 Ra-226 11.7i0.6 11.0t1.7 Ra-228
<3 0
STW-306 Water Mn. 1983 Sr-89 20.018.7 29.215 Sr-90 21.7t8.4 17.2 1.5 STW-307 Water Jan.1983 Gross alpha 29.014.09 29.0i13 l
Gross b,e,tc '
29.310.6 31.018.7 STM-309 Milk Feb. 1983 Sr-89 35i2.0 3718.7 Sr-90 13.710.6 18t2.6 I-131 55.713.2 55i10.4 1
Cs-137 29tl.0 2618.7 Ba-140
<27 0
K-40 1637tS.8 1512i131 A-4 3
4 2
Table A-1.
(continued)
Concentration in acl/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis 12ac 33o, n=1d STW-310 Water Feb. 1983 H-3 2470180 2560i612 STW-311 Water March 1983 Ra-226 11.9il.3 12.713.3 Ra-228
<2.7 0
STW-312 Water March 1983 Gross alpha 31.6t4.59 31113.4 Gross beta 27.012.0 2818.7 STW-313 Water April 1983 H-3 3240180 3330i627 STW-316 Water May 1983 Gross alpha 9417 64fl9.9 Gross beta 13315 149t12.4 Sr-89 1911 24f8.7 Sr-90 1211 1312.6 Ra-226 7.9 0.4 8.5t2.25 i
Co-60 3012 30f8.7 l
i Cs-134 27i2 33t8.7
~
Cs-137 2911 2718.7 STW-317 Water May 1983 Sr-89 59.712.1 57*8.7 Sr-90 33.711.5 3813.3 STW-318f Water May 1983 Gross alpha 12.8tl.5 1118.7 Gross beta 49.4t3.9 5718.7 I
STM-320 Milk
&ne 1983 Sr-89 2010 2518.7 Sr-90 10i1 1612.6 I-131 30 1 30110.4 Cs-137 52 2 4718.7 K-40 1553157 14861129 STW-321 Water
&ne 1983 H-3 1470 89 15291583 STW-322 Water
&ne 1983 Ra-226 4.310.2 4.811.24 Ra-228
<2.5 0
i STW-323 Water 11y 1983 Gross alpha 311 718.7 Gross beta 2110 2218.7 i
STW-324 Water August 1983 1-131 13.310.6 14110.4
- O A-5
I
()
Table A-1.
(continued)
Concentration in pCi/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis 12cc 13o, n=1d STAF-326 Air August 1983 Gross beta 42 2 3618.7 Filter Sr-90 1412 10i2.6 Cs-137 19 1 1518.7 STW-328 Water Sept. 1983 Gross alpha 2.310.6 Si8.7 Gross beta 10.7tl.2 918.7 SIW-329 Water Sept. 1983 Ra-226 3.010.2
-3.110.81 Ra-228
- 3. 2i0. 7 2.0i0.52 STW-331 Water Oct. 1983 H-3 1300i30 12101570 STW-335 Water Dec. 1983 I-131 19.6fl.9 20t10.4 STW-336 Water Dec. 1983 H-3 28701100 23891608 STAF-337 Air Nov. 1983 Gross alpha 18.010.2 1918.7 Filter Gross beta
- 58. 6tl. 2 50t8.7 f)
Sr-90 10.9i0.1 15i2.6 Cs-137 30.li2.5 20i8.7 STW-339 Water Jan. 1984 Sr-89 47.211.9 36t8.7 3
Sr-90 22.Si4.0 2412.6 STW-343 Water Feb. 1984 H-3 2487i76 23831607 STM-347 Milk March 1984 I-131 5.3 1.1 6tl.6 STW-349 Water se ch 1984 Ra-226 4.010.2
.4.lil.06 Ra-228 3.610.3 2.010.52 STW-350 Water March 1984 Gross alpha 3.811.1 518.7 Gross beta 24.2 2.0 20t8.7 STW-354 Water April 1984 H-3 3560 50 35082630 STW-355 Water April 1984 Gross alpha 21.0i4.1 35t15.2 Gross beta 127.8i4.1 147t12.7 Sr-89 29.3i2.0 23t8.7 Sr-90 16.6t0.7 2612.6 Ra-226 4.0 1.0 4.0tl.04 Co-60 32.3 1.4 3018.7 Cs-134 33.6i3.1 3018.7
()
Cs-137 33.312.2 26i8.?
v A-6
Table A-1.
(continued)
Concentration in pCi/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis t2cc i3o, n=1d STW-358 Water May 1984 Gross alpha 3.0 0.6 318.7 Gross beta 6.7tl.2 618.7 I
STM-366 Milk June 1984 Sr-89 21 3.1 25i8.7 Sr-90 13i2.0 1712.6 I-131 46t5.3 43i10.4 Cs-137 3814.0 35 8.7 K-40 1577t172 1496t130 STW-368 Water July 1984 Gross alpha 5.111.1 618.7 Gross beta 11.9i2.4 13i8.7 STW-369 Water August 1984 I-131 34.315.0 34.0i10.4 STW-370 Water August 1984 H-3 30031253 2817i617 STF-371 Food July 1984 Sr-89 22.0i5.3 25.018.7 Sr-90 14.7t3.1 20.0i2.6 f'y'-
<172 39.0t10.4 Cs-137
- 24. Oi5. 3 25.018,7 i
K-40 25031132 2605t226.0 STAF-372 Air August 1984 Gross alpha 15.3tl.2 17i8.7 Filter Gross beta 56.010.0 5118.7 Sr-90 14.3tl.2 18i2.4 1
Cs-137 21.0 2.0 1518.7 STW-375 Water Sept. 1984 Ra-226 5.li0.4 4.9il.27 Ra-228 2.210.1 2.310.60 STW-377 Water Sept. 1984 Gross alpha 3.3 1.2 5.0i8.7 Gross beta 12.712.3 16.018.7 STW-379 Water Oct. 1984 H-3 2860 312 2810 356 STW-380 Water Oct. 1984 Cr-51
<36 4018.7 Co-60 20.311.2 20 8.7 Zn-65 150i8.1 147 8.7 Ru-106
<30.
4718.7 Cs-134 31.317.0 31i8.7 Cs-137 26.7tl.2 24t8.7 O
A-7
~
. Table A-1.
(continued)
[}
Concentration in pCi/lb Lab Sample Date TIML Result EPA Result Code Type Collected-Analysi s 12 c i3, n=1d STM-382 Milk Oct. 1984 Sr-89 15.7i4.2 2218.7 Sr-90 12.7tl.2 16i2.6 I-131 41.7 3.1 42t10.4 Cs-137 31.316.1 32i8.7 K-40 1447166 1517i131 STW-384 Water Oct. 1984 Gross alpha 9.7tl.2 14t8.7 (Blind)
Sanple A Ra-226 3.310.2 3.0t0.8 Ra-228 3.4tl.6 2.1 0.5 Uranium nae Si10.4 Sample 8 Gross beta 48.315.0 64i8.7 Sr-89 10.7i4.6 11f8.7 Sr-90 7.3tl.2 12i2.6 Co-60 16.3tl.2 14i8. 7 Cs-134
<2 218.7 Cs-137 16.7tl.2 14i8.7 4
STAF-387 Air Nov. 1984 Gross alpha 18.7tl.2 15i8. 7 Filter Gross beta 59.0t5.3 52 8.7 Sr-90 18.3tl.2 21i2.6 Cs-137 10.311.2 1018.7 STW-388 Water Dec. 1984 I-131 28.0*2.0 36t10. 4 STW-389 Watar Dec. 1984 H-3 35831110 31821624 STW-391 Water Dec. 1984 Ra-226 8.4tl.7 8.6t2.2 Ra-228 3.li0.2 4.1 1.1 STW-392 Water Jan. 1985 Sr-89
<3.0 3.0 8.7 Sr-90 27.3 5.2 30.012.6 STW-393 Water Jan. 1985 Gross alpha 3.3 1.2 518.7 Gross beta 17.3 3.0 15t8.7 STS-395 Food Jaa. 1985 Sr-89 25.316.4
.34.0i5.0 Sr-90 27.0 8.8 26.0il.5 I-131 38.0i2.0 35.016.0 Cs-137 32.7 2.4 29.0 5.0 K-40 1410t212 13821120 O
A-8
O Table A-1.
(continued)
Concentration in pCi/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis i2cc 130, n=1d STW-397 Water Feb. 1985 Cr-51
<29 48i8.7 Co-60 21.3t3.0 20i8.7 Zn-65 53.715.0 5518.7 Ru-106
<23 2518.7 Cs-134 32.311.2 3518.7 Cs-137 25.3t3.0 2518.7 STW-398 Water Feb. 1985 H-3 38691319 3796t634 STM-400 Milk March 1985 I-131 7.312.4
- 9. 0f 0. 9 STW-402 Water March 1985 Ra-226 4.620.6 5.0il.3 Ra-228
<0. 8 9.0i2.3 Reanalysis Ra-228
- 9. 0i0. 4 STW-404 Water March 1985 Gross alpha 4.7i2.3 6t8.7 Gross beta 11.311.2 1518.7 Q
STAF-405 Air March 1985 Gross alpha 9.3 1.0 10.018.7 Filter Gross beta 42.0tl.1 36.018.7 Sr-90 13.3tl.0 15.012.6 Cs-137
- 6. 3tl. 0 6.0i8.7 STW-407 Water April 1985 I-131 8.0i0.0 7.5fl.3 STW-408 Water April 1985 H-3 3399i150 35591630 STW-409 Water April 1985 (Blind)
Gross alpha 29.711.8 32.0i5.0 Sample A Ra-226 4.4i0.2 4.110.6 Ra-228 nae 6.2i0.9 Uranium nae 7.016.0 Sample 8 Gross beta 74.3t11.8 72.0i5.0 Sr-89 12.317.6 10.015.0 Sr-90 14.7 2.4 15.0 1.5 Co-60 14.7t2.4 15.015.O Cs-134 12.0 2.0 15.015.0 Cs-137 14.0i2.0 12.0i5.0 O
A-9
O Table A-1.
(continued)
Concentration in pCi/1b Lab Sample Date TIML Result EPA Result Code Type Collected Analysis 12 c 13, n=1d STW-413 Water May 1985 Sr-89 36.0t12. 4 39.0i5. 0 Sr-90 14.314.2
- 15. 0tl. 5 STW-414 Water May 1985 Gross alpha 8.314.1
- 12. 0i5.0 e
Gross beta 8.7tl.2 11.015.0 1
STW-416 Water June 1985 Cr-51 44.7i6.0 44.0i5.0 Co-60 14.3tl.2 14.015.0 Zn-65 50.3t7.0 47.015.0 Ru-106 55.315.8 62.015.0 Cs-134 32.7tl.2
- 35. 0i5. 0 Cs-137 22.7i2.4 20.0i5.0 I
STW-418 Water June 1985 H-3 2446t132 2416t351 STM-421 Milk June 1985 Sr-89 10.314.6 11.018.7 Sr-90 9.012.0 11.012.6 O'
I-131 11.7tl.2 11.0i10.4 Cs-137 12.711.2 11.018.7 K-40 1512i62 1525i132 STW-423 Water July 1985 Gross alpha 5.0i0.0 11.0i8.7 Gross beta
- 5. 0i2. 0
- 8. 0i8. 7 STW-425 Water August 1985 I-131 25.7i3.0 33.0110.4 STW-426 Water August 1985 H-3 4363183 4480i776 STAF-427 Air August 1985 Gross alpha 11.310.6 13.0i8.7 Filter Gross beta 46.011.0 44.018.7 Sr-90 17.710.6 18.012.6 Cs-137 10.3i0.6
- 8. 0i8. 7 STW-429 Water Sept. 1985 Sr-89 15.7t0.6 20.0 8.7 Sr-90 7.010.0 7.012.6 STW-430 Water Sept. 1985 Ra-226 8.210.3 8.912.3 Ra-228 4.1 0.3 4.6tl.2 STW-431 Water Sept. 1985 Gross alpha 4.7 0.6
- 8. 0i8. 7 Gross beta 4.711.2 8.0i8. 7
()
A-10
O T.able A-1.
(continued)
Concentration in pCi/lb Lab Sample Date TIML Result EPA Result Code Type Collected Analysis 12cc i3c, n=1d STW-433 Water Oct. 1985 Cr-51
<13 21.018.7 Co-60 19.30.6 20.0i8.7 Zn-65 19.7t0.6 19.018.7 Ru-106
<19 20.0i8.7 Cs-134 17.0tl.0 20.018.7 Cs-137 19.311.2 20.0t8.7 STW-435 Water Oct. 1985 H-3 1957150 19741598 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 Branch,. Environmental Monitoring and Support Laboratory, U.S.
Environmental Protection Agency, b (EPA), Las Vegas, Nevada,All results are in pCi/1, except for elemental potassium (
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 12 standard deviations for three determinations.
d USEPA results are presented as the known values i control limits of 3a for n = 1.
e NA = Not analyzed.
f Analyzed but not reported to the EPA.
9 Results af ter calculations corrected (error in calculations when reported to EPA).
a J
A-11
O O
O Table A-2.
Crosscheck program results, thermoluminescent dosimeters (TLDs).
mR d
Teledyne Average 12o Lab TLD Result Known (all Code Type Measurement 12aa Value participants) 2nd International Interconiparisonb CaF :Mn Gama-Field 17.011.9 17.1c 16.417.7 115-2b 2
Bulb Gamma-Lab 20.814.1 21.3c 18.817.6 3rd International Intercomparisone CaF :Mn Gama-Field 30.713.2 34.914.8f 31.513.0 115-3e 2
?
Bulb C
Gama-Lab 89.616.4 91.7114.6f 86.2124.0 4th International Intercomparison9 CaF :Mn Gama-Field 14.111.1 14.lil.4f 16.019.0 115-49 2
Bulb Gama-Lab (Low) 9.311.3 12.212.4f 12.017.6 Gama-Lab (High) 40.411.4 45.819.2f 43.9113.2 5th International Intercomparisonh CaF :Mn Gama-Field 31.411.8 30.016.01 30.2114.6 115-5Ah 2
Bulb Gama-Lab 77.415.8 75.217.61 75.8140.4 at beginning Gama-Lab 96.6i5.8 88.418.81 90.7131.2 at the end
O
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Table A-2.
(Continued) mR d
Teledyne Average i 20 Lab TLD Result Known (all Code Type Measurement 12aa value participants) ll5-5Bh LiF-100 Gamma-Field 30.314.8 30.0161 30.2114.6 Chips Gamma-Lab 81.117.4 75.217.61 75.8140.4 at beginning i
Gamma-Lab 85.4111.7 88.418.8I 90.7131.2 at the end J,
a Lab result given is the mean 12 standard deviations of three determinations.
um 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.
c Value determined by sponsor of the intercomparison using continuously operated pressurized ion chamber.
.i d Mean *2 standard deviations of results obtained by all laboratories participating in the program.
e Third International Intercomparison of Environmental Dosimeters conducted in summer of.1977 by Oak Ridge National Laboratory and the School of Public Health of the University of Texas, Houston, Texas.
f Value 12 standard deviations as determined by sponsor of the intercomparison using continuously operated pressurized ion chamber.
4 9 Fourth International Intercomparison of Environmental Dosimeters conducted in summer of 1979 by the School of Public Health of the University of Texas Houston, Texas.
h Fifth International Intercomparison of Environmental Dosimeter conducted in fall of 1980 at Idaho Falls, Idaho and sponsored by the School of Public Health of the University of Texas, Houston, Texas and Environmental Measurements Laboratory, New York, New York, U.S. Department of Energy.
i Value determined by sponsor of the intercomparison usinn continuously operated pressurized ion chamber.
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1 Appendix B Data Reporting Conventions t
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B-1
-v
,-a,
---ne-,-
-,,------vwcen--
a a,
-.,n--
-,-,,,n--~n.v.-,-e.,n.w.--aam--.e---,
- -- w a m w. -m me m swnn, m-w wm
- vee,
f' Data Reporting Conventions 1.0.
All activities are decay corrected to collection time.
2.0.
Single Measurements Ehch single measurement is reported as follows:
xis where x = value of the measurement; s = 2a counting uncert'ainty (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 detecticn based on 3a uncertainty for a background sample.
3.0.
Duolicate Analyses 3.1.
Individual results: x1 i s1 x2 i s2 Reported result:
xis where x = (1/2) (x1 + x2) 2+s2 s=
s 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; 4
<L otherwise B-2
(
O 4.0.
Computation of Averages and Standard Deviations 1
4.1 Averages and standard deviations 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 4
of quarterly standard deviations.
The average x and standard deviation (s) of a set of n numbers x1, x2, Xn are defined as follows:
x=fEx Z I*'*I2 i
s=
n-1 4.2 Values below the highest lower limit of detection are not included in the average.
1 4.3 If all of the values in the averaging group are less than the highest LLD, the highest LLD is reported.
O.
4.4 If all but one of the values are less thn the highest LLD, the 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.
i 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 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.
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Appendix C Maximum' Permissible Concentrations
[
i of Radioactivity in Air and Water.
j Above Background in Unrestricted Areas i
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Q.
.i'
.C-1 1
i
> - -..... - -.. - -. -.. -.., = -.., -, _
.-,n,,,---,,_,a
O Table C-1.
Ma~Num permissible concentrations of radioactivity in air and water above 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 pCi/1 Barium-140 20,000 pCi/l Iodine-131 300 pCi/1 Potassium-40c 3,000 pCi/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 f actor of 700 to reduce the dose resulting from the air-grass-cow-milk-child pathway.
c A natural radionuclide.
1 O
C-2
NRC-86-24 WISCONSIN PUBLIC SERVICE CORPORATIO5 February 28, 1986 Mr. Richard C. DeYoung, Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Changes Made to Facility Per 10 CFR 50.59 Enclosed is a copy of the 1985 Kewaunee Nuclear Power Plant (KNPP) Annual Operating Report. This report is being sent to you in accordance with 10 CFR 50.59(b). Section 3 of the report describes those facility changes allowed by 10 CFR 59.59(a)(1).
The 1985 KNPP Annual Operating Report satisfies the reporting requirements of KNPP Technical Specification 6.9.1.b (annual reporting requirements),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).
Very truly yours, l
D. C. Hintz Manager - Nuclear Power GWH/jms Enc.
cc - Mr. Robert Nelson, US NRC -w/o attach.
Mr. George Lear, US NRC - w/o attach.
600 North Adams = P.O. Box 19002
- Green Bay, WI 54307-9002
' t