ML20030A489

Semiannual Operating Rept,Jul-Dec 1974

ML20030A489
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/31/1974
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090740
Download: ML20030A489 (93)

Text

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CONSUhERS POWER COMPANY Docket 50-155 License DPR-6 21ST SEMIANNUAL REPORT OF OPERATIONS OF BIG ROCK POINT PLANT July 1, 197h - December 31, 197h I. INTRODUCTION - SEMIANNUAL OPERATING REPORT Our 12th refueling outage ended on July 27, 1974 and marked the beginning of Cycle 13 which produced a total of 228,796 MWhe gross through December 31, 197h. The unit was base loaded at approximately 63 !Ge gross. No plant outages occurred during this period with December 31, 197h marking 157 days of consecutive power generation. ( I-1

(, 'H ,) 's i 'II. OPERATIONS

SUMMARY

A. Changes in Plant Design Facility changes are as follows: 1. C-250A This change involved relocation of the stack area warning siren to the warehouse area to provide coverage to personnel within the ware-house and, at the same time, provide greater coverage in the outdoor area. Included in this change was the extension of plant public ad-dress system speaker power to the warehouse so that the warehouse speakers would have a power supply in common with the total public address system 4 (A-C emergency lighting power supply). l i-2. C-263A This change relocated electrical outlets and storage shelves above I the Instrument Lab workbench to allow safer and'more efficient utilization t of the work area. Prior to the change, the outlets were located above the j shelf which made the use of the outleta inconvenient. Design function is not altered by this change and it can be concluded that the design safety of the electrical outlets is not reduced; rather, it is increased. 3 C-270 This change added an open-ended sight glass water column on the shell side of the core spray heat exchanger. This water column will give. j indication of tube-side leakage (during special testing) by registering an increase in the shell-side water inventory with the tube bundle pressurized in the event of a tube leak. The design function and safety of the core spray heat exchanger was not altered by this modification. b. C-271 This facility change Gdded cathodic protection tect boxes for i more accurate test results to ensure full cathodic protection of underground plant piping and the underground section of the sphere. A preliminary sur-vey of the cathodic protection system indicated that the protection had de-creased to the extent to require either additional anodes or a rectifier system to maintain full cathodic protection. Since much of the piping is located beneath the asphalt pavement, an accurate measurement of the cathodic I protection was difficult. In order to facilitate the measurements, the installation of six test boxes was needed. This change did not affect the II-l t

i design safety or function capabilities of the cathodic protection system. It will,~ in the long run, however, increase the reliability of the system by allowing easier access to obtain more meaningful test results. 5 C-275 This change involved the addition of an oscillator and control relay to the plant public address system to provide a " Tone-Call" (4500 Hz) from the control room. This system will be used by the control room operator to provide instant alert to Operations personnel throughout the plant should an oper-ating problem or emergency arise. The " Tone-Call" does not remove any of the original features of t the system and provides priority over the telephone (PBX) entry the same as the control room handset. 6. C-276 This facility change allowed fbr the moving of the temporary main-tenance trailer offices from an area near the rear of the plant to one nearer the shop area. The movement was based on insurability factors and, in the j eyes of the insuring agency, statistically decreases the potential for per-j sonnel hazard. The design safety of the plant is not affected. 7 C-281 This facility change involved the addition of a speaker-amplifier and two speakers placed in the Service Building Annex (fermer Information Center) to the plant public address system. The speakers are included in Zone II of the plant system (office areas) and were deemed necessary since the building is now occupied by permanent plant personnel, i B. Performance Characteristics At the start of the report period, the plant was shut down for the twelfth refueling of the reactor. All irradiated fuel and core inter-i nals were temporarily located in the spent fuel pool. Representatives of General Electric were on site for inspection of failed F-type fuel assem-blies (refer to Semiannual Report #20). Removal of two reactor surveillance coupons for subsequent ship-ment to the Naval Research Laboratory, Washington, DC, and visual inspec-l ( tion of accessible coupons within the vessel were completed on July 6. It j II-2 L J ~

I was determined from the binary codings on the coupon buckets that two coupons (#120 and #121) were in reverse positions in the reactor vessel and that coupon #123 was in the accelerated vessel position. Reconstitution of reactor internals we aompleted on July 9 I Twelve core periphery fuel assemblies were loaded with nonfueled corner rods. Irradiated neutron source rods were loaded into two new G-type l fuel assemblies f9r operation during cycle 13 l Fuel loading was completed on July 18. All bundles were visually checked for location and orier.tation within the reactor vessel. Following successful shutdown margin tests, the reactor was taken critical with the I head off. Saturation checks of the in-core detectors revealed that a total of seven had failed during handling. This was in addition to two previously known failures and left a maximum of 15 out of 24 detectors operational for cycle 13 The reactor vessel head was installed on July 20 and the Temperature Coefficient Test was conducted on July 21. Results showed an increase of 10.64 in core reactivity while increasing primary system temperature from 70 F to 120 F. The temperature coefficient was negative above 120 F. During the hydrostatic test of the nuclear steam supply system, leaks were found at the control rod drive A-2 and at in-core detector #18. The control rod drive leak was stopped, but the in-core leak necessitated removal of the reactor vessel head. The in-core pressure seal seat was found to be scored in several placcs, and the assembly was replaced with an electronically failed assembly. The maximum number of operational in-core detectors was thus reduced to 13 Following reinstallation of the reactor vessel head, the primary syste.t hydro was successfully accomplished on July 26. During the outage, the two 3-way solenoid valves controlling operation of the contolnment supply ventilation valves were replaced with solenoid valves having larger air exhaust ports. The change was necessary to return operation of the supply vent valves to within the Technical Specifications limit of equal to, or less than, six seconds. (Technical i Specifications Change h2 allowed temporary operation of these valves.) II-3

i I m ( Also_'during the outage, a six-inch tie line and two motor-operated isolation valves were : installed between the fire system and the condenser, s hot well to assure a supply of water to the reactor feed pumps during a postulated small break Loss of Coolant Accident. The system is actuated. byftwo manual' switches from the control room. This change is temporary until the Reactor Depressurization System is installed during the latter part of 1975 The plant was returned to power operation on July 27 Power was escalated slowly according to General Electrir 's recommendations for fuel l' preconditioning, Output reached 200 MWt on August'26. The initial off-gas release rate at this power level was 950 pCi/s. Plant output is being administratively limited to a nominal 63 Mnsc (200 MW ) for core reactivity t . considerations in order that cycle 13 can be extended to autumn 1975 After power operation was reached, the total number of in-core detectors i determined to be in service was 12. Plant output was maintained at a nominal 63 MWeg for the remainder of the report period.except for three occasions. On September 28, power was reduced temporarily to h9 MWe to permit removal from service of the No 1 condensate pump for replacement of two upper motor thrust bearings. This power reduction lasted approximately 12 hours. On November 21 and on November 23, power was reduced to 10 MWe to permit access to the condenser pipe tunnel for repair of the turbine intermediate pressure extraction line to the intermediate pressure feed-vater heater. The power reductions lasted approximately 7 hours and 't Fours, respectively. On September 17, er y ac r. detector was assessed as having failed i 4 electronically and was remo, vi 11 ~.. er vice. On October 6, another in-core i detector had failed and was also removed from service. Subsequently, due to the reduced number of operational in-cores (10), plant operation was administrative 1y controlled to meet 80% of the Technical Specifications thermal hydraulic limits (MCHFR, MAPLEGR, etc). Thic resulted in placing the plant in.a coastdown mode from November 5 to November 26, at which time the administrative limit was reevaluated and set at 90% of the Technical L Specifications. Operation was resumed at 200 MW and maintained at that t fpoint for the remainder of the report period. e II-b i ~. -., -,.

'g The semiannual. containment component leak rate test was conducted from September 19 to September 28 and resulted in a leakage rate equal to 17% of that allowed by 10 CFR 50, Appendix J. Two shipments of irradiated fuel assemblies were made to Nuclear Fuel Services, West Valley, NY, on November 5 and November 26. These assem-blies (13) had been removed from the reactor at the end of cycle 11. Three shipments of irradiated fuel rods were made to General Electric's Vallecitos Nuclear Center, Pleasanton, CA, on July 3, Nove=ber 18 and November 26, respectively. These included 11 uranium rods removed from Bundle F-3h at the end of cycle 11 and 8 mixed oxide rods from the EEI-M02 fuel program. t The off-gas release rate was maintained at approximately 1,200 pCi/s from August 26 until December 1,1974. At that time, a correction factor, based upon a reevaluation of the specific gravity of radiolytic gas / air in-leakage mixture, was included in the off-gas release rate calcu-lation and resulted in an increase in the calculated off-gas release rate to 1,700 uCi/s. The release rate remained constant through the end of the report period. C. Changes in Procedures Manual Which Were Necessitated by Preceding A and B or Which Otherwise Were Recuired To Improve the Safety of Facility Operations The following procedural changes were made with respect to plant operations: Change No Section and System Change 73-7h Dl.6.0 Fire System Annunciator Rewrite D1 9 0 Condensate Demin System Annunciator Rewrite D1.10.0 Radvaste System Tabulation Rewrite D1.ll.0 Diesel Fire Pump Tabulation Review 75-Th B8.3.lh Thru B8.3.15.3 Added Fire Water to Main Condenser 76-7h A3.7.3-A3 7.h Addresses Use of Visitors Log 78-Th Dl.h.0 Station Service System Annunciator Tabulation Rewrite 79-Th A7 1.12 Handling of Maintenance and Facility Design Changes 80-7h A3.5 QA Department Changes, Reporting Responsibilities 81-Th A3 9 Revised Plant Review Committee Makeup and f. Responsibilities 4 II-5

{ Change No Section and System Change 83-7h Ak.3.1 Redefined " Cold Shutdown" 84-7h A4.3.5 3 Added Section on Semiannual Reporting' 85-74 'B9 3.3 Thru B9.'3 3.10 Deleted Under Fuel Pool System 86-74 A10.0 New Section Added on Commitment Reporting 87-Th' B29.0 Referenced Hydro Test to Special Test Procedure - Removed From Procedures Manual 88-74 B8.3 16.2.2 Deleted Testing of Lights on Motor Operators in Fire Line to Hotwell 89-Th Bl.3.3.h When Picoammeters Are Recalibrated, Stickers Are To Be Placed on Recorders Signifying Date, Time and Initials 93-74 B29.2.h Action Points - Primary System Coolant Chemistry 9h-Th Bl.2.h.6 Deleted Plotting Off-Gas and Extrapolation Over h8 Hours l 97-7h B29.1.2-329 3 2 Chan6ed Section on Removing Reactor Recirculation Pump From Service l 98-Th B8.3.16.1.1 Revised To Observe Telltale Leakage From Once/Round to Once/ Shift i 99-74 A3 9 6 Clarified Record of Review of All Procedures Changes 10&.4 A3.8.5 Designates Responsibility for All Drawing Changes 102-74 E2.0(d) Radiation Area Redefined 103-74 E3.1.11 New Section. Locking of Vacuum Cleaners Used in Radiation Areas 105-7h B2.2.6.1 Specifies Frequency of Cleaning Refueling Tools and Reactor Deck During Heavy Use 106-7h Bl.h Defines Normal Shutdown From Power 107-74 A7 0 Revised To Show How Maintenance Orders Are Processed ill-Th A8.9 Added Changes in Filing of Procedures - Includes Reactor Engineering Files 114-74 A2.2 3 Emergency Call Signal - Operations 118-74 Places Radwaste Batch Discharges to Canal on Special Procedure-119 B22.3.3 7 Raises Normal Operating Pressure Air Ejector Steam Supply to 250 Psig 120-75 A6.1.2 Defines Airlock Use (Inner Door) 121-Th A2.2 Specifies Manning of Control Room 127-Th-A3 7 New Section Added on Key Control ] II-6

I Change No -Section and System Change 128-74 Bl.h Specifies Use of Control Rod Drive Withdrawal Sequence Cards -131-74 Bl.0 In-Core Flux Detectors.and Flux Wire Use 132-74 B6.h.3 & Dl.3 Emergency Condenser Use With Tube Bundle Uncovered 13h 'f4 Bil.0 Radwaste Procedure Rewrite .D. Results of Surveillance Tests and Inspections Required by Technical Specifications The following listings show the systems tested, the required test frequency, the dates tested during this report period and the results of the tests. 1. Containment Isolation a. System - Containment Isolation Valve Controls and Instru-mentation. Required Frequency - Quarterly (conducted monthly). Test-Dates - T30-01 was performed on July 26,197h; August 26,197h; September 2h,1974; October 22,197k; November 19,197h; and December 17, 1974. ] Results - The automatic controls and instrumentation for eight of nine isolation valves were checked and found to function properly. One valve (main steam drain, MO-7065) is maintained in the closed position, de-energized and not used. Therefore, testing the automatic controls of this valve is not required. b. System - Isolation Valve Leak and Operability Test. Required Frequency - Twelve months or less. Test Date - Test T365-Oh was not required during this 1. report period. System - Containment Sphere Penetration Inspection.(Visual). c. Required Frequency - Twelve months or less. Test Date - This test was not required during this report period. d. System - Containment Sphere Integrated Leak Rate Test. Required Frequency - Every two years. Test Date - This test was not required during this report' period. I II-7

1 e. System - Containment Sphere Component Leak Rate Test. i Required Frequency - Six months or less. Test Date - Test T180-01 was performed on September 30, 1974. Results - The containment component leak was conducted; the - results. of this test showed a total leakage of 17% of the allowable limit. Problems were encountered briefly with the bursting of a rupture disc on the test fixture. Recall that the previous procedures called for running the test at 20 psig. A new test fixture was devised utilizing a relief valve. 2. Control Rod Drive System and Associated Tests a. System - Reactor Safety System Scram Circuits (Not Requiring i Plant Shutdown To Test). Required Frequency - One month or less. Test Dates - T30-01 was performed on July 26, August 26, September 2h, October 22, November 19 and December 17, 1974. f Results - Satisfactory. Minor problems were encountered with

1. 1 picoammeter range switch as it was showing signs of dirty contacts during

+ the August, September, October and November tests. It is scheduled for re-pairs at the next shutdown. b. System - Control Rod Performance - Run. Required Frequency - Each major refueling and at least once every six months during power operation. Test Date - July 14, 1974. Results - The control rod drive (CRD) continuous withdrawal and insertion test, including withdrawal timing, was performed for each CRD. This test was performed during the reactor refueling outage following com-pletion of other CRD performance tests and adjustments and represents the results of the final timing of each CRD under cold conditions. Tra results of this test showed all CRD to be operating satisfactorily with most with-drawal times between 36 and 38 seconds. No withdrawal time was set less than 36 seconds. c. System - Control Rod Performance - Jog. Required Frequency - Each major refueling and at least t every six months during power operation. Test Date - July 14, 1974. Results - Satisfactory latching and unlatching of all CRD. II-8

d. System - Control Rod Performance - Scram. t Required Frequency - Each major refueling and at least once every six months during power operation. I Test Date - July -13,1974. Results - The CRD scram test was performed for each CRD. The-test included time from system trip to 100% of insertion. The results of this test were all within the Technical 3pecifications of < 2.5 seconds for 90% travel. The longest time recorded was 1 30 seconds. System - Reactor Safety Systems Scram Circuits (Requiring e. Plant Shutdown). Required Frequency - During each major refueling outage but I not less frequently than ence every twelve months. Test Date - Test T365-13 was performed on July 21, 1974. Results - Satisfactory. f. System - Reactor Safety System Response Time (Requiring Plant Shutdown). Required Frequency - During each major refueling shutdown, but not less frequently than once every twelve months. l Test Date - Test T365-lh was not performed during this report period. g. System - Control Rod Withdrawal Permissive Interlocks Function. Required Frequency - Twelve months or less - the refueling interlocks will be tested prior to each major refueling. Test Date - Test TR-02 was performed on July 23, 1974. Results - Satisfactory. h. System - Control Rod Drive Friction Test. Required Frequency - During each major refueling, but not less frequently than once each year. Test Dates - Test TR21 was performed on July 14 and July 25,1974. Results - The July 1h test was run following core internals reconstitution and the July 25 test was run following core reloading. Both were satisfactory. II-9

t 3 Emergency Cooling a. System - Core Spray System Check Valves. Required Frequency - Twelve months or less. Test Date - Test T365-05 was not performed during this period. b. System - Post-Incident Spray System Automatic Control Operation. Required Frequency - Twelve months or less. Test Date - Test T365-10 va.s performed on July 13, 1974. Results - Reactor containment building sprays electrical i circuits were tested satisfactorily. System - Reactor Emergency Core Cooling System Trip Circuit. c. Required Frequency - Twelve months or less. Test Date - Test T365-10 was performed on July 13, 1974. Results - Satisfactory. d. System - Containment Sphere Isolation Trip Circuits. Required Frequency - During each major refueling shutdown, but not-less frequently than once every twelve months. Test Date - Test T365-15 was performed on July.13,1974. I Results - Satisfactory. e. System - Emergency Condenser Outlet Valves Test. Required Frequency - Twelve months or less. Test Date - Test T365-17 was performed on July 21, 1974. l Results - Satisfactory. f. System - High Energy Piping Leakage Inspection. Required Frequency - Monthly. Test Date - This is a newly acquired Technical Specifications requirement. Test T30-15 will be run monthly while the turbine generator is in service. Results - The tests were begun at the first of the calendar year (January 1975) and were not perfbrmed during this report period. g. System Primary System Leakage Test. Required Frequency - Daily. Test Dates - The calculation was made daily from July 27 through December 31, 1974. l-Results - Satisfactory, ie < 1 gpm leakage unidentified. II-10 _, _ ~

f 1 'I i '4. Miscellaneous Systems a. System - Reactor Shutdown Margin Test. Required Frequency - After each refueling, after certain core component changes if the system-is cooled to atmospheric conditions and after 35,000 mwd /T have been generated. l Test Date - Test RE-08 was performed on July 19, 1974. Results - The shutdown margin of 0.003 Ak/k with the strongest control rod fully withdrawn from the core was verified. In addition, the shutdown margin of 0.003 Ak/k was verified with two adjacent control rods fully withdrawn from the core. b, System - Nil Ductility Transition Temperature (NDTT) Calculation. Required Frequency - At least once each year. Test Date - The calculation was not made during-this report period. c. System - Moderator Temperature Coefficient Test. - Required Frequency - Following each major refueling outage. 5 Test Date - Test performed on July 21, 1974. Results - From ambient temperature to 123"F, the reactivity added to the core was 2 90 x lo-Ak/k with the beta fraction equal to -3 5.113 x 10 Ak/k. At this temperature, the increase in system temperature produced negative reactivity insertions. d. System - Suberiticality Checks. Required Frecuency - During core alterations which increase reactivity. Test Date - July 1974. I Results - Satisfactory. These checks are incorporated into the core reloading procedure and are performed throughout the reconstitution of the core with fuel bundles. Two fission chambers were utilized in-core for extra suberitical neutron multiplication visibility as two control rods were withdrawn in the area of the reactivity change; one to notch 06, the adjacent control rod fully withdrawn. By monitoring instrumentation re- - sponses, verification was made that no critical condition was approached during core loading, i II-11

L e. System - In-Service Primary System Inspection. Required Frequency - A continuing program being conducted during some major refueling outages. i Test Dates - July 3 - July 12,1974. Results - Acceptable; see Special Report No 20. f. System - Refueling Operation Controls. Required Frequency - Each major refueling. i Test Date - Test TR02 was performed on July 23, 1974. Results - Satisfactory. g. System - Reactor Refueling Safety System Sensors and Trip Devices. Required Frequency - Each major refueling. Test Date - Test T365-13 was performed on July 13, 1974. Results - Satisfactory. h. System - Recire Pump Valve Interlock Test. Required Frequency - Twelve months. Test Date - Test T365-22 vac not performed during this report period. -5 Poison System a. System - Liquid Poison System Firing Circuit Test. Required Frequency - Two months o'; less. Test Dates - Test T60-01 was performed on August 26, October 31 and December 30, 1974. Results - Satisfactory, b. System - Explosive Valve From Equalizing Line. Required Frequency - Twelve months or less. Test Date - Test T365-ll was not performed during this report period. c. System - Explosive Valve From Nonequalizing Lines. Required Frequency - Twelve months or less. Test Date - Test T365-12 was not performed during this report period. f II-12

4 1 u ' ~ 6. Radiation Monitoring a. System - Air Ejector and Off-Gas Monitoring System. Required Frequency - One month or less. Test Dates - July 25, August 22 September 26, October 2h, November 26 and December 20, 1974. Results - Checks showed the calibration to be satisfactory 3 '(within 12% of the 2 5 x 10 cps isolation valve trip setting). The' auto-matic closure function of the isolation valve timer was checked and showed the timer calibration to be satisfactory (within 3% of the maximum timer { setting) and the isolation valve closed as specified. The ionization chambers on the off-gas volume chamber were externally calibrated and the results agreed very favorably with the' 1964 results. b. System - Calibration and Functional Test of the Stack-Gas Monitoring System. Required Frequency - One month or less. Test Dates - July 26, August 22, September 26, October 2h, November 26 and December 20, 1974. 1 Results - The stack-gas monitoring system was checked using the built-in Us-137 calibration source. The instrument check showed the cali-bration to be satisfactory, resulting in the alarm occurring within the speci-fled 0.1 curie per second release rate. An additional calibration of stack-gas monitoring system is a comparative calibration used to demonstrate opera- ) tions of the monitor and to detect gross calibration changes and/or instrument drift. All calibrations were within the acceptance criteria of + 30% since recalibration of the monitor with standard liquid sources. c. System - Analyses of Stack-Gas Particulate and Iodine Filters. Required Frequency - Weekly. Test Date - The analyses were conducted weekly. Results - The results of analyses of the stack-gas particulate filter and iodine filter are reported in terms of curies released in Appendix A of this report. 4 II-13

^ ] d. System - Calibration of Emergency Condenser Vent Monitor. Required Frequency - One month or less. - Test Dates - July 25, August 22, September 26, October 26, l November 27 and December 19, 1974. Results - The emergency condenser vent monitors are checked by comparing with a calibrated portable instrument. The checks showed the vent monitor calibrations to be satisfactory with all monitor checks within 1 10% of full scale. Alarm points were found to be less than 10 mR/ hour plus background. e. System - Calibration of Canal Liquid Process Monitor. Required Frequency - One month or less. Test Dates - July 26, August 22, September 26, October 2h, t November 26 and December 20, 1974. Results - The calibration of the canal liquid process monitor . is a comparative calibration used to demonstrate operations of the monitor and to detect bross calibration changes and/or instrument drift. All cali-brations were within the acceptance criteria of 130% since recalibration of the monitor with standard liquid sources. f. System - Canal Liquid Collection Sample. Required Frequency - Daily. Test Date - The analyses were conducted daily. Results - Satisfactory. 3 g. System - Calibration of area monitors. Required Frequency - One month or less. Test Dates - July 25, August 22, September 26, October 26, November 27 and December 19, 197h. Results - The area monitor calibrations are checked by comparing readings eith a calibrated portable instrument. The checks showed the area monitor calibration to be satisfactory with most monitors within 110% and all monitor calibrations within i 15%. I h. System - Calibration of All Liquid Process Monitors (Except Canal Monitor). Required Frequency - Three months or less. ( Test Dates - July 26, August 22, September 26, October 2h, November 26 and December 20, 1974. 4 II-lh i'

'f J Results~ - The calibration of the liquid process monitors (except the canal monitor which is reported separately) is a comparative l calibration used to demonstrate operation of the monitor and to detect i gross' calibration changes and/or instrument drift. All calibrations were within the acceptance criteria of + 30% except the condensate monitor which required a recalibration with liquid standard sources. 7 Instrumentation a. System - Start-Up Channels No 6 and No 7 Operations Checks. Required Frequency - Each major refueling. Test Date - Test TR19 was performed on July 12, 1974. Results - Satisfactory. b. System - Fissi. Counting Instrumentation Calibration and Neutron Response Check. Required Frequency - Each major refueling. Test Date - Test TR20 was-performed on July 11, 1974. Results - Satisfactory. E. The Result of Any Periodic Containment' Leak Rate Test Performed During the Report Period The biannual containment' integrated leak rate test was not required j during this report period. F. Technical Specifications Changes [ Durin6 this report period, Technical Specifications Changes were 3 authorized by the Co==ission as follows: Change h3 - This change allows for removal of the four lover rollers on each of 15 peripheral control rods at the plant. There are 16 peripheral control rods with the lower rollers previously removed from one ro d. Change kh - This change sets operating power level limits in terms of maximum average planar linear heat generation rate (MAPLHGR) for the various types of fuel assemblies used in the reactor and describes the re-quirement for a fire system water supply to the main condensate system to cope with an LOCA. Change h5 - This change adds interim surveillatee requirements for high-energy piping system integrity in the area of the turbine building pipe tunnel. II-15 N - - - _, - ~,, -n---- w --,--,,n -a e

l-G. Changes in plant operating organization involving key supervisory personnel: 1. Mr. Edward M. Evans, age 22,and a recent graduate from Michigan Technological University (June 197h) with a BS in Electrical Engineering, was added to the staff in July 197h. His present posicion is as a i " Graduate Engineer." 2. Mr. E. F. Peltier, Assistant Shift' Supervisor, having achieved his Senior Reactor Operator License, was promoted to Shift Supervisor on July 1,1974. 3 Mr. James J. Zabritski, Quality Assurance Engineer at Big Rock, achieved his Senior Reactor Operator License on July 1,197h. Also, as of July 1,197h, the position of Quality Assurance Engineer reports directly to the Quality Assurance Administrator in the Corporate Office instead of the Site Plant Superintendent. h. The following promotions were not reported in the Semiannual Report fbr the period of January 1 - June 30,197h and reflect additional g staff personnel, mainly added to assist key personnel: Mr. H. E. Black, Plant Repairman "A",was promoted to a. position of Assistant Maintenance Supervisor on February 1. b. Mr. T. M. Brun, Senior Plant Technician, was promoted to the position of '.scistant Chemical and Radiation Protection Supervisor on February 1. ( II-16 4 e- . - -.--~

( III.- POWER GENERATION Report Period Total to Date 1. Tnermal Power Generated (MWh ) 726,448 13,319,586 t -2., Gross Electric Power Generated -(Wheg) 228,796 4,2h1,702 3-Net Electric Power Generated (M@e) 216,h54.8 -h,017,611.8 k. Hours Critical 3,797 0 ~ Th,86h.8 5 Hours Generator On-Line-3,785 7 73,0h1 9 6. Maximum Dependable Capacity (We Net) 71 71 7 Reserve Shutdown Hours 0 0 i f - III-1 g,,- r --,--m -m-- s- ---a-- e

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t- +im. i ; +: _. T. g t u.v 25 510152025 .s 510152025 S1015202; i t..s. m. > a t.., g wis, StJ. J ANU ARY FEBRUAPY MARCH APRIL d %T JUNE JULY AUGUST S E PTE u BE R OC 'C B E R NOVEMBER DEC E M B E R O N g Outages M 1. Maintenance Outage for Emergency Condenser.Npairs

======. 2. Eleventh Refueling Outage g 3 Maintenance Outage Extended To Include the Twelfth Refueling E .1cm r---

.. I t IV. SHUTDOWNS A. Type - Forced Unit Off Line: At beginning of report period (June 2,197h - 23h3 hours). Unit On Line: July 27, 1974 - 0724 hours. Length of Outage: 1303 Hours, 41 Minutes. Discussion: The plant was forced out of service during the previous report period when a steam leak in the 3" stage drain from the HP section of the turbine to the HP feed-water heater resulted in a controlled, delib-erate shutdown. An attempt to start up on June 5, 197h, with Control Rod Drive (CRD) C-h valved out of service,vas unsuccessful as a second CRD (B-5) also could not be withdrawn. This occurred during CRD timing tests. An orderly return to cold shutdown was initiated, the reactor head was removed and in-vestigations were conducted. Because the 12th refueling outage was scheduled for July 19,197h, it was decided to advance the refueling operation and incorporate the current outage with the 12th refueling of the reactor. I IV-1

V. SAFETY-RELATED MAINTENANCE l NOTE: Dates contained in this section generally refer to the weekly period when the maintenence was performed. [. A. Reactor Protection and Control System Instrumentation 1. Neutron Monitoring Channel No 2 7/18/7h - The compensated ion chamber in this channel was a. replaced to correct slight overcompensation characteristics of the chamber in service. The chamber in service had been installed earlier in the ouwe along with new cables when high-humidity conditions were suspected at the chamber cable location. Prior to chamber replacement, the chamter was response checked (as well as the new chamber following replacement) with the neutron source and responded properly. During shutdown conditions, chamber currents are in the order of 10-amperes and the slightest degrada-l tion of insulation resistance can cause this type of problem. l The reactor was in shutdown conGtion at the time of failure and during the repair period. b. 11/lk/Th - The picoamme eer in this channel was replaced with a spare unit following variations in the recorder trace of 5 to 10%. Bench testing of the unit removed revealed no significant problem other than mar-ginal electron tubes. Subsequently, the problem returned (the next day) and trouc.ie- . shooting of the channel was confined to range switch exercising and cable connector inspection. The problem was alleviated by repositioning of both calibration control potentiometers to a new position while maintaining the same output level. The potentiometers will be tested for noise at the next range switch inspection. Technical Specifications and plant design allow for the temporary removal of one power range flux monitor from service without compromising safety. 11/27/7h - The dual high-voltage power supply in this chan-c. nel was replaced with a spare unit following variations in the negative (polarizing) output of 300 to 800 volts. No effect was evident on the power level reading, however, as the chamber is saturated at less than 100 volts t with the reactor at power. V-1

Bench testing of the failed unit resulted in electron tube replace-ment and cleaning of contacts in the 30-second (varm-up) time delay relay. Technical Specifications and plant design allow for the recmval of one power range flux monitor from service without compromising safety. 2. Neutron Monitoring Channel No h 7/25/Th - Upscale drift in this channel was corrected b,' a. replacement of the coaxial cables from the chamber drive head to the chanber and heat-drying of the chamber connector posts. Subsequently, it was noted that the drift was recurring. Further maintenance consisted of additional heat-drying and sealing of the connector posts and reinsulating the chamber to reduce leakage to ground. This problem occurred while the reactor was in shutdown condition and was attributed to the high-humidity conditions in the reactor contain-ment atmosphere. A decrease in the insulation resistance of the signal cable to 0 10 ohms vill cause approximately one half decade increase in reading of the most sensitive portion (10~ % power) of the Log-N amplifier. The signal 1 cable resistance is usually greater than 10 ohms with the chamber attached and in its normal operating position (during plant shutdown conditions). 3 Neutron Monitoring Channel No 5 12/26/7h - The Log-N period amplifier in this channel was a. replaced with the spare unit following erratic behavior of the power level indication. Bench testing of the failed unit resulted in repla'ement of defective electron tubes. Failures of this type are within the decign limi-tations of the equipment. The Technical Specifications and plant design do not require this instrument to be in service when reactor power is above 5% rated power. h. Neutron Monitoring Channel No 7 7/18/7h - The log count rate meter in this start-up channel a. was repaired following indication of low test counts in both the 10 and 10 test positions. Repairs consisted of replacement of a defective filter capacitor, transistor and regulating diode in the -150 volt internal power supply and i a weak electron tube in the bistable multivibrator circuitry. Failures of this type are within the design limitations of the equipment. V-2

g 5 Fuel Loading Neutron Monitor No 8 a. 7/25/7h - The log count rate meter in this channel was re-paired following erratic operation. The unit would break into oscillation, causing an extreme upscale talft in counting rate. Bench testing of the unit resulted in replacement of a defective preamplifier tube socket. Failures of this type are within the design limitations of the equipment. This failure occurred during reactor shutdown prier to performing reactor shutdown margin verification, at which time the fuel loading moni-tors are placed in the reactor vessel above the core to provide additional monitoring (to the out-of-core instrumentation). B. Radioactive Effluent Monitoring Systems - 1. Air Ejector Off-Gas System 8/1/7h - Facility Change C-26h was completed on this system. a. This change consisted of replacing the original off-gas flow transmitter with a new unit to improve instrument performance. (See Facility Change Section.) b. 8/15/Th - The booster relay in the off-gas flow transmit.er was repaired following a decrease in transmitter output and a gradual loss of sensitivity. Repairs consisted of replacing the gasket between the booster relay and the relay pressure cavity. c. 10/10/7h - Checked calibration of the off-gas flow trans-mitter as a follow-up step following replacement (Facility Change C-264). As-found calibratica in the normal operating range was within 2% of initial calibration; minor adjustments were required at the low end to correct for zero (0) flow indication. 2. Stack-Gas Monitoring System ll/lk/Th - Repaired the stack-gas radiation recorder chart a. drive by replacement of the paper drive worn gear. This failure did not impair the indicating or alarm functions of this recorder but affected only the chart paper drive mechanism. b. 11/27/7h - The gross isotope channel of this recorder was repaired following loss of sensitivity. Repairs consisted of adjustment to the pen carriage drive cable, sensitivity adjustment and a thorough V-3

5 cleaning of the traverse block and rod. The loss of sensitivity was apparent on minor changes in count rate and did not affect overall re- . corder response. ~ Removal of this system from service is permitted by the Technical Specifications provided repairs are promptly mad' and the system returned to service. 3 Liquid Process Monitoring System a. 8/8/7h - Main Condensate Monitor - The linear count rate meter in this channel was replaced with a spare unit following downscale failure. Bench testing of the failed unit resulted in replacement of a shorted regulating diode in the count rate meter alarm circuitry, the power fuse and three overheated resistors in the +150 volt power supply to correct the problem. Additional repairs during performance testing and calibration consisted of replacement of one electron tube and range switch component adj ustment. Following calibration, the unit was returned to " ready spare" status. A failure of this type is considered to be within the design lLmitations of the equipment. b. 10/10/7h - Canal Discharge Monitor - Testing of this system j using high-level pulse instrumentation was initiated (Procedure No LPM 1010). Initial operc tions indicated that a considerable amount of the inter-ference (spurious upscale spike) was eliminated; however, full evaluation of the data has not_been obtained. (For additional information on this subject, a see Semiannual Operations Report No 20, Section V.h.e.) c. 10/17/74 - Canal Sample Pump - The spring tension on the canal sample pump shaft sr. was increased following failure of the pump to pump at rated capacity. d. 10/31/74 - Reactor Cooling Water Monitor - The detector for ' this monitor was repaired following low response during monthly calibration. Bench testing of the scintillation detector revealed that the output was considerably lower than comparable detectors, indicating a possible loss ( of photomultiplier gain. A spare scintillation detector unit was installed. The detector was returned to the vendor for testing and repair. A V-b

l (- e. 10/31/7h - Condensate Liquid Process Monitor _ - The detector for this monitor was also repaired following low response during monthly calibration. The scintillation detector in this unit also exhibited low output and -will be. returned to the vendor for testing and repair. (Subse- ~ quently, information from the vendor indicated that the crystals in both units were fractured and will.oe replaced.) f. 11/7/7h - Canal-Discharge Monitor - The flush inlet valve on the canal sample monitor was replaced due to leakage from the same. ~ g. 11/7/7h - Canal-Sample Line - An extension on the canal sample line was installed to facilitate future maintenance work on that line.- Removal of these systems from service is permitted by the Technical Specifications.provided repairs are promptly nade and the systems returned to service. C. Containment Sphere Isolation Systsm 1. Containment Sphere Supply Ventilation System Solenoid valves SV-9151 and SV-9152 and the associated air lines i utilized for exhausting the valve operator air cylinders on containment sphere supply ventilation valves CV-4096 and CV-4097 vere replaced with I larger capacity components. The capability of processing a larger volume of air through these lines provided for the reduction in closing time for 2 'CV-h096 and CV-4097 from 9 and 8.5 to P.5 and 3 seconds, respectively. These valves were, therefore, returned to' full compliance with Technical Specifications. Tech Spec Change No h2 allowed temporary relaxation.of the valve closing times. The modification was described in Facility Change C-263 and performed ' uring cold shutdown under procedural control. d D. Emergency Power-Systeg 1. Emergency Diesel Generator - None. 2. Emergency Diesel Starting Motor - None. 3. Emergency Diesel Generator Starting Batteries a. 11/21/7h - Emergency diesel starting difficulties were corrected by cleaning the starting battery terminals. As a corrective {- measure, subsequent semiannual inspections of the batter 7 terminals have been.added to the BRP Preventive Maintenance Program. .V-5 ~

l h. Emergency Diesel Generator Fuel System a. 11/27/7h - The fuel oil supply pump on the diesel generator (fuel oil tank to the day tank) was replaced. Failure of the fuel oil supply pump as reported in Abnormal Occurrence Report /.0-15-Th necessitated the change. 5 Emergency Diesel Generator Output -Instrumentation a. 12/12/7h - Emergency Diesel Generator - Loss of voltage, current and frequency readings on the emergency diesel generator local panel (Control Room indication was normal) was corrected by replacing power supply-fuses to the instrument control transformer. Plant operating requirements permit removing the emergency diesel from service for periods in excess of thirty minutes with the approval of the Plant Superintendent. In each of the cases noted above, the Plant Super-intendent's approval was obtained for removing the diesel from service only for the time specifically required for troubleshooting and replacement and/or repair. E. Emergency Condenser System 1. Emergency Condenser a. 8/8/Th - Leakage from the emergency condenser inlet valve (bD-7062) packing was corrected through packing adjustment. The valve was subsequently test-operated satisfactorily, b. 9/5/7h - Leakage from the emergency condenser loop No 2 inlet valve, MO-7052, was corrected by adjusting the packing. The valve was test-operated satisfactorily following the packi g adjustment. acking adjustments can be made on these valves so long as the valves are subsequently test-operated. Test operation presented no oper-ational difficulty and thus did not affect plant safety. F. Primary Coolant System 10/3/7h - The spare reactor recirculating water pump seal cartridge (Cartridge #3) was rebuilt utilizing a newly fabricated shaft sleeve seal (the replaced sleeve had been badly eroded as a result of excessive leakage). The newly rebuilt seal cartridge displayed an acceptable controlled leakage rate of 0.676 gpm under a test pressure of 2hD psig. The seal was rebuilt f i i using procedural control and replacement components with materials certifi-cations. This operation was performed with the plant in the cold shutdown mode. V-6

G. Shutdown Cooling System 11/7/Th - Leakage on the cooling water side of the #1 shutdown cooling system heat exchanger was repaired by tightening the flange stud bolts. Tightening the flange bolts did not affect system operability (which is required only during plant shutdown and refueling operations) and thus plant safety was not affected. H. Control Rod Drive System 1. CRD Mechanisms (CRDM) a. 7/11/7h - CRDM in Positions A-2 and B-1 were replaced with previously overhauled CRDM. CRDM Serial No 266 in Position A-2 was replaced with CRDM Serial No 268 and CRDM Serial No 252 in Position B-1 was replaced with CRDM Serial No 264. b. 7/11/Th - Drive Mechanism A-4 (Serial No 2h0) was removed for inspection relative to coupling difficulties. The mechanism was found to be in satisfactory condition. New flange 0-rings were installed and the thimble J-veld was successfully ultrasonically inspected while the drive was removed. The drive was then reinstalled with procedural controls and successfully recoupled. The above-noted maintenance activities were performed with the plant in the colf shutdown condition. In each instance, the support struc-ture module beneath the respective drive was subsequently repositioned according to the design requirements on vertical clearance and lateral alignment (see Item e below). c. 7/ll/Th - Control Rod Drive Blades - The lower rolleru jere removed from fifteen (15) control blades with procedural controls as directed by the Reactor Engineer. This operation was performed with the plant in cold shutdown. d. Insert and Withdrawal Speed Control Manifolds (1) 7/11/Th - The E-6 position speed centrol manifold ball-check valves and filters were inspected relative to sluggish drive withdrawal. The manifold filters were replaced and the ball-check valves were inspected revealing no indications of malfunction. The manifold was reassembled and I.. returned to service. i V-7

1 ,1 I (2) 9/96. - Leakage from the D-3 drive speed control mani-fold-was corrected by replacement of the three manifold filters and the i withdrawal rate set valve. These repairs were conducted with the plant in cold shutdown.. CRD Valves .(3) 10/10/7h - D-3 Selector Valve - Leakage from the with-drawal selector valve was corrected by replacing the valve ball, stem, seats and 0-rings. With the reactor at power, one drive may be removed from service to permit repairs to the selector valve. (h) 10/2h/7h - Pump Relief Valves - Leakage through the No 1-and No 2 rod drive pumps relief valves was corrected by replacing both valves with reconditioned and reset valves. This repair was performed with the reactor at power and with procedural controls. One of the two CRD pumps may be removed from service and still maintain normal operational status. e. T/18/Th - Drive Support Structure and Flux Wires - An inspec-tion of the support structure modules and flux vire tubing was made (per Maintenance Procedure MCRD-1) subsequent to the rod drive maintenance ac-tivities. All components were found to be in satisfactory condition (see Items a and b above). f. 12/19/Th - Control Rod Drive No 252 and No 266 were dis-assembled, inspected, cleaned and rebuilt per Maintenance Procedure MCRD-2. Both appeared to be in very clean condition with no appe. rent discrepancies. Both were returned for subsequent installation as replacement mechanisms. 2. CRD Filters 8/1/Th - The'No 1 control rod drive filter was replaced due a. to high differential pressure, b. 9/6/Th - The No 2 rod drive filter was changed due to high differential pressure, 9/19/74 - The No 1 rod drive filter was changed due to high c. differential pressure. d. 10/3/7h - The No 2 rod drive filter was changed due to high differential pressure. 10/17/Th - The No 1 rod drive filter was changed due to e. high' differential pressure. y_8

-j. f. 10/2h/7h - The No'2 rod drive filter was. changed due to high differential pressure, g. 12/26/7h - The No 2 control rod drive filter was changed due to high AP and the bottom 0-ring on the filter was replaced upon dis-closure of defects found in preinstallation -inspection. The filter changes above were made with the-reactor at power. Only one of the two CRD filters is required for service under normal oper-ation. All filter changes were made per established procedural control. 1 3 CRD Accumulators a. 7/25/Th - Accumulator E-3 leakage was repaired by replacing the water cylinder bladder, b. 9/5/7h - C-5 drive accumulator leakage between the accumula-tor halves was corrected by replacing the seals. c. 11/7/Th - Rubber 0-rings and inner and outer Teflon backup rings were replaced on the C-h accumulator to stop a gas leak at the neck of the lower bottle. d. 11/27/7h - Control rod drive accumulator C-6 leakage was 1 corrected by replacing the gss-side 0-ring. e. 12/26/7h - Leakage on accumulator E-h between the upper and lower cylinder _ was repaired by replacing 0-ring seals and backup rings. 3 With the exception of the E-3 accumulator repair above, all work I was performed while the reactor was at power. Under this condition, the f primary hydraulic source for CRD scramming comes from the reactor vessel. This design feature permits repair of an accumulator without affecting reactor safety. In each case, aonly one accumulator was removed from service and only for the time required to perform the corrective maintenance. The E-3 accumulator repair was performed with the reactor in the cold shutdown condition. h. CRD Pumps a. 9/1/7h - Leaking packing on the No 2 control rod drive pump was replaced per Maintenance Procedure MCRD-5 b. 8/8/7h - Leakage from the No 2 control rod drive pump east i piston packing was corrected by packing adjustment. V-9

'I c. 8/22/Th - Leakage from the No 2 control rod drive pump piston packings was corrected by reple:ing the packing in all three cylinders. d. 9/5/7h - Excessive leakage from the east piston packing on the No 2 control rod drive pump was corrected by adjusting the packing. e. 10/3/7h - Leakage from the No 2 rod drive pump was corrected by adjusting the packing. I f. 10/10/7h - Excessive packing leakage from the No 2 rod drive pump was corrected by replacing the packing in the No 1 cylinder and adding one ring of packing to the No 2 and No 3 cylinders. g. 10/31/7h - Adjusted packing on No 1 control rod drive pump to correct excessive leakage. h. 11/21/Ik - Leakage from the No 2 control rod drive pump was corrected by adjusting the pump packing.

i. 11/27/Th - Packing in the No 1 cylinder of CRD pump No 1 was replaced due to leakage.

J. 12/19/7h - The air supply line to SV-h858 on CV-4016, the CRD pump suction supply, was repaired by replacing a coupling in the line to cor-1 rect leakage caused by the tube rubbing on adjacent piping. I k. 12/26/74 - Control rod drive pump No 2 center plunger was repacked to repair leakage from the same. The above CRD pump repairs were performed with the reactor at power and with procedural controls. One of the two CRD pumps may be removed from service and still maintain normal operational status. Pump packing adjustments can be made without affecting the operational status of the pump. 5 Control Rod Drive System Instrumentatien a. 7/18/74 - Adjusted the position switch for the rod bottom light on D-2 position probe to provide better switching action. b. 7/25/7h - The control rod drive continuous withdraw (jog bypass) feature was defeated following core loading and prior to physics i testing. This will insure that the control rod drives cannot be continu-I- ously withdrawn with the reactor at operating pressure and temperature. This change was made following SARB review of the rod timing l'~1 characteristics and was performed under Procedure CRD-74-hS5(1). N Y-10

l' The reactor was in the shutdown condition during the above maintenance action. c. 8/1/Th - Intermittent operation of the control rod drive pump low-pressure alarm was corrected by cleaning the pressure switch sensing line. This allowed the switch to reset properly following a low-pressure alarm. d. 8/22/74 - Repairs were made to the print wheel assembly on the control rod drive temperature recorder following improper printing action. This failure did not inhibit the indicating and alarm functions, which were still available if a high temperature had been received. 10/10/Th - Spurious high-temperature alarms from the con-e. trol rod drive temperature re, corder were traced to a defective thermocouple (or connection) on B-1 control rod drive. The thermocouple was shorted (in the recorder case) and gave alarm indication on the remaining drives. Repairs will be made at the first convenient outage. During the interim period, manual observation of B-1 control rod drive cooling water will be performed. f. 10/31/Th - Checked calibration of the control rod drive filter differential pressure indicator and alarm (DPIS-RD39). I. Reactor Vessel 1. 7/11/7h - Reactor Vessel Baffle Plate Open Locking Devices - The bolts retaining each of the four baffle plate open locking device arm essem-blies were replaced as a result of the fa' lure of two of these bolts. The replacement bolts were fabricated from certified bar stock and NDT inspected prior to installation. The tack welds retaining the bolts in place were visually inspected after installation. This work was performed with all fuel removed from the reactor and the plant in cold shutdown. 2. 7/25/7h - In-Core Assemblies - Position No 18 in-core was removed due to flange leakage. Inspection revealed the in-core assembly seating area to be grooved in three positions, preventing a complete seal in the flange stellite seat. The flange seat was also inspected and found to be in satis-factory condition with the thimble clean and free of foreign materials. The removed in-core was stored for later disposal and a previously used assembly V-11

i with an intact seal was installed. The maintenance activities performed were controlled by Maintenance Procedures MRVI-l and MRVI-2 and carried out with the reactor in the cold shutdown mode. 3 7/25/Th - Reactor Bleed-Off Valves - Repair and adjustment of l blowdown Valves CV-h0hD and CV-bllh was performed to correct minor leakage l through the valve seats. The maintenance activity was performed with the plant in the cold shutdown condition. J. Steam Drum 10/31/Th - Packing leakage of the steam drum upper vest end level sensing root valve was corrected by adjusting the packing. The repair was performed while the plant was in operation without altering valve positioning. 1 l ll/1h/7h - Leakage from the steam drum, east end, lower water level I reference line valve was corrected by adjusting the valve packing. The re-repair was performed while the plant was in operation without altering valve positioning. K. Fuel Handling Systems 1. Fuel Pool Bridge Crane 7/3/Th - The deteriorated vinch cable was replaced and the cable travel upper and lower limit switches were reset. 9/26/7h - Failure of the fuel pool hoist brake to hold properly vis corrected by adjusting the brake shores. 2. Fuel Transfer Cask 7/3/Th - The vinch control box, switches and lead cables were replaced due to general deterioration. The vinch contactors and cable travel limit switches were inspected and adjusted as required. The vinch was satisfactorily test-operated and the transfer cask returned to service. Maintenance may be performed on the above equipment without affecting plant operation or safety. L. Main Steam Supply System 1. Main Steam Bypass Valve Hydraulic System 8/29/7h - Replaced unir,ader piston on turbine bypass valve a. and adjusted the low-and high-pressere alarm settings to correct operating / iange of the "B" hydraulic pump unloader. V-12

f t. b. ll/1h/Th - Failure of the "B" unit hydraulic pump to main-tain adequate pressure was corrected by flushing out the unloader set valve. c. 12/19/Th - The turbine bypass valve "B" unloader was re-placed with the spare. Following removal, the unloader was rebuilt and placed in stock as a spare. The main steam bypass valve hydraulic system may be removed from service for specified periods without affecting plant operation or safety. ~M. Feed-Water System 1. Reactor Feed Pumps 8/15/Th - Rough pump operation and excessive heat from the a. i outboard packing on the No 2 reactor feed pump was determined to be the result of some degree of "snt" acquired by the packing during the recent outage when the pump was not run. Proper pump operation was restored when the packing was " broken in" following start-up. In addition, correct pack-ing leak-off was established once the packing " break-in" period was com-l pleted. b. 12/12/7h - Excessive leakage on the No 2 reactor feed pump i outboard packing was corrected by adjusting pump packing. The above repairs were performed while at power without affecting plant operation or safety. i l 4 t ) s V-13

VI. CHANGES, TEST, EXPERIMENTS A. 10 CFR 50.59 Facility Changes 1. Facility Change C-263 This change replaced solenoid Valves SV/9151 and Sv/9152 which control the operation of the supply ventilation Valves CV/h096 and CV/h097 by controlling the air supplied to and exhausted from the operators fcr these valves. The volume of air which these solenoid valves can process in a given time determines the speed with which the CV valves can be closed. Technical Specifications (as well as the FHSR description) re-quire that this time is to be six seconds or less. The recently installed replacement for CV/h097 (Maintenance Procedure MCIS-1, Rev 0) utilizes an air operator which requires the processing of a much larger volume of air for opening and closing than was required by the valve replaced. Process-ing of this larger volume of air takes ten seconds as opposed to the re-quired six seconds. Te hnical Specifications Change No h2 allowed tem-porary relaxation of the valve closing times. The controlling procedure (MCIS-5, Rev 0) and this facility change outlined the requirements for replacing the former CV valve operator exhaust system with one capable of processing a larger volume of air. Specifically, the changes were as follows: Replace 1/2-inch pipe,'Ith 3/4-inch pipe. a. b. Replace 1/2-inch tubing and fittings with 3/4-inch tubing and Swagelok fittings. Replace original ASCO solenoid Valves SV/9151 and SV/9152 c. with larger ported valves. Comparative statistics on the original and replacement ASCO valves are as follows: Original Replacement Pipe Size 3/8 3/4 Orifice Size 5/8 11/16 Minimum Operating Pressure 10 Psi 10 Psi Maximum Operating Pressure 125 Psi 250 Psi Safe Body Working Pressure 300 Psi 300 Psi Maximam Fluid Temperature 77 F 180 F l CV Flow Factor 3 5.5 Voltage 125 V D-C 125 V D-C Wattage 10.5 16.8 Catalog No 831612 8316Chh VI-l

4 Review of the above valve data indicates that the newer valves require slightly higher current. However, the higher current is well within j the capabilities of the 125 V d-c supply utilized. Also, it is noted that the never valves have a slightly higher flow factor and it is for this higher flow capability that these valves were purchased. The primary effect of the higher flow capability (in conjunction with larger exhaust lines) will be to reduce the closing time of the CV valves (as required). A secondary effect will be an increase in opening times, for which there is no Technical Specifications, FHSR or plant operating requirement. Specifications on the replacement exhaust line component are as follows: Compcnent Specification 3/4-Inch Pipe Commercial Grade Carbon Steel 4 Schedule 80 Pipe Nipple, ATSM A106, Gr B 3/k-Inch Tube Soft Annealed Sesaless Copper Tub-ing, ASTM-B75 Equivalent, 576 Psi Allowable, Shelf Items 3/4-Inch Swagelok Fittings Commercial Grade Brass Fittings Manufactured to Applicable ASA B31.3, 1959; 631+ Psi Allowable The replacement exhaust line components were not purchased as certified stock but are equivalent to the original components and have specifications which far exceed the requirements. It should be noted that air control systems are not specifically treated in the ASME BPV Code nor in AEC Regulatory Guide 26. However, the ) air control aystem on the containment ventilation valves indirectly affects l the function of a Class MC (ASME BPV Code) Tessel and the replacement com-j ponents for this air control system have, therefore, been purchased with considerations designed to preclude system failure; ie, certification of solenoid valve compliance with cetalog specifications and tubing and piping sized for service pressures far in excess of actual system pressures. In addition, the controlling procedure (MCIS-5) calls for integrity and opera-tional tests which will definitely establish the acceptability of the j replacement components. VI-2

i In the light of the above discussion, it is concluded that the proposed change in no way compromises or changes the characteristics of the system as described in the Technical Specifications, FHSR or plant operating requirements. Indeed, the facility change described returns the system to full compliance with the above documents and, therefore, does not represent an unreviewed safety question as described in 10 CFR 50 59 2. Facility Charge C-26h This change consisted of replacing the original off-gas flow transmitter with a new unit to improve instrument performance. The old unit would not remain in calibration following plant shutdown and the information at normal flow rates was questionable. The new transmitter provides 0-100% flow indication for 0-21 cefm of airflow. In addition to the flow output, it also has a differ-ential pressure indicator so that the pressure drop across the flow orifice can be measured directly. Initial calibration figures indicate that the instrumentation system (differential pressure indicator, flow indicator and flow recorder) is within 2% from 0 to 100% flow; while in the normal operating range (10 to 60%) the system is within 1% accuracy. A safety evaluation concluded that on a performance basis this equipment is better than original design. Also, the static pressure rating of the new equipment is equal to the piping pressure rating of the off-gas system (150 psig) whereas the original was rated at only 50 psig. Therefore, it can be concluded that the design safety of the system would not be affected and, hence, there is no unreviewed safety question involved as described in 10 CFR 50 59 3. Facility Change C-268 This change added a fire water emergency makeup tie condenser line designed for flooding the reactor core from the time of initiation of a small LOCA to the point of depressurization of the reactor vessel and initiation of the core sprays. This interim line is to be supplied from the fire system and vill deliver fire water to the condenser hot well i via the hot well recire line. The plant Technical Specifications and VI-3

FHSR specify use of the fire system in supplying water for actuation of the core spray system assuming a fire header capacity of 1,000 gpm with either the diesel or electric fire pumps in service (capacity greater if both pumps are in service). With the design function of the interim core flooding line specified only for the period up to activation of the core sprays, there will be no requirement of the fire system beyond its designed capabilities, and both the interim core flooding line and the core sprays vill have uncontested use of the fire system during their respective periods of design function. The availability of water from the fire system is based on the following essumptions: a. Either the 46 kV or the 138 kV lines available for service. b. The electric fire pump backed by the emergency diesel generator or the diesel fire pump available for service. The condensate system (pumps and valves) available for c. service. d. The feed-water system (pumps and valves) available for

service, e.

Interim core flooding line limitorque valves available for service. f. The recire pumps are not specifically required (h x 10 lb/h natural circulation flow available) but would be available through the h6 kV or 138 kV lines. The assumption of the availability of either the h6 kV or 138 kV lines is based on the historical performance of availability of these lines. The remaining assumptions are based both on line availability and equipment availability and are justified by past operational and maintenance perfor-mance. Design and fabrication of the interim core flooding line was in compliance with the applicable portions of ANSI B31.1, Popwer Piping Code, as noted below: a. Design - All materials, vall thicknesses and pressure ratings have been previously set by the Bechtel Piping Specifications via the P& ids; ie, the hot well recire line was built to Spec D-3 and the fire header line was built to Spec E-11. The D-3 spec, being the more conservative, VI h

I was used as the basis for this construction. The fabrication utilizes all commercial components assuring no unforeseen stress risers or lack of wall thickness in weld or pipe. The total run of the interim core flooding line is supported by three pipe supports. In addition, each end of the line attaches to existing piping in the immediate vicinity of a pipe support. b. Materials - The components utilized in the interim core flooding line are standard, commercial grade shelf items: ASTM A-106 pipe, ASTM A-23h fittings and flanges and 300 lb (minimum) valves. c. Fabrication - Welding was performed per CP Co Weld Procedure MA-1 by a qualified welder. d. Inspection and Test - Maintenance Procedure MGP-6 i specifies hydrostatic testing to establish compliance with ANSI B31.1. Installation of the interim core flooding line required removal of a portion of the previously existing emergency core flood line. This line was cut off upstream of the manual isolation valves preceding the point of junction with the feed pump discharge line. No pipe cap was installed on this line, but the two manual isolation valves were closed, tagged and locked to assure maintained integrity of the feed-water header. Review of the above discussion will establish that installation of the interim core flooding line does not compromise or alter the design function of any of the associated systems or equipment as described in the plant Technical Specifications and/or FHSR, reference Technical Specifica-tions Change No hh. In addition, the installation conforms in all respects to the existing applicable piping codes. It can, therefore, be concluded j that installation of the interim core flooding line does not constitute an unreviewed safety consideration as described in 10 CFR 50 59 h. Facility Change C-269 This change provided a one-half inch hole in the control rod drive pump casing for the purpose of overflow to prevent water from enter-ing the oil crankcase in the event the casing bottom drain were to plug. Overflow water from this hole will go directly to the base drain and from there to the dirty sump collection system. In addition to overflow pro-I i ' tection, the hole will also. indicate high water in the piston housing. VI-5

This change does not affect the pump operation or design function and, in fact, gives additional safeguards to prevent water from entering the oil. It can, therefore, be concluded that the design safety of the sys-tem is not affected and there are no unreviewed safety considerations as described in 10 CFR 50 59 5 Facility Change C-278 This change replaced the 125 volt d-c station batters charger with one of a greater charging capacity. It also relocated a ground detector and an undervoltage relay. The modification does not change the design intent of the system; rather, it replaces the previous charger with one of a higher charging capacity and, therefore, does not affect the system function. RIt can be concluded that the design safety of the system is not altered and there are no unreviewed safety considerations as de-scribed in 10 CFR 50.59 B. Performed Pursuant to 10 CFR 50 59(b) 1. Reactor Coolant and Steam Phase Iodine Carry-Over The clean-up demineralizer "no clean-up test" is designed to i measure the fractional carry-over of iodine in the steam phase of a boiling water reactor. The technique is to measure the difference in the equilib-rium level of the radiciodines in the reactor water in normal operation and in operation with the clean-up demineralizer removed from operation. 4 I -From this relation, the iodine removed from the coolant water by the clean-up demineralizer can be calculated. Hence, the fractional carry-over is obtained. The value of the iodine carry-over obtained by taking the ratio of the I-131 concentration in the condensate sample to that in the. primary 4 coolant is 0.5 0.2%. Both the condensate (68%) and clean-up demineralizer (995) efficiencies were measured in these series of measurements. The results of the analyses are that the iodine carry-over is 0 7 0.3%. Based on PRC review of the Procedure NUCP (no clean-up test), it was concluded that there were no unreviewed safety questions involved an described in 10 CFR 50.59 VI-6

N VII. RADIOACTIVE EFFLUENT RELEASES A. Introduction Releases of radioactive material both to the atmosphere and to Lake Michigan from January 1 to December 31, 1974 were well within the facility licensed limits and the AEC's regulations, particularly Title 10, Code of Federal Regulations, Part 20. Table 1 of Appendix A of the Twentieth Semiannual Report has been revised to reflect the results of studies described in the November 1, 1974 letter to the Directorate of Licensing and is included, as revised, in Appendix A of this report. Briefly, these revisions are: (1) a 30-minute off-gas holdup time, (2) an increase of 38% in the off-gas flow rate, (3) a 40,000 can stack flow, and (h) removal of the arbitrary tripling of iodine releases. Also, per the November 1, 197h letter, all reported particulate releases prior to 197h should be increased by h/3; all iodine releases reported prior to 1973 should be increased by h/3 and I-131 releases for 1973 should be decreased by h/9 In addition, g all radioiodine releases to the atmosphere are plus or minus 50%. B. Gaseous Effluent Gaseous releases to the atmosphere totaled 1.88E+05 curies of fission and activation gases. This corresponds to 0.61% of the licensed Technical Specifications limit of 1 Ci/s. Particulate releases totaled 0.09 curie or 0.08% of the licensed limit while halogen releases were measured to be 0.36 curies or 0.bl% of the licensed' limit. Gross alpha measurements on the particulate filter revealed that the release of alpha emitting nuclides totaled 2.80E-CC curies. Tritium releases for the period totaled 28.7 curies or 3.00E-05% of limit based upon meteorolo6 cal dispersion to the point of maximum ground concentration. i 1. Gaseous Effluent Calculational Methods A sample of off-gas is obtained weekly during power operation i and analyzed by gamma spectrometry for **six noble gas radionuclides. Based upon the mixture of the six nuclides, a stack release rate, which i includes a total of 22 noble gas radionuclides, is determined. The stack 4 release rate is based on a 30-minute holdup time for off-gas plus a 1%

• • The six nuclides are: Kr-85m, -87, -88 and Xe-133 -135 and -138.

VII-l s. l

l contribution from the turbine sealing steam system utilizing a 2-minute holdup. The 1% turbine seal contribution has the same distribution of nuclides as the off-gas corrected for a 2-minute decay period. This is reflected in the monthly totals shown in Appendix A. Activation gas releases are composed primarily of N-13. The rate of release is power-level dependent and is incorporated in the total l monthly releases shown in Appendix A. Particulate and halogen releases to the atmosphere are measured l by counting particulate and charcoal filters weekly. These filters col-lect stack effluent continuously at a rate of three cubic feet per minute. Determination of release rates in this manner assumes radioactivity is continually being deposited uniformly throughout the week on the filters and, hence, a decay correction to the time of analysis is applied, depend-ing on the half-life of the nuclide observed. The net beta activity, as reported in Appendix A, represents the unidentified portion of the total activity present on the particulate filters (ie, gross beta activity minus j the ider.tified isotopic activity). The net unidentified beta activity is corrected for decay based on a half-life of 27 7 years (ie, Sr-90). Tritium releases to the atmosphere are calculated, based upon ocasurements made in the primary coolant and containment air and using identical concentrations for all releases as rollows: Off-Gas - The average flow rate containing 90% radiolytic a. gas by volume at primary coolant tritium to hydrogen ratio and at 100% l relative humidity is used to determine tritium releases both in vapor and molecular form. b. Turbine Sealing Steam - The design flow rate at 100% relative humidity and primary coolant tritium to hydrogen ratio. c. Containment Ventilation - The design flow rate and measured containment building tritium concentration. The results of these calculations are also shown in Appendix A. C. Liquid Effluents Liquid vaste releases totaled 1.07 curies of radioactive mate-rial. This release corresponds to 2.1% of Technical Specifications limits. [ Addit,ionally, 5.07 curies of tritium were released corresponding to 0.0016% of 10 CFR 20 permissible concentration in the discharge canal. VII-2

~. i 1. Liquid Effluent Calculational Methods The release psthway to Lake Michigan for all liquid effluents is through the plant's condenser circulating water discharge canal. A 49,000-53,200 gpm dilution for liquid effluents is obtained flow rate of through the use of the condenser circulating water pumps, two at 2h,500 gpm each and house service water pumps, two at 2,100 gpm each. Each collected tank of liquid is sampled, analyzed for radioac-tive content, and discharged at a controlled rate to assure that permissible concentrations are not exceeded in the canal prior to dilution in Lake i Michigan during the time of discharge. Each sample is analyzed by gamma spectrometry to identify as many of the component nuclides as possible, f (See Appendix B for results.) Permissible concentrations in the canal are determined from the following: C g 1 MPC1 4 .where C is the concentration of the ith isotope in the canal at the given y 3 concentration measured in the tank diluted by the known canal flow rate. t Those isotopes not identified by gamma spectrometry but measured by gross beta analysis are presemed to ' 1r-90 and released on that basis. Tritium releases are based on average concentrations in both " clean" and " dirty" waste tanks. D. Solid Wastes ~ A total of 5.kO8E+06 curies of radioactive material was shipped off site during the period covered by this report. Of the total, irradi-ated cobalt accounted for 3.113E+05 curies, spent fuel 5 06TE+06 curies and solid radwaste 9.h52E+01 curies. (See Appendix C.) E. Environmental Dose Calculations Levels of radioactive materials in environmental media indi-cate that public intake is well below 5% of that which could result from continuous exposure to the concentration values listed in Appendix B, Table II, 10 CFR Part 20. 1. Atmospheric Releanea Currently, a computer model is used to calculate radiation dose resulting from plant releases of noble gases. The integrated population VII-3 1p. m e ,y-.- y +. - - -,.m. y

dose, out to 50 miles, for the reporting period is shown on the following page. The computer model utilizes the following: a. X/Q values for the five sectors are averaged over both stability class and vind frequency. b. Doses are calculated for each of the 22 noble gas radio-nuclides and daughter products based on individual decay energies. Total dose is then the summation of the individual nuclide contributions, c. The 1974 population is estimated from the 1970 census of Population on a township basis corrected by the census-determined State of Michigan growth rate of 1.3% per year and includes transient population as 1/k residents. The total estimated 1974 population resides 24 hours per. day all year at the same location. d. The actual mixture found during the weekly off-gas analy-sis is used for that week's releases and the total release is further corrected by daily gross measurements of off-gas. e. Site boundary doses are finite cloud shine doses. Semi-infinite cloud geometry is utilized to calculate doses after the plume reaches ground level. f. No credit is taken for the meandering of the plume before it reaches the different annuli. The maximum enlculated radiation dose at the site boundary re-sulting from noble gas releases was 6.7 millirems. The integrated dose to the population out to 50 miles was 5.5 manrems. Doses from particulates and halogens releases as shown in Appen-dix A vere negligible compared to that received from noble gases due to the conservative limits in the plant Technical Specifications. 2. Liould Releases The nearest municipal drinking water supply intake is located' in Charlevoix, Michigan, which is generally upstream of the prevailing current flow in Lake FEchigan at this location. However, since current patterns do occur that could, at times, carry the discharged water in the direction of Charlevoix, a population dose based upon this flow has been calculated. A conservative dilution factor of 800 is taken from t 'the point of discharge to the city of Charlevoix based upon the report, VII-4

I 4 CALCULATED RADIATION DOSES FROM GASEOUS RELEASES January 1, 1974 to December 31, 1974 (Manrems) Distance Sector (Miles) 1 2 3 4 5 Tota) 1-2 Population 13 75 0 10 0 98 Population Dose 1.6E-02 5.0E-02 0.0 1.lE-02 0.0 7.8E-02 2-3 Population 26h 270 0 51 73 658 Population Dose 2.6E-01 1.1E-01 0.0 h.5E-02 6.7E-02 h.8E-01 3-h Population 562 397 0 h8 58 1,065 Population Dose 3.8E-01 1.3E-01 0.0 3.hE-02 4.0E-02 5.9E-01 h-5 Population 722 3,344 0 103 0 4,169 Population Dose 2.1E-01 1 3E-00 0.0 5.3E-02 0.0 1.6E-00 5-10 Population 2,102 2h 0 534 0 2,660 Population Dose h.hE-01 3.3E-03 0.0 1.1E-01 0.0 5.6E-01 10-20 Population 8,987 395 Th7 1h,115 327 24,571 Population Dose k.5E-01 1.0E-02 4.9E-02 9.2E-01 1.5E-02 1 5E-00 1 20-30 Population 9,651 3,504 1,902 h,623 327 20,007 Population Doce 1.2E-01 3.3E-02 3.9E-02 9 1E-02 5 1E-03 2.9E-01 30 40 Population 22,775 4,081 2,916 h,8h7 0 34,619 j Population Dose 1.2E-01 1.2E-02 2.5E-02 h.0E-02 0.0 2.0E-01 40-50 Populatiod 40,790 8,888 5,873 12,101 0 67,652 Populatic Jose 1.0E-01 1.hE-02 2 5E-02 5.0E-02 0.0 1 9E-01 0-50 Population 85,866 20,978 11.h38 36,hh7 785 155,409 Population Dose-2.1E-00 1.7E-00 1.hE-01 1.kE-00 1.3E-01 5 5E-00 Site Boundary Dose (Rem) 5.8E-03 3.8E-03 6.7E-03 6.6E-03 VII-5

1 " Big Rock Point Hydrological Survey, Great Lakes Research Division, University of Nuchigan, Special Report No 9," by John C. Ayers,1961. In addition, the population dose is calculated to the entire population which receives its drinking water from Lake Michigan, based on a uniform concentration, resulting from plant releases, throughout Lake Michigan. Also, radiation dose to human populations can occur as a result of plant releases through the consumption of fish caught in Lake Michigan. Utilizing the mea,sured values of radionuclides released as shown in Appendix B, the following formula, and the standard man model, drinking water doses can be calculated as follows: I C pgh '(Limiting Dose Rem /Yr) D, = E { ij vhere: D, is the individual dose in rem /yr, C is the average mcentration in Lake Michigan of the individual 1 nuclides er .ed, in pCi/ml, FPC is the concentration of each nuclide measured, required to produce the limiting dose at continuous intake in pCi/ml and the limiting dose is the dose produced at continuous exposure to MPC concentrations. In calculating ingestion dose from the consumption of fish, an equation similar to the one used for drinking water dose is used except that a standard daily diet of 50 grams of fish flesh is used in contrast to the 2,200 ml of fluid consumed daily by the standard man. This, in effect, increases the MPC by 2,200/50 or hb. g The calculation of individual doses, both from drinking water e and consuming fish, are per the previous formula while integrated popula-tion doses in manrem are calculated utilizing the following parameters: a. For drinking water, the individual doses are summed over 4 the entire population that receives its drinking water from Lake NEchigan with discharge canal flow appropriately mixed with the lake. This is approximately 10 million people of which approximately 7 million reside in the Chicago metropolitan area. VII-6

b. The population dose due to drinking water to Charlevoix residents is based on a population of 3,500 people. c. 'For fish consumption, the average concentration in Lake Ndchigan water, resulting from plant releases, is used with a bioaccumu-lation factor to determine the average concentration in fish. d. Fish do not reside continuously in the discharge canal but migrate. Population doses based apon' drinking water from the Charlevoix municipal system were 0.025 manrem and total Lake Michigan drinking water consumption population dose was 1.1 manrems. The consumption of all of the Lake Michigar. fish harvested resulted in a population dose of 0.18 manrem. As a measure of total environmental impact, the radioactive liquid releases from the plant are averaged over the entire lake and then used to determine the population dose from fish caught throughout the entire lake and total water consumed from the lake. Both of the dose calculations are conservative in that: a. Equilibrium is not obtained in the human body for most isotopes released. b. No credit is taken for precipitation and deposit in sediment or uptake by life forms other than fish. c. No credit is taken for radioactive decay which for I-131 is si nificant. d Results are shown in the following tables. l

• ERG Special Report No 2, " Trace Element Distributions in Lake F3chigan Fish: A Basel.'.ne Study With Calculations of Concentration Factors and Equilibriu= Racioisotope Distributions," March 1973.

1 VII-7

~

x CONSUMERS POWER COMPANY-

~ Big Rock Nuclear Power Plant Calculated Radiation Doses From Liquid Effluents - Population Drinking Water Dose January 1, 197h to December 31,.19Th. Avg Concentration (CI/MPCI). Population. Population Dose Critical Curies in Lake Michigan MPDI Dose Charlevoix, Mich Vector-Isotope MPC Organ _ Releas ed ( Ci/ml) CI/MPCI (mrem /Yr),, (Man-Res) (Man-Rem) Water-Zn-65 1.OE-Oh Whole Body 2 3hE-02 4.88E-15 4.88E-11 2.hhE 2.hE-Oh 5.kE-06' Water I-131 3 0E-07 Thyroid 5 2]E-03 1.09E-15 3.63E-09 1.81E-06 1.8E-02 -k.0E-04 Water Cs-13k '9 0E-06 Whole Body T.5hE-02 1 5TE-lh 1 75E-09 8.T3E-0T 8.TE-03 1 9E-Oh Water Cs-137 2.0E-05 Whole Body 1.8hE-01 3.0hE-lk 1 92E-09 9.61E-07 9 6E-03 2.1E-Oh d Water BaLa-lho 2.0E-05 GI Tract 1.16E-02 2.hlr-15 1.20E-lO 1.81E-07 1.8E-03 k.0E-05 7'

• Vater Co.60 3.0E-05 GI Tract 1.10E-01 2.29Fclk T.62E-10 1.lhE-06 1.1E-02 2.6E-Oh Water Mn-;k 1.0E-04 GI Tract 4.92E-02 1.03E-lh 1.03E-10 1 55E-07 1.6E-03 3.5E-05 Water others 3 0E-0T Whole Body 5.8hE-01 1.22:1-13 k.06E-07 1.10E-OL 1.1E00 2.hE-02 Total - Whole Body 1.lE00 2 5E-02

- Thyroid 1.8E-02 h.OE-Oh- - GI Tract 1 5E-02 3 3E-04 - Bone 0 0 Ave: age concentration in Lake Michigan = curies released / volume of Lake Michigan. Volume of Lake MichiJan 4.8 x 10 ' liters. ,3, eased on a finie intaxe er 1200 m1,ea,. (h) Population taking its drinking water from Lake Michigan is approximately 10,000,000 people with 7,000,000 in the Chicago area. Using average concentration in discharge canal diluted by 800. ($)10 CFR 20 MPC for unknown mixture with certain isotopes not present. . =.

~, CONSUMERS POWER COMPANY Big Rock Nuclear Power Plant Calculated Radiation Doses From Liquid Effluents - Fish Consumption Dose January 1, 197h to December 31, 197h Average Concentration Average (2)Bioaccum-in Lake Concentration (CFI/MPCI) Population Critical ulation Mi:higan in Fish MPDI Dose ( Vector Isotope MPCI organ Factor (pCi/ml) (uCi/g) (mrem /Yr) (Man-Rem)' Fish Zn-65 h.h0E-03 Whole Body 900 h.88E-15 4.39E-12 5.0E-07 3.0E-Oh Fish I-131 1 32E-05 Thyroid 500 1.09E-15 5.hhE-13 2.1E-05 1.2E-02 Fish CS-134 3 96E-Oh Whole Body 2,360 1 57E-lh 3.TlE-ll h.TE-05 2.8E-02' Fish CS-13T 8.80E-Oh Whole Body 2,360 3.8hE-lh 9 07E-11 5 2E-05 3 1E-02 'I Fish BaLa-lho 8.80E-Oh GI Tract 365 2.klE-15 8 79E-13 1 5E-06 8.9E-Oh Fish Co-60 1 32E-03 GI Tract 330 2.29E-lh - T.5hE-12 8.6E-06 5 1E '3 Fish Mn-Sh h.h0E-03 GI Tract 280 1.03E-lh 2.88E-12 9 8E-07 5.8E-04 l Fish Others 1.32E-05 Whole Body 80 1.22E-13 9 76E-12 2.00E-Oh 1.2E-01 Total - Whole Body 3.0E-Oh 1.8E-01 - Thyroid 2.lE-05 1.2E-02 - GI Tract 1.lE-05 6.6E-03 - Bone 0 0 MPCI = MPC (Water)

• 2200/50 ERG Special Report No 2 (3) Population dose based on annual catch of 2 3E+T pounds consumed at 50 gm/ day / person.

s

I VIII. ENVIRONMENTAL MONITORING A. Environmental Survey Environmental levels of radioactivity as found.in the vicinity of the plant were composed almost entirely of naturally occurring radio-active materials. 'n the vicinity of the circulating water discharge canal, radioactive material of plant origin has been found. These mate-rials occurred primarily in aquatic organisms. The levels of radioactive materials, however, were extremely low and are of no significance to the health and safety of the organisms or the public. Purther, the levels of radioactive material found in the resident biological community are consistent with levels found in previous years and show no upward trend. The environmental surveillance program includes continuous sampling of air for particulate and halogen activity at seven locations including background sample locations at Traverse City and Boyne City, Michigan, about 50 miles south-southwest and 40 miles southeast of the plant, respectively, to determine increased concentrations, if any, of radioactivity of plant origin. Thermoluminescent dosimeters (TLD), placed at each of these-locations plus six additional locations on the site. property boundary, measura direct dose in the environment. In addition to the exposure re-l ceived in the field, dosimeters also receive an exposure in transit. To account for this exposure, two control dosimeters accompany the field dosimeters during shipment to and from the environmental contractor and are stored in c lead shield while the other dosimeters are in the field. The average dose received by the control dosimeters is then subtracted from each field dosimeter to obtain a net exposure. The average net doses at the site, inner ring and background stations are compared monthly and any difference, at the 95% confidence level, is determined using standard "F" and "t" tests. The results of these dosimeter analyses are given in Appendix D. While all the dosimeters record-doses from natural occurring sources, the dosimeters on site can also be expected to receive doses from not only the plume but direct radi-ation from the plant. Differenecs between the site and background stations \\ were observed on only two occasions..A difference of 2.1 1 1.8 mR was noted for May and a difference of 1.3 1 0.6 for November. During the same VIII-l

i.

period ~of time, the inner ring dosimeters did not show a dose rate significantly above background though a difference of 1.6 i 1.1 mR between the inner ring and background stations was observed for March. The July TLDs, including controls, received exposures of 200 to 500 mR. Since the only time the control dosimeters are with the field dosimeters is during shipment to and from the lab, it must be assumed that the donimeters were exposed while in transit. The December inner ring and background dosimeters were not read upon receipt by the contractor. The December dosimeters were sent j to the contractor along with the January 1975 dosimeters which were being returned unexposed due to their late arrival at Big Rock Point. The re-turn of two sets of dosimeters resulted in the contractor reading only the December site dosimeters. Air samples gathered continuously and analyzed weekly at the stations shown in Appendix D showed no difference, at the 95% confidence level, in the level of radioactivity measured at those stations close to the site and those remote from the site. Both particulate filters and carbon cartridges are used to measure potential concentration of radio-active materials resulting from plant operations. From the known meteoro-logical dispersion conditions, the following maximum concentrations can be calcu_Tted: Particulates (March) = (1.2 pCi/sec) x (0.003) x (5.0E-lh sec/cc) = 1.8E-16 pCi/cc Halogens (December) = (1.2 pCi/sec) x (0.017) x (5.0E-lh sec/cc) = 1.03E-15 pCi/cc These compare to the minimum detectable activity values and normal background concentrations as follows: Maximum Calculated Minimum Detectable Normal Background 3 Release Concentration pCi/cm Activity pC1/cm3 Activity pCi/cm3 Particulates 1.8E-16 1.0E-lh 7 0E-14 Halogens 1.0E-15 2.0E-14 Hence, the negative data obtained in the program was expected. The method used by the environmental contractor to analyze charcoal cartridges for I-131 is currently being reevaluated. This was prompted by the fact that VIII-2

( approximately 10% of the charcoal cartridge was reported to have positive iodine concentrations. Cartridges are counted using a No 1 crystal connected to a single channel spectrometer adjusted to accept pulses from the 0.36 MeV gamma. For those results showing greater than 0.02 pCi/m3 (approximately 0 5 cpm), the presence of other nuclides increases the counts in the 0.36 MeV channel above background levels. Printouts of these spectra fail to show any peak in the 0.36 MeV region - and, hence, it is questionable whether or not radioiodine was truly de- . tected in a majority' of the cases when levels greater than 0.02 pCi/m were reported. Also, at the Big Rock Point Plant, daily composite condenser circulating water inlet and canal water discharge samples are taken and analyzed for radioactive content. In addition, a monthly composite of these samplee is analyzed for radioactive content (gross beta and tritium). These results are shown in Appendix D. Additional aquatic samples are taken and analyzed during the spring and fall and these results are also tabulated in Appendix D. Based upon the liquid release of 1.07 curies of radioactive mate-rial (less tritium and noble gases) which results in an average concentra-tion in the discharge canal of 1.0hE-08 pCi/ml, the analysis of discharge canal water should indicate an increase of radioactive material in dis-charge canal water samples since the minimum detectable activity for gross beta measurements is about 2.0E-09 pCi/ml or about five times lower than the predicted discharge concentration. The results sho"n plotted in Appendix D indicate an average concentration of about 1.8 1 0.3E-08'pCi/ml above intake concentration which is in relative agreement with the calcu-lated concentration. Also contained in Appendix D are the results of the analysis of the site well water. Two samples were collected in November. The first, collected on November 14, had a gross beta activity of 1.73 1 6.3 pCi/1. A second sample was collected, after notification of the high activity by the contractor lab, on November 27 The second smsple's activity was 2 5 i 1.0 pCi/1. Subsequent investigation failed to pinpoint the exact cause of the high activity found in the November lh sample though VIII-3 ,,n- ~---,

t indications are that the activity was the result of cross-contamination of the sample in the plant laboratory. I 4 VIII-b

IX. OCCUPATIONAL PERSONNEL RADIATION EXPOSURE The twelfth refueling and maintenance shutdown occurred from June 2 through July 27, 1974. This report will cover the period July 1 through July 27, 1974. Personnel exposure for the remaining portion of the report period will appear at the end of this section. The following is a list of personnel exposure by. job breakdown durind a portion of the twelfth refueling shutdown (July 1 - July 27). These exposures were tabulated from pocket dosimeter records, radiation protection logbooks, weekly radiation exposure record work sheets and high radiation area work sheets. The tabulation is considered to be 10% high since the pocket dosimeters normally read 10% higher than the film dosimeters. Summary Total Exposure No of mrem Personnel l Operations 8,756 20 Shift Supervisors 1,178 6 Administration & Supervision 2,155 16 Maintenance Big Rock Plant 12,969 13 Traverse City Region Repair 4,981 4 Grand Rapids, Campbell 5,677 5 Radiation Protection & Chemistry h,722 5 Plant Technicians (I&C) 1,135 h Radiographers 359 2 Total for CP Co Employees h1,932 General Electric 270 2 Southwest Research Institute 7,960 8 Hartford Insurance 158 2 Total for Outside Personnel 8,388 j Grand Total for.This Portion of the Shutdown 50,320 Grand Total for the Complete Shutdown 6/2-7/27/74 96,631 l Pocket Dosimeter Accumulation - [ IX-1 l

. _. _.-_=._. 4 .~ ~ AFP1 F12 I Crestyyps powIn emmy Big Rock Troint Plant. Docket Bo 50-155 Atmospherie Release of Ra*loactive Material Sin-Noeth Unito January Feb ruary March Arr11 May June Tot al 3 5 3.30 x 10 3 95 a 10 153 x 10 Total poble Cases Curies 3.75 m 10 k.k1 x 10 3.43 a 10

• Total Halogens 1.10 m 10-2 1.55 a 10-2 6.87 x 10'#

2.k3 x 10-2 1.65 x 10'# 7.37 a 10 0.21 3 1.87 m 10-2 4 1.92 x 10-3 1.16 x if Total Particulates (e.i) 8.39 x 10 1.93 a 10-3 1.06 a 10-2 2.26 a 10 Total Tritium 4.96 6.39 5.20 0.61 6.13 1.ok 2k.3 4 2.02 x 10-7 1.13 x 10*I 3 05 10'I k.01 x 10-7 1.25 x 10 Total Partleulate - Gross Alpha 9.01 x 10' 1.k0 m 10 2 Maaksian poble Gas Belease Rate uC1/s 1.13 x 10 1.89 x 10 1.76 m 10' 3 21 x 10 1.70 a 10 3.21 x 10 Tercent of Tech Syt Limits for 5 1.23 0.15 0.98 soble Games 1.40 1.82 1.28

• nalogens 0.27 0.ko 2.00 0.78 0.k1 2.32 1.03 Particsistes 0.01 0.03 0.23 0.04 0.02 0.02 0.06 4

Isotopes Released Curies Halogene 2 1

• I-131 7 95 x 10'3 1.01 a 10-2 6.26 x if 2.k2 x 10'#

1.23 a 10-2 7.18 x 10'# 0.189 4 3 3 1.93 x 10-3 2.08 a 10 l 1-133 3.08 a 10'3 5.38 x 10'3 6.1h if 1.19 x 10 k.18 x a f Particulates i Co.134 2.86a10j 3.30 x 10' 8.71 x 10j 6.79 x 10'f 1.17a10'[ 1.50 x 10-3 Co-137 6.Okn10]' 9.25 x 10'g 2.37 x 10'3 3 99 s Qg 9 39 a 10'4 7.65 x 10'S 3.65 s 10'3 1.06 a lo 1.92 x 10 1.k2 a 10' 3.27 a 10 Bala-1ko 2.92 x 10 1 33 x 10' 3.96 x 10 Mn-54 1.95 x 1 7.35m10*g 2.98x10j 2 98 s 10-5 3 j Zn-65 6.13 a 10 {2 58 a 10'3 1.30 x 10 3.50 x lo-Co-60 1.69 x 1 1.k3 a 10*g 2.k2 s 1 5.53 a 10,g 2.13 x 10-j 4 1.93 a 10'3 1 Net UniSectified Beta 2.98 x 1 6.54 x 10'g 6.57 x 10-3 1.10 m 10-3 5.10 a 10 2.81 x 10 9.42 m if 3 3 2 6.48 x 10 8.01 a 10 3.94 a 10 Noble cases 1.06 a 10 1.21 a 10 9.ko x 10 3 3 3 3 2 5.75 x 10 6.22 x 10 2.19 a 10 No-138 k.77 x 10 6.03 a 10 k.77 x 10 3 3 3 3 Kr-87 2.87 x 10 3.30 x 10 2.60 x 10 3.13 a 10 3.43 s 10 1.23 a 10 4 3 3 I 2 Kr-88 1.79 a 10 2.06 a 10 1.5d a 10-1.87 a 10 1.94 x 10 7.k9 m 10"3 3 3 3 3 Kr-85e 5.97 x 10 6.87 x 10 5.32 s 10 6.71 x 10 6.57 x 10 2.55 s lo 3 3 3 3 2 g Ie-135 2.0k a 10 2.8h a 10 2.12 x 10 k.33 a 10 3.06 x 10 1.16 a 10 Ie-133 0 0 0 0 0 l Ie-1k) 0 0 0 0 0 0 0 Kr-94 0 0 0 Kr-93 0 0 0 0 0 1e-141 0 0 0 0 0 i Kr-92 0 0 0 0 0 el el 6.19 Kr-91 1 38 1.59 1.22 Ie-Iko 18.6 21'S 10* 5.57 2.03 64.2 2 2 2 2 Kr-90 2.1h a 10 ' 2.kT s 10 1.89 a 10 66.6 23.3 7.40 a 10 2 3 Ie-lk1 3.18 x 10 3.66 x 10 2.81 x 10 1.00 x 10 3b.5 1.10 a 10 l 3 3 5.18 a 10 1.56 a 10 5.03 x 10 Kr-89 1.bb x 10 1.65 a 10 1.27 x 10 Ie-137 3.30 x 10 3.81 a 10 2.93 x 10 1.22 x 3 0 3.59 s 10 1.16 a 10' 3 3 3 3 2 Ie-135m 3.05 x 10 3 52 x 10 2.70 x 10 1.k2 a 10 3.12 a 10 1.10 m 10 s 3 2 2 2 3 Kr-83m 1.06 x 10 1.22 x 10 9.40 m 10 8.26 x 10 1.15 a 10 4.16 a 10 Ie-133m 85.5 98.6 75 7 2.55 x 10 9.30 5.2h a 10 Ie-131m 3.52 k.06 3 12 h4.4

• 1

$6.1 Kr-85 2.79 3.22 2.kT 3.16 x 10

• 1 3 25 x 10 2

3 N-13 2 38 a 10 3.32 a 10 2.63 x 10 3.50 x 10 22.3 1.21 a 10 I b 'W assred value Multiplied by Three l

AFFENDIX B C0KOUMERS POVER COMPANY Big Rock Point Plant. Docket No 50-155 Radioactive Liquid Release Six-Month Units January February March April May June Total Total Radioactivity Released (Except Tritium. Disst:1ved Cases and Alpha) Curies 9.8'x 10-2 2 50 x 10~ 1.k6 x 10 9 96'x 10-2 1.66 x 10 T.23 x 10-2 0.606 Voltane of Waste Discharge Liters k.06 x 10 1.18 x 10 5.95 x 10 5.k5 x 10 8.11 x 10 k.22 x 10 2.90 x 105 Average Concentration of -3 p,33,go-3 2.k5 x 10~3 1.83 x 10~3 2.05 x 10~3 1.71 x 10~3 2.09 x 10-3 Waste Prior to Discharge uC1/ml 2.k2 x 10 Volume of Circulating 9 9 9 9-Discharge Water Liters 8.79 x 10 T.94 x 10 8.79 x 10 8.50 x 10 8.62 x 109 '8.50 x 109 1U 511 x 10 Averese Concentratios Released (Except Tritium. Dissolved Gases and Alpha) pCi/mi 1.12 x 10~ 3.15 x 10~9 ~ 1.66 x 10~ 1.17 x 10~0 1 93 x 10 8.51 x 10-9 1.19 x 10 ~0 Maximum Concentration (Except Tritium, Dissolved Genes and Alpha)a pC1/a1 2.73 x lo-T 3,3g, 2n-T 3.67 x 10-T 4.96 x lo-T g,35,in-T 3.31 x 10-7 k.96 x 10-7 Percent of Applicable Limits 5 1.95 0.60 3 28 2.38 4.62 1.55 2.ko Trittum Released Curies 8.53 x 10~ 2.k8 x 10-2 1.25 x 10~1 .1.1% x 10~1 1.k6 8.75 x 10-2 3,99 i Averare Tritium Ccncentration -8 Relea.v.d pCi/ml 9.T x 10-9 3.1 x 10~9 1.42 x 10 1.3h x 10-0 1.69 x 104 1.03 x 10~3 3.72 x 10 Total tross Alpha Released Curies 5 38 x 10-6 3.51 x 10 6.09 x 10 9.80 x lo 3.k2 x 10~k 9.59 x 10 1.38 x 10-3 -5 d Avere,pr Alpha Concentration sci /mi 6.12 x 10-12 k.k2 x 10-13 6.93 x 10 1.15 x 10 3.97 x 10~11 1.13 x 10-10 2.70 x 10-11 42 42 Isotwes Curies I."31 1 98 x 10-3 1.25 x 10~3 2.Tk x 10~3 2.27 x 10-3 k.65 x 10~2 3 Cr.13h 1.27 x 10~ 1.93 x 10 ~ 1.26 x 10-2 3 -Co-137 2 57 x 10~3 6.98 x 10-3 3.02 x 10-2 6.TT x 10 2 1.39 x 10-2 T.37 x 10 3 5.52 x 10- -3 175 x 10 2 2.55 x 10-2 7,73,39 2 3,3g,194 Co-60 9 15 x 10-2.01 x 10 1.73 x 10-2 1.07 x 10 7.28 x 10-3 1.29 x 10-59kxloi Bala-lho ~3 In-65 1.82 x 10~3 8.15 x 10 2.63 x 10 ~ Mn 54 k.k1 x 10~2 T.53 x 10~k 5.16 x 10~3 1.03 x 10 j -2 Fe-59 1.24 x 10-k.58 x 10-3 7.12 x 10~3 2.k1 x 10 3 -2 3.43 x 10-3 1.63 x 10~ 5 06 x 10-3 Total Identified Released Radioactivity Curies 4.95 x 10-2 1.10 x 10 6.01 x 10-2 5.22 x 10-2 5.54 x 10-2 4.k3 x 10-2 3 03 x 10~1 -2 Percent of Total Identified 50.!. kk.2 kl.2 52.k 33.k 61.2 50.0 i s ~

APPENDIX C Off-3ite ShiInent of Radioactive Material Shipnent Transfer Number Date From Transfer To Radioactive Material Disposition 353 h/ 2/74 DPR-6 Oyster Creek Plant Fuel HandiIng and Gamma Scan Equip-Reuse 29-12773-01 ment - 5 mci 35h h/18/Th DPR-6 NECo, 16-NSF-1 152 Gallon Barrels (Filters) Burial Sheffield, IL Low-Level Wastes - 349 mci i' 355 h/24/7: DPR-6 NECo, 16-NFS-1 15 Gallon Barrels (Filters) Burial Morehead, KY High-Level Wastes - 7.38 Ci 356 h/29/Th DPR-6 NAC Battelle Fuel Bundle Inspection Device Reuse Columbus Lab 3 mci-34-06854-05 35T h/30/Th DPR-6 NECo, 16-NSF-1 4' x 6' Steel Tank, Misc Waste - Burial Morehead, KY 9 78 Ci 358 h/30/Th DPR-6 NPI c/o Battelle 20 Irradiated Cobalt Rods - Storage Columbus Lab 311,300 Ci 3h-06854-05 359 5/ 1/7h DPR-6 Exxon Nuclear Fuel Inspection Tools - 2 mci Reuse Richland, WA WN-1062-1 360 5/16/74 DPR-6 NECo, 16-NSF-1 12 -- 59-Gallon Barrels, High-Level Burial Morehead, KY Waste (Filters) - 77 CI 361 6/1T/74 DPR-6 GE, Vallecitos, CA 6 Irradiated Fuel Rods - 29,942 Ci Inspection I SNM-960, Amend 70-20 & Analysis V

APPENDIX C Off-Site Shipsent of Radioactive Material Shipment Transfer No Date Frm _ Transfer To Radioactive Material Disposition 353 h/2/Th DPR-6 Oyster Creek Plant Fuel handling and gansaa scan equipenent Re-use 29-12773-01 5 mci 35h h/18/74 DPR-6 NECo, 16-NSF-1 152 55-gallon barrels (filters) Burial Sheffield, Ill. Low level vastes (3h9 mci) 355 h/2h/74 DPR-6 NECo, 16-NFS-1 15 55-gallon barrels (filters) Burial Morehead, Ky. High level vastes (7 38 Ci) 356 h/29/74 DPR-6 N.A.C. Battelle Fuel bundle inspection device Re-use Columbus Lab 3 mci 3h-06854-05 357 L/30/Th DPR-6 NECo, 16-NSF-1 k' x 6' steel tank, mise. vaste Burial Morehead, Ky. 9.78 Ci 358 h/30/Th DPR-6 NPI c/o Battelle 20 irradiated cobalt rods Storage Columbus Lab 311,300 Ci 34-06854-05 359 5/1/74 DPR-6 Exxon Nuclear Fuel inspection tools Re-use Richland, Wash. 2 mci WN-IO62-1 360 5/16/74 DPR-6 NECo, 16-NSF-1 12 55-gallon barrels, high level vaste Burial Morehead, Ky. (filters) 77 Ci 361 6/17/74 DPR-6 G.E.,Vallecitos, Cal. 6 irradiated fuel rods Inspection SNM-960, Amend.70-20 29,942 Ci & Analysis

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..( APPENDIX D -Difference in Average TLD RenMngs mR/ month Month Site vs Backgrcind Inner Ring vs Background Stations Stations January (1}N. D. N. D. February N. D. N. D. March N. D. 1.6 I 1.1 April' N. D. N. D. .May 2.1 1 1.8 N. D. June N. D. N. D. Average 0 35 1 0.3o 0.27 1 0.18 (1)N. D. - No difference at the 9% confidence level i .- l

Appendix D Aquatic Biota Gross Gesuna Identification epm /g Discharge Algae 1.94 9 Mile Pt Algae 0.09 9 Mile Pt Periphyton 0.13 d Mile So Discharge Periphyton 0.15 h Mile No Discharge Periphyton 0.78 Mt MSSuaba Algae 0.19 Discharge Crayfish 0.28 h Mile So Discharge Crayfish 0.05 h Mile No Discharge Crayfish 0.05 Mt McSuaba Crayfish 0.02 9 Mile Pt Crayfish 0.03 Fish Flesh Suckers Mt McSuaba 0.03 Fish Flesh Lake Trout o.03 Fish Flesh Sucker Lake 0.02 9 Mile Pt Fish 0.03

APPENDIX D (Contd) High, Low and Average Concentrations For Highest Average Sampling Location January 1,1974 to June 30, 1974 Type Type of Analysis Units Location High Low Average PCi/h TC o.26 (.01 0.11 Air Gross Beta-Gamma 3 I-131 PCi/m All o.11 (0.2 (0.2 Lake Water Gross Beta PCi/l BR ST LWO' 26.5 2.1 11.4 H-3 pCi/1 BR ST LWo 1316 299 587 Well Water Gross Beta BR ST W 15.7 40.28 6.7

• rLD Dose mR/mo E

6.5 6.4 6.4

• In excess of control dosimeter.

APIENDIX D (Contd) Samplir4 and Analysis Suecary January 1, 197h to June 30, 1974 Number of Samples Frequency of Medium Description Location Collected Type of Analysis Analysis Air Continuous at All 170 Gross Beta, 131 Weekly 1 Approximately 1 cfm Lake Water 1 Gal Grab ST 12 Gross Beta, Gross Gansna Nnthly Sr, 134,,65,Cs, 5#m, Quarterly 137 W g 58Co, 60Co, 3, 59Fe Well Water 1 Gal Grab ST 6 Gross Beta Monthly Gamma Dose Continuous All 60 TLD Dose Monthly Aquatic Biota Grab ST, UM, Gross Beta, Gross Gesuna Semiannual Mt McSauba Spectrum

OCCUPATIONAL EXPOSURE l 7 2.2.1.2.a(1)(h) Number of Persons Within Exposure Range 'arem Dose 7/1/Th-7/31/Th 8/1/74-8/31/Th 9/1/7h-9/30/74 0-100

• Maint 3

Oper-2

• Maint 9

Oper 13

• Maint 7

oper 14' Supy 17 Tech 2 Supv 18 Tech 7 Supy 21 Tech 7 Others 36 others lh others 15 101-500

• Maint 1

Oper 12 'Maint 4 Oper 7

• Maint 8

Oper 6 Supy 3 Tech 5 Supy 4 Tech 3 Supv 1 Tech 3 others 7 Others 0 Others 2 501-1250

• Maint 3

oper 5

• Maint 0

Oper o

• Maint 2

Oper o Supv 2 Tech 3 Supy 0 Tech 0 Supy 0 Tech 0 Others 3 others 0 Others 0 1251-2500

• Maint 9

oper 1

• Maint 0

oper o

• Maint 0

Oper o Supy 0 Tech 0 Supy 0 Tech 0 Supy 0 Tech 0 Others 6 Others 0 Others 0 >2500

• Maint 0

Oper 0

• Maint 0

Oper o

• Maint 0

Oper 0 Supv 0 Tech 0 Supy 0 Tech 0 Supy 0 Tech 0 Others 0 Others 0 Others 0-Total' Number of People Badged 120 79 86 mrem Dose 10/1/Th-10/31/Th 11/1/Th-11/30/Th 12/1/74-12/31/Th 0-100

• Maint 3

oper 12

• Maint 5

oper 18

• Maint 7

oper 19 Supv 15 Tech 7 Supy 22 Tech 7 Supv 21 Tech 8 Others 32 others 3h others 26 101-500

• Maint 9

Oper 8

• Maint 8

oper 2

• Maint 7

Oper 1 Supy 7 Tech 2 Supy 0 Tech 2 Supv 1 Tech 2 Others 0 Others 2 Others 0 501-1250

• Maint 0

Oper 0

• Maint 0

Oper 0

• Maint O.

Oper 0 Supy 0 Tech 1 Supv 0 Tech 1 Supv 0 Tech 0 Others 0 Others 0 Others 0 1250-2500

• Maint 0

Oper 0

• Maint 1

Oper 0

• Maint 0

Oper 0 Supy 0 Tech 0 Supy 0 Tech 0 Supy 0 Tech 0 Others 0 Others 0 Others 0 >2500

• Maint 0

Oper 0

• Maint 0

Oper 0

• Maint 0

Oper 0 Supy 0 Tech 0 Supy 0 Tech 0 Supy 0 Tech 0 Others 0 Others 0 Others 0 Toal Number of People Badged 96 102 92 28 people received an annual exposure greater than 2500 mrem because of the major causes stated in this section (IX). I -Others include office secretaries, general office personnel, contract personnel, vendor s personnel, plant guards, Region repairmen (other than from Traverse City) and visitors.

• Maint includes Region repairmen from Traverse City.

IX-2 j

.l. Exposure - Job Break.iovn (July 1-July 27, 1974) I Total Exposure No cf mrem Men Radwaste or Fuel Pool Filter Change 117 h Recire Pump Room To Check for Leaks 748 3 Valve Lineup in RCP Room 89 3 Shutdown Heat Ex Room Valve Lineup 61 3 Routine Refueling 2,103 18 Inspection in Steam Drum Area 20 1 Reactor Head Installation 4,750 12 Sipping Fuel 204 h Off-Site Cask - Treat 2 30 2 Reconstituting, Core Hardware 1,h89 8~ Recirc Pump Seal Inspection 261 1 Baffle Plate Inspection 315 2 Monitoring for Inservice Weld Inspection 231 3 Changing Rod Drives 2,09h 5 Removal of Rollers From Blades - Monitoring 175 1 In-Cores - Monitoring 70 1 .l Changing Rod Drives - Monitoring-h31 3 . Baling Solid Waste - Monitoring 40 1 Recire Pump Room - Monitoring 813 5 Steam Drum Area - Monitoring 318 3 Working on Start-Up Channels 100 1 CU Demin Pit - Monitoring 165 3 CU Demin Pit - Valve Calibration 100 1 Control Rod Drive Room - Monitoring 109 3 Reactor Head Removal - Monitoring 75 1 Control Rod Drive Room Work on In-Cores 1,142 7 Cleaning Welds for Inservice Inspection 789 3 Steam Drum Enclosure To Check Valve Packing 94 2 Removal of Rollers From Blades 904 h Fuel Pit Filter Area Work on Steam Pop Valve 175 3 f . Cleaned Up Demin' Resin Line (Plugged Line) 888 2 IX-3 ~

I. Exposure - Job Breakdown (July 1-July 27,1974) (Contd) } Total Exposure No of j, mrem Men Work on Baffle Plate 753 1 Remove Insulation in Recire Pump Room 489 4 -Recire Pump Room Work on Valves 1,989 6 Head Removal and Installation 1,182 5 CU Pit Work on Valve 1 h71 h Hydro Leak Repair RCP Room & CU Pit 88 2 CRD Room Flux Wire Tubes 147 2 RCP Room Installing Insulation and Work on CV Valve 529 5 RCP Room Insulat'ng and Cleanup 3,179 h Steam Drum Ins sting 2,173 h SWRI Inservice 4eid Inspection 7,960 8 Hartford Ins 6 2 GE Fuel Inspection 270 2 j CP Co Radiographers 359 2 All remaining exposure was assigned to: Routine Plant Surveillance and Inspection and Routine Work for Entire Plant Staff 10,881 i The following is a tabulation of exposure - job breakdown for the' entire twelfth refueling and maintenance shutdown. Exposure - Job Breakdown Total Exposure No of mrem Personnel Operations l Main Condenser Area To Check for Steam 62 2 Routine Refueling 8,650 20 Unloading Core Hardware 2,917 10 Sipping 3,605 10 Reconstituting Core Hardware 1,489 8 Radwaste or Fuel Pool Filter Change 322 6 Tagging in Recire Pump Room & Valve Lineup 170 5 Off-Site Cask - Treat 2 (M 30 2 Shutdown Heat Exchanger Room - Valve Lineup 61 5 1 Recire Pump Room - Checking for Leaks 350 1 Routine Plant Surveillance & Inspection 10,233 20 Total 27,81h

,w i ' Exposure - Job Breakdown (July 1-July 27,1974) (Contd) Total Exposure No of mrem Personnel 4 Supervision Main Condenser Area To Check for Steam Leak 92 l' Routine Refueling 572 5 Recire Pump Room Insulation Removal 198 2 Recire Pump Room - Replace Seal in #2 Pump 638 1 Reactor Vessel - Inspection of Baffle Plate 334 1 Reactor Head' Installation 330 2 Steam Drum Area Inspection 20 1 Recire Pump Room - Checking for Leaks 398 2 Routine Plant Surveillance & Inspection 5,160' __ 23 Total 7,742 Radiation Protection Technicians Monitoring Reactor Head Removal (Twice) 175 2 Monitoring in hhh 18 1 Monitoring Main Condenser To Check for Leak 20 1 Monitoring in CRD Room 820 1 Monitoring in Recire Pump Room 1,397 h Monitoring at Solid Radwaste Area 30 1 i Monitoring on Reactor Deck - Baffle Plate Inspection 287 2 Monitoring in Clean-Up Demin Pit 195 3 Monitoring in Clean-Up Heat Exchanger Room 9 1 Monitoring in Slutdown Heat Exchanger Room 10 1 Monitoring in Steam Drum Enclosure 318 3 Monitoring in Removal of Rollers From Blades 175 1 Monitoring Baling 40 1 Monitoring Reactor Head Installation (Twice) 280 2 Monitoring Tank Room - To Remove Iodine Sampler 90 1 l Routine Work (Includes all Chem Lab and Routine Rad' Protection Work) 3,139 5 { Total 7,003 i IX-5 l

Exposure - Job Breakdown-(July 1-July 27, 1974) (Contd)_ Total Exposure No of arem Personnel Instrument Technicians Removed Switch Covers in Recire Pump Room 45 1 Repaired Thermocouple in,CRD Room 45 1 Disconnected In-Core Cables in CRD Room 127 2 Work on Start-Up Channels 100. 1 Calibration of Valve in CU Pit 100 1 Work on #8 In-Core CRD Room 30 1 Routine Work 1,378 4 Total 1,825 Maintenance Reactor Head Removal (Twice) 2,031 8 Cleaning in hhh 54 2 Repair of Rod Drives in RD Access 1,020 h CRD Room Removing In-Cores 138 2 Work on Valve To Clean Sump RCP Room 40 1 Deposited Solid Radwaste in Vaults 182 3 Work on T-2 Cask - Rods 3kh 3 Deck Work for Exxon 167 1 Replace #2 RCP Seal 1,h09 3 RCP Room Insulation Removal 1,9h9 5 Changed Location of Electric Box in RCP Room 365 2 Clean-Up Heat Exchanger Room Insulation Removal 35 2 Shutdown Heat Exchanger Room Repack Valves 26 1 Steam Drum Enclosure To Check Packing 94 2 Removal of Rollers From Blades 904 4 Puel Pit Filter Area - Worked on Pop Valve 175 3 Clean Demin Resin Line (Plugged Line) 880 2 Steam Drum - Remove Insulation 671: 3 Changing Rod Drives 2,094 h { Replacing In-Cores 772 3 RCP Room - Work on Valves 764 2 1 IX-6 4

f ^ f(. Exposure - Job Breakdown-(July 1-July 27, 1974) (Contd)- ~ Total Exposure-No of mrem Personnel ~ Maintenance (Contd) .CRD Room - Flux Wire Tubes 147 2 RCP Room - Installing Insulation 529 5 CU Demin Pit - Work on Valve 571 3 Reactor Head Installation ('n ice) 2,364 9 w .CRD. Room #8, In-Core 340 3~ 'RCP Room - Valve Packing 330 2 Hydro Leak Repair RCP Room & CU Pit 88 2 Routine Work h,784 _ 13 Total 23,270 All Other CP Co Maintenance -Baffle Plate Work 753 1 Reactor Head Installation. 1,596 4 RCP Room Insulating & Cleanup 3,179 4 Steam Drum Insulating 2,173 h Reactor-Head Removal 897 h-RCP Room Packing Reactor Level Sensor Valve 895 2 CU Demin Pit - Work on Valves 900 2 Routine Work 265 9 Total 10,658 j Outside Personnel General Electric 1,390 2 Exxon Nuclear 6,h67 5 Battelle 1,287 1 SWRI 7,960 8 SAI 70 1 Newkirk 3h4 3 Hartford Ins 158 2 Plant Guests 28h 2 Total 17,960 i 1 l IX-7

1 k. 1 --( ~ Exposure - Job Breakdown l (July 28-December 31,1974) Total Exposure arem ' Operations Routine Plant Surveillance & Inspection 7,953 Radwaste~(Filters,Etc) 151 -Pipe Tunnel Steam Leak Inspection and/or Repair 1,459 e Recire Pump Steam Leak Inspection 240 Cleanup After Refueling Operation 599 Moving Fuel in Fuel Pool 650 ' Decontaminated Reactor Level 240 NFS Cask 25 Total 11.317 Maintenance Personnel Routine Plant Surveillance & Inspection 6,863 Solid Radwaste 200 Repair Recire Pump Seal 276 Resin Transfer (Tank to Tank) 115-Lifting Beam (CU Pump) 133 Main Steam Bypacs Valve 300 ^ Radwaste Pumps (Repair) 64 CRD Pump Packing 26 Reactor Cooling Water Monitar 5 Pipe Tunnel (Steam Leak Repair) 9h6 Pipe Tunnel (Extraction Line Repair) 312 Recire Pump Room (Sensing Line Repair) 600 Shutdown Heat Exchanger 107 NFS Cask 506 Steam Drum Pop Valves 45 Steam Drum Area (Steam Leak) 1,500 Decontaminated CRD Room Access 728 (- Total 12,726 l I(< w % ' IX-8 i I

i l l

(

Exposure - Job Breakdown (July 26-December 31, 1974) (Contd) Total Exposure mrem Chemical and Radiation Protection Techs Routine Plant Surveillance & Inspection 3,734 Pipe Tunnel - Steam Leak Inspection 1,777 Recire Pump Room - Steam Leak Inspection' 340 Resin Transfer (Tank to Tank) 280 Main Steam Bypass' Valve 304 CU Demin Heater Exchange Room 23 Shutdown Heat Exchanger 33 -Recirc Pump Room - Steam Leak Inspection / Monitored Maintenance 862 Steam Drum Area (Steam Leak) 30 Steam Drum (Valve Packing) 770 Total 8,153 Instrument and Control Technicians Routine Plant Surveillance 839 Pipe Tunnel Dew Cell 20 Total 859 Supervision Routine 5,161 Pipe Tunnel - Steam Leak Inspection 2,479 Recire Pump Room Steam Leak Inspection 610 Resin Transfer (Tank to Tank) 222 Refueling (Carry-Over Work) 596 Steam Drum Area (Leak Inspection) 30 Reactor Level Work 15 Total 9,113 1 IX-9 r ,en-

u .( Exposure - Job Breakdown- '(July.28-December 31, 1974) (Contd) Total Exposure mrem Off-Site Personnel Repair Plant Heating Boiler Sh Resin Transfer (Tank to Tank) (Traverse City Region Repairmen) 2,88h f McDermit Co 80 General Office 58 Suntac k0 General Electric-357 Plant Security 703 j AEC Inspections 25 Total h,201 Total Pocket Dosimeter Accumulation Twelfth Refueling Total 7/1-7/27 (arem) 7/28-12/31 (mrem) mrem Operations 8,756 11,317 20,073 Maintenance (BRP) 12,969 12,726 25,695 Traverse City Region Repair h,981 2,88h 7,865 Campbell Repairmen 5,677 0 5,677 Radiographers 359 0 359 Chem & Rad Protection Techs h,722 8.153 12,875 ' I&C Technicians 1,135 859 1,994 Supervision 3,333 9.113 12,h46 Off-Site Personnel (Exclude Traverse City Repairmen) 8.388 1.317 9.705 Total 50,320 h6,369 96,689 [ IX-10 L--_-________________________.

i i i s X. . RADIOACTIVE LEVELS IN PRINCIPAL FLUID SYSTEMS A. Primary Coolant Minimum Average Maximum Reactor Water Filtrate pCi/m1("} 2 9 x 10 2 9 x 10 9 5 x 10~ ~ Reactor Water Crud -1 pCi/ml/ Turbidity Unit 1 7 x 10 5.8 x 10-1.5 x 10 ~ Iodine Activity -2 pCi/ml h.0 x 10 1.0 x 10-2.0 x 10 B. Reactor Cooling Water System Reactor Cooling Water pCi/m1(*) 1 5 x 10 8.7 x 10 5.8 x 10 -3 -3 -2 C. Spent Fuel Pool Fuel Storage Pool (" 5 0 x 10 2.0 x 10 2.0 x 10 -I ~ ~ -3 -3 -2 Fuel Pool Iodine 2 9 x 10 5.8 x 10 1.0 x 10 4 " A counter-efficiency based on a decay scheme consisting of one gamma photon per disintegration at 0.662 MeV used to convert count rate to microcuries. All count rates were taken two hours after sampling. Based on efficiency of Iodine-131 two hours after sampling.

1. Based on APHA turbidity units and 500 ml of filter sample.

X-1

APPENDIX D Differences in Average TLD Readings mR/ Month Site Vs Background Inner Ring Vs Background Month Stations Stations January ND ND February ND ND March ND 1.6 i 1.1 April ND ND May 2.1 1 1.8 ND June ND ND Jdy(2) August ND ND September ND ND October ND ND November 1.3 0.6 ND December Average 4 ND: N0 difference at the 95% confidence level. July TLDs were exposed in transit. (3)Due to' late arrival of the January 1975 dosimeters, these dosimeters were returned to the contractor, unexposed with the December dosimeters. This resulted in an additional mix-up at the contractor's lab and resulted in the December inner ring and background dosimeters not being read.

_J: APPENDIX D Aquatic Biota Grot,s Beta (pCi/g) Gross Gamma (cpm /g) Location-Spring 1974 Fall 1974 Spring 1974 Fall 1974 Crayfish Discharge k.9 1 0.5 2.5 1 0.3 0.28 0.12 - 1/4 Mi South 1 7 i 0.2 3.5 0.4 0.05 0.04 - 1/4 Mi North 1.1 1 0.1 2 5 1 0.3 0.05 0.06 - Mt McSauba 1.2 0.1 1.9 1 0.2 0.02 0.02 - Nini Mile Pt 0.8 i 0.1 2.8 1 0.3 0.03 0.03 Algae - Discharge 8.7 1 0 9 NS 1 94 NS - 1/h Mi South NS(I} NS NS NS - 1/4 Mi North NS NS NS NS - Mt McSauba 5 0 1 0.5 6.9 07 0.19 0.21 - Nine Mile Pt 2.3 09 NS 0.09 NS Periphyton - Discharge NS 13.3 13 3 NS 1.55 - 1/4 Mi South 1.3 1 0.1 60.8 6.1(4) 0.15 0.38 - 1/4 Mi. North 5 0 1 0.5 9.8 1 1.0 0 78 0 73 - Mt McSauba NS 2.6 1 0.3 NS 0.37 - Nine Mile Pt h.8 1 0 5 17.h 1 7(5) 0.13 0.25 2) Fish - Lake Trout, Flesh Only 2 7 i O.3 1.6 0.2 0.03' O.01 -Suckers,FleshOnly(2) i 2.3 1 0.3 4.0 0.4 0.02 0.03 - Shore Fish - Nine Mi Pt 2.2 1 0.2 NS 0.03 NS - Suckers, Flesh Only, Mt McSauba 2.2 1 0.2 NS 0.03 NS - Fish, Flesh, Assorted (2) NS 2.1 0.2 NS 0.0h - Shore Fish, Discharge N3 1.h i O.1 NS 0.06 2) - Lake Trout, Flesh Only NS 0.8 0.1 NS 0.01 (1)NS: No sample collected.

2) Samples collt.cted approximately 0 5 miles off discharge at a depth of approximately 20 feet.

(3)Cs-13T,Zr-95,Co-60,K40. (4 Cs-137 [ (5)Cs-137,Zr-95,K-ho. ,e,

m. -APPENDIX D (Contd) ~ High, Lov and Average Concentrations for Highest Averve Sampling Location January 1,-1974 to June 30, 1974 Type -Type of Analysis Units Location M-Low Average 3 Air-Gross Beta-Gamma pCi/m SL 0.24-0.01 0.089 3 I-131 pCi/m All 0.11- <0.2 <0.2 Lake Water _' -Gross Beta-pCi/1 BR ST LWO 16 9 2.1 1.8 H-3 pC1/1. BR ST LWO 1316 250 503 Well Water Gross Beta BR ST WW 15 7 <0.28 4.7 ' CTLD Done mR/mo l

1. Values were not computed for TLDs due to the "in transit" exposure received by

. the July dosimeters. 4 i t,

J4 e APPENDIX D (Contd) Sampling and Analysis Sumary January 1,197h to June. 30,197h ' Number of Samples Frequency of' Medium Description Location Collected Type of Analysis Analysis Air Continuous at All 352 Gross Beta, 131 Weekly i 7 Approximately 1 cfm 4 Lake Water 1 Gal Grab ST, CH 33 Gross Beta, Gross Gamma Monthly ST 2h 90 134Cs, 137Cs,- Sh ~Ma,- Quarterly ~ Sr,. i 58Co, 60Co, 65Zn,. 59 Fe Well Water 1 Gal Grab ST 12 Gross Beta ' Monthly-Gamma Dose Continuous All 145 TLD Dose. Monthly ~ Aquatic Biota Grab ST, NM, 31 Gross Beta, Gross Gamma Semiannually Mt McSauba Spectrum i i f 1

3 S APPENDII A. TABLE 1 CON'AMERS 19fER COMPANY l Big Rock Point Plant, Docket Bo 50-155 .Atmospherie Release of Radioactive hterial I Six-Mmth " Units Jgnuary. February Moreb ' April Mar-June Total (Total Noble Cases. Curles-2 30E+0k -5 30E+0'4 k.13E+0h k.28E+0k k.70E+03 1.65E+05 h Total Nalogens T.6kE 1.1TE-02 3 60bO2 1.09E-02 1.10E-02 3.k5E-02 1.12E-01 - Total Particulates (8, y)- 1.125-03 2.575-03 1.k1E :3 02E-03 2 56E-03 1 55E-03 2.k9E-02 I Total Tritium. 2.89E00 3 90EDO. 3.0TE00 1 57 bo2 3.8kE00 3 2kE00 1 76E+01 ~I Total Particulates-Grose 1.865-07 2 70E-07 1 51bo7 k.0TE-07 5 35E-07 1.6TE-06 Alpha 1.205-07 ' Maziar.se poble Ons Release k.19E+04 2 51E+0h b.19E+0h I pate

pct /s 1 5kt+0h 2.6TE+0h 2 72E+0h Percent of Tech Spee '

l. Liette for 5 {. Noble Cases 8 59E-01 2.19E00 1.5kE00 ' 1.60 1.81E-01 1.06 ~ Halogens 1,L25.2.1TE-01 ~ 9 298-01 3.kTE-01 2.13E-01 1.05E00. h.8kE-01 Partteulates 1.65E-02 3 81E-02' 3.02E-01 5.95E-02 3 03E-02, 2.26E-02 7.92E-02 i Isotopes Released. Curies Ba.logens 1-131 3 53E-03 k.51E-03 2.78E-02 1.0TE-02 5.k5E-03 3 19E-02 8.kOE-02 I I-133 4.10E-03 7.1SE-03 8.18E-03 1 59E-Oh 5.58E-03 2 58E-03 2 78E-02 I - Particulates 3 81E-05 .k.39E-05 1.16E-05 9.0$E-05 1 56E-05 2.00E-Ok Cs-134 Cs-137' 8.05E-05 1.k1E 3 16E-03 5 31E-Ok 2.56E-04 1.89E-Ok k.36E-03 BaLa-1ko 3 89E Ok 1.23E-03 1.83E-03 5 29E-05 1.25E-03 1.02E-Oh k.86E-03 jg En-65 3 97E-05 3 97E-05 ' h 54 2.60E-05 9.80E-05 1.TLE-04 k.66E-05' 3.kkE-04 co-60 2.25E-04 1.90E-04 3 23E-Ok. 7.37E-04 2.8kE-Ok '8.1TE-Oh 2 58E-03 ' set Unidentified Beta 3.98E-04 8.72E-Ok 8.76E-03 1.kTE-03 6.80E-Oh 3 75E-Ok' 1.26E-02 i Noble Cases 'I Xe-138 T.k6E+03 1.6TE+0h 1 30E+04 8 9hF+03 1.10E+03 k'.72E+0k Er-87 3 39E+03 8.32E+03 6 58E+03 7 93E+03 8 55E+02 ' 2 71E+0h i Kr 2.03E+03 - k.5kE+03 3.71E*03 k.31E+03

k. TIE +02 1 51E404 1

Kr-85m 1.26E+03 2.8kE+03 2.18E+03 2 58E+03 2.6TE+02 9 13E+03. Xe-135 k.21E+03 9.kTE+03 7 3kE+03 9.26E+03 9.08E+02 3.1?E+0h Xe-133 1.kkE+03 3.91E+03 2 93E+03 5 96E+03 k.21E+02 1.kTE+0k ~Xe-1k) f . Kr-9k[ Er=93 Xe-1k t Kr - Kr-91 9 72E-01 2.19E00 1.68E00 5 5TE-01 2.06E-01 5.61E00 Xe-140 1.32E+01 2 96E+01 2.28I+01 T.66E00 2 79E00 T.61E+01 Kr-90 1 51E*02 3.40E+02 2.61E+02 9 16E+01 3 20E+01 8.76E+02 Xe-139 - 2.2kE+02 5 0$E+02 3.88E+02 1 38E+02 k.7kE+01 1.29E+02 Kr-89 3 2TE+02 7 37E+02 5.66E+02 2.30E+02 6.93E+01 1 93E+03 Xe-137 5 69E+02 1.28E+03 -9.85E+02 k.10E+02 1.21E*02 3 3TE+03 l Xe-135m 1.18E*03 2.6TE+03 2.05E+03, 1.0TE+03 2 51E+02 7.22E+03 1 Kr-83m' 6.88E+02 1 55E+03 1.19E+03 1.0kE+03 1.46E+02 4.61E+03 3 52E*02 1.2TE+01 6.6kE+02 Xe-133s 6.01E+01 1 35E+02 1.0kE+02 ./ Xe-131s - 2.k9E00 5.60E00 k.30E00 6.15E+01 5 26E-01 7.kkE+01 i l Kr-85 1 97E00 k.kCE00 3.k1E00 k.36u C2 k.18E-01 k.k6E+02 i N-13 9 16E+01 1.28E+02 1.01E+C2 1.35E+02 8.59E00 .k.6kE+02 6 i -+c e-..

APPENCII A. TABLE 2 CONSIMEPS POWER COMPAM Big Rock Point Plant, Docket No 50-155 Atmospheric Releases of Radiometive Material Six-Woth Annual Units July At,r2st _ Setterber Octeder November Deeerber Tot al Total html Noble Cases Curies 2.62E+02 3.b7E+03 k.61E+03 k.kOE+03 5.03E+03 b.96E+03 2.2TE+0L

1. 88DO Total Halogess 1.35E-03 T.kOE-03 3.71E-03 2.k2E-03 1.0TE-02 2.1TE-01 2.k3E-01 3.55E-0 Total Particulates (8, y) 9 9 5 03 3.kkE-03 3 76E 03 T.93E-Oh 6.92E-03 k.10E-02 6.58E-02 9 0TE-0 Total Tritium 6.71E-01 k.18E00 k.0TE00 k.12E00 3 96E00 k.10E00 2.11 D01 3.87E+ 0 html Farticulates-Oross Alpha 2.23E-07 2.395-07 1 92E-07 1.kTE-07 1 30E-07 1 98E-07 1.13E-06 2.80Z-0 MaxizJa Noble Oma Release Rate WC1/s 6.37E+02

1. 53E+03

3. 48E+ 03 1.88E+03 3 36E+03 2.88E+03 3.kBE+03 k.19PO Percent of Tech Spee Limits forr 5

Noble Gases 9 72E-02 1.34E-01 1.78E-01 1.G E-01 1.9kE-01 1.85E-01 1.59E-01 6.13E-0 M*10c m k.18E-C2 T.03E-02 5.OkE-02 5.31E-02 9.k1E-02 1.72 3.38E-01 b.13I-0 Particulates 1.62E-02 k.00E-02 4.TkE-02 2.2TE-02 8.22E-02 2.60E-01 7.81E-02 T.87E-0 Isotopes Released Curies Malogens 1-1 31 1.35E-03 6.63E-Oh 8.52E-Ok 1.b7E-03 k.k2E-OL 1 39E-03 6.10E-03 9 01E-0 1-133 6.TLE-03 2.E6E-03 9 51E-Ok 9.TSE-03 2.16E-01 2.36E-01 2.G E-0 1 135 k.L8E-Ok 1.79E-Oh 6.2TE-OL 6.2TE-0 Particulstes cs-1% 5.G E-OL 3.95E-Oh 1.kBE-Ok 8.kTE-06 T.03E-05 2.08E-OL 1.39E-03 1.59E-0 co-137 T.29E-Ch 1.25E-03 5.92E-Ok 1 3kE-05 3.82E-Ch L.68E-Ok 3.k3E-03 T.79E-0 Bate-1k0 k.62E-05 8.8CE-Ok 1.63E-03 3.76E-03 3.kCE-02 4.08E-02 L.57E-0 En-65 3 97E-0 %-5h k.55E-Ok 6.86E-05 1.00E-Ok 3.26E-Ok 9.k6E-Oh 1.29E-0 Co-60 k.01E-Oh 9.k1E-05 2.65E-04 T.00E-05 T.96E-Ok 1.03E-03 2.65E-03 5 23E-0 Cr-51 7.69E-03 T '9E-03 7.69E-0 Net Un!Dr.tifle3 Beta 3 93E-05 8.15E Ok 1.06E-03 T.01E-OL 1.81E-03 k.38E-03 8.PJE-03 2.1LE-0 Noble Gases Xe-139 1.23D02 1.L8E+03 1 71E+03

1. 30E+03 1.51E+03 1.k6E+03 T.58E+03 5.k BDO Kr-87 2 97E+01 k.k9E+02 5 0AD02 6.65E+02 6.73 +02 T.25D02 3.05D 03 3.02E+0 Kr-88 2 96E+01 3.TkE+02 4.TTE+02 k.65E+02 5 3kE+02 5.k3E+02 2.k 3E+02 1.75E+0 Kr-85a 9 76E00

1. 3k D02 2.0TD02 2.13DC2 2.kkE+02 2 3kE+02 1.OkE+03 1.02E+0-Ie-135 3.87E+01 5 38D 02 8.16D02 8.03E+02 1.0kE+03 9.79E+02 k.21D03 3.%E + 0 Xe-133 6 kOE00 1 56E+02 3 6TE+02 b.09D02 k.31E'02 b.15E+02 1.78E+03 1.65E+0:

ge-lk 3 u.9k 0%93 xe-1k1 Kr-92 Kr-91 1.03E-01 1 59E-01 1.6kE-01 1.88E-01 1.80E-01 T.94E-01 6.kCE00 Xe-1ko 1.02E-01 1.kOE00 2.16E00 2.22E00 2.5%E00 2.kkE00 1.09E+01 8.70D01 Kr-90 1.1TE00 1.60D 01 2.k8E+01 2 57D 01 2.92E+01 2.8CE+01 1.25E+02 1.00D0 Xe-139 1.TkE00

2. 38E+01 3.6TE+01 3.TBD01 b.33D 01 k.16E+01 1.85E+02 3.1kE+0; Kr-89 2.5kE00 3.b7E+01 5 36E+01 5.52E+01 6.32D 01

6. 0TD01 2.70D 02 2.20D O:

Xe-137 k.k2E00 6.OkE+01 9.33E+01 9 59D01 1.10E+02 1.06E+02 k.70E+02 3 8kE+02 Ke-135m 9 20F00 1 26D02 1 9kt+02 2.00 D02 2.29E+02 2.209 02 9 78D02 8.20E+03 Kr-8 h 5.3kE00 T.30E+01 1.13D02 1.16E+02 1 38E+02 1.28E+02 5 73E+02 5 182+03 Re-13 b k.6TE-01

6. 38E30 9.86E00 1.01D01 1.16E+01 1.12E+01 k.96E+01 T.1kD02 1e-131a 1.93E-02 2.6kE-01 k.08E-01

%.19E-01 k.80E-01 k.61E-01 2.05E00 , T.(5E+01 [ Kr-85 1.53E-02 2.10E-01 3 23E-01 3 33E-01 3.81E-01 3 66E-01 1.63E00 .k.kEE+02 N-13 1.6k UO1 1 52D 02 1 GE+02 1.70E+02 1.60D02 1.69E+02

8. 31E+02 1 30DC3

~_ m. 4 n APPENDIX B TABLE l' CONSUMERS FO'JER COMPANY Big Rock Point Plant, Docket No 50-155 Radioactive Liquid Release 4 Six-Month-Units January February March April May-June' Total . Total Radioactivity. Released 1 (Except Tritium, Dissolved T.23 x'10-2 0.606' Gases and Alpha) Curies 9.8 x 10-2.50 x 10-2 1.k6 x 10-1 9 96 x 10-2 1.66 x 10-1 Volume'$f Waste Discharge Liters k.06 x 10 1.18xlok 5.95'x 10 5.45 x 10 8.11 x lo

k.22 x 10

'2 90 x 10 5 . Average Concentration of Waste Prior to Discharge pCi/mi 2.k2 x 10-3 2.12 x 10-3 2.k5'x 10-3 1.83 x 10 2.05 x 10-3 1.T1 x 10-3 2.09 x'10-3 'i

Volume of Circulating 9

9 9 9 9 9 10 " Discharge Water Liters 8.79 x 10 7 9h x 10 8.79 x 10 8.50 x 10 8.62 x 10-8.50 x 10

5 11 x 10 1 Average Concentration Released i

(Except Tritium,-Dissolved -8 Gases and Alpha) pCi/ml-1.12 x 10-3.15 x 10-9 1.66 x 10-0 1.1T'x 10-0' 1 93 x 10 8.51 x 10-9 1.19 x 10 Maximum Concentration 1(Except Tritium,' Dissolved Gises and Alpha) DCi/mi 2.73 x 10-7 1.1h x 10-I 3.67 x lo-T k.96 x lo-I' k.35 x lo-T 3,31,ig-T 4 96 x lo-I Percent of Applicable Limits 5 1 95 0.60 3.28 2.38 k.62 1 55 ~ 2.k0 Tritium Released curies 8.53 x 10-2 2.h8'x 10 1.25 x 10-1 1.1k.x 10-1 1.h6 8.75 x 10-2 1 90 -2 i-L Average Tritium Concentration -9 -g -T 1.03 x lo-g-3 72 x lo-g Released pC1/ml 9 7 x 10 3.1 x 10-9' 1.42 x lo 1.3h x lo-g 1.69 x lo 5 38 x 10 ' 3.51 x 10 6.09 x 10 -6 -5 9.80 x 10-6 3.42 x 10 9 59 x lo- .1.38 x 10 [ -3 Total Cross Alpha Released Curies Average Alpha Concentration pCi/ml 6.12 x 10 h.h2 x'10-13 '6.93 x 10-12 1.15 x 10-12 3 97 x 10-11 1.13 x 10-10 2 70 x 10-11 -12 Isotopes Curies i I-131 1 98 x 10-3 1.25 x lo 3 2.Th x 10-2.27 x 10-3 4.65 x 10 -2 -3 -3 i Cs-13h 1.2T x 10 1 93 x 10-1.26 x 10-2 6.77 x 10 1.39 x 10-2: 7 3T x 10-3 5 52 x 10-2 -2 -2 1 Cs-137 2 57 x 10 6.98 x 10-3 3.02 x 10-2 175 x 10 2.55 x 10-2 7.73 x 10-3 1.1h x 10 2 Co-60 9 15 x 10-3 2.01 x 10-3 1.73 x 10-2 1.07 x 10-2 Bala-140 1.82 x 10-T.28 x 10-3 1.29 x 10-2 5 9h x 10- -3 7.53 x 10-k-8.15 x 10-3 2.63 x lo 2 3~ Zn-65 k.k1 x 10 5.16 x 10 3 1.03 x 10 2 Mn-Sk 1.24 x 10-2 k.58 x 10-3 T.12 x 10 Fe-59 3.43 x 10-3 1.63 x 10-3 2.k1 x 10 3 5 06 x 10-Total Identified Released Radioactivity Curies 4.95 x 10 1.10 x 10 6.01 x 10-2 5 22 x 10-2 5.5h x 10-2.. 4 k3 x 10-2 3,g3,1g-1 -2 -2 Percent of Total Identified 5 50.4 kh.2 kl.2. 52.h - 33.k' 61.2 50.0 4 )

~ m APPENDII Be TAllE 2 CONS'JMERS PCNTR CCfGANY Big Rock Point Flant, Docket No 50-155 Radioactive Liquid Release Six-Month Twelve-Month Units July August September October Neverber December Total Tetal Total Radioactivity Released (Except Tritium, Diseolved Noble Gases and Alpha) Curies 1.60E-01 1.03E-03 2.47E-02 7.15E-02 2.04E-01

h. tile-01 1.07E00 Volume of Waste Discharged Liters 5.60E+04 1.63E+0h 2.01E+04 2.05E+C4 2.96E+04 1.k3E+05 4.33E+05 Average Concentration of waste Prior to Discharge uCi/c1 2.86E-03 6.33E-05 1.232-03 3.60E-03 6.89E-03 3.22E-03 2.47E-03 Volume of Circulating Discharge Water Liters 8.79E+09 8.45E+09 8.79E+09 8.79E+09 8.50E+o9 8.62E+09 5.19E+10 1.03E+11 Average Concentration Released (Except Trit.ium. Dissolved Noble Gasec and Alpha)

LCi/ml 1.82E-08 1.222-11 2.81E-09 8.66E-09 2 37E-08 8.88E-09 1.04E-08 Maximum Concentration (Except Tritium, Dissolved Noble Cases and Alpha) uCi/c1 5.03 -07 7.33E-77 1 96E-07 2 38E 07 3.86E-07 7.33E-07 7 33E-crt Percent of Applicable Limits 2.'f7E00 2.1hE-02 3 79E-C1 1.21E00 6.22E00 1.80E00 21.CE00 Tritium Released Curies 1.18E-02 1.09E00 k.22E-02 4.26E-02 1.98EOO 3 17E00 5.07E00 Average Tritium Concentration Released uCi/mi 1.34E-09 1.29E-07 4.80E-09 5.08E-09 2 30E-07 6.10E-08 4.92E-08 Total Gross Alpha Released Curies 2.04E-05 2.94E-06 1.27E-06 5 5kE-06 9 00E-07 3 11E-05 1.41I-05 Average Alpha Concentration uCi/mi 2 32E-12 3.h8E-13 1.4hE-13 6.51E-13 1.0kE-13 5 98E-13 1 37E-13 Released Isotopes Curies I-131 3 18E-05 6.33E-05 4.84E-04 5.79E-04 5 23E-03 Cs-134 7.38E-03 3 71E-03 6.10E-03 2.82E-03 2.00E-02 7 52E-02 Cs-137 1.60E-02 3.73E-05 9 92E-03 2.29E-02 2.13E-02 7.02E-02 1.84E-01 Co 60 2.59E-02 2 91E-04 1 35E-03 1.08E-02 1.17E-03 3 95E-02 9.89E-02 Bala-140 3.49E-Oh 1.10E-02 1.13E-02 1.16E-02 Zn-65 1.09E-02 1.09E-02 2.12E-02 Mn-54 1.66E-02 8.48E-03 2.71E-02 4.92E-02 Fe-59 5.17E-03 5.17E-03 1.02E-02 I-133 1.53E-04 1.53E-04 1.53E-04 Total identified Radio-activity Released 8.26E-02 3.60E-04 1.50E-02 L.00E-02 4.53E-02 1.63E-01 4.86E-01 Percent of Total Identified 5.13E+ 01 3 50E+01 6.01E+01 5.59E+1 2.22E+01 3 98E+01 k.54E+01

m APPENDIX C Off-Site Shipment of Radioactive Material ~ Shipment Transfer Number Date From Transfer To Radioactive Material Disposition 353 h/ 2/Th DPR-6 Oyster Creek Plant Fuel Handling and Gamma Scan Equip-Reuse 29-12773-01 ment - 5 mci 35h h/18/Th DPR-6 NECo, 16-NSF-1 152 Gallon Barrels (Filters) Burial Sheffield, IL Low-Level Wastes - 349 mci 355 h/2h/74 DPR-6 NECo, 16-NFS-1 15 Gallon Barrels (Filters) Burial Morehead, KY High-Level Wastes - 7 38 Ci 356 h/29/74 DPR-6 NAC Battelle Fuel Bundle Inspection Device - Reuse Columbus Lab 3 mci 3h-06854-05 357 h/30/Th DPR-6 NECo, 16-USF-1 4' x 6' Steel Tank, Misc Waste - Burial Morehead, KY 9 78 Ci 358 h/30/Th DPR-6 NPI c/o Battelle 20 Irradiated Cobalt Rods - Storage Columbus Lab 311,300 Ci 34-06854-05 359 5/ 1/74 DPR-6 Exxon Nuclear Fuel Inspection Tools - 2 mci Reuse Richland, WA WN-1062-1 360 5/16/74 DPR-6 NECo, 16-NSF-1 12 - SS-Gallon Barrels, high-Level Burial Morehead, KY Waste (Filters) - 77 CI 361 6/17/Th DPR-6 GE, Vallecitos, CA 6 Irradiated Fuel Rods - 29,942 Ci Inspection SNM-960, Amend 70-20 & Analysis. 4

APPENDIX C Off-Site Shipment of Radioactive Material Shipment Transfer No. Date From Transfer To Radioactive Material - Disposition 362 7/3/74 DPR-6 G.E. Val-Cal. 5 irradiated fuel rods Inspection and SIN-960 Amend 71-20 29,9h2 Ci Analyses 363 7/31/74 DPR 6 Exxon Nuclear Contaminated fuel handling tools Re-use R chland, Wash. 2 mci i WN-IO62-1 364 10/10/74 Received 10 Ci Cs-137 for instrument calibration. 365' 11/5/74 DPR 6 NFS, W. Valley, N. Y. 6 irradiated fuel tundles . Reprocessing CFS-1 2,355,330 C1 366 11/18/74 DPR-6 G.E. Val-Cal. 7 irradiated fuel rods Inspection and SIM-960 Amend. 71-20 28,729 Ci Analyses 367 11/26/74 DPR 6 NFS, W. Vaney, N.Y. 7 irradiated fuel bundles Reprocessing CFS-1 2,624,090 C1 368 11/26/74 DPR-6 G.E. Val-Cal. 7 irradiated fuel bundles Inspection and SIN-960 Amend 71-20 28,729 Ci Analyses 369 12/6/74 DPR 6 Argonne Nati. Lab. 1 gal. liquid waste sample Analyses Exempt <0.1 mci

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